ML19305B896

Response to Chesapeake Energy Alliance First Set of Interrogatories.Includes Info Re Storage of Contaminated Water from TMI-2 Containment,Primary Coolant Sys Leakage & Radioactive Effluent Monitoring Provisions.Affidavit Encl

ML19305B896
Person / Time
Site:	Crane
Issue date:	03/17/1980
From:	Swartz L NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
To:	Chesapeake Energy Alliance
References
NUDOCS 8003200573
Download: ML19305B896 (48)

Text

72

's?'

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION-BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In.the Matter of

)

3

)

-METROPOLITAN EDISON COMPANY

)

Docket No. 50-289

)

(Three Mile 1 Island, Unit 1)

)

URC STAFF RELPONSE TO CHESAPEAKE ENERGY ALLIANCE FIRST SET OF INTERROGATORIES Pursuantto10LC.F.R.12.720and10C.F.R.'s2.744,theNRCStaffhascoupleted most of its responses to Chesapeake Energy Alliance's (CEA) Interrogatories to the IRC Staff dated February 13, 1980.

Responses not provided today will

'be filed on or before March 31, 1980.

Each interrogatory is restated and a response-provided. Where appropriate, the NRC Staff has invoked that portion of the Commission's Order of August 9, 1979 (slip op. at 11) which allows as an adequate-response to'a. discovery. request a' statement that information is available in the Local Public Document Rooms and guidance as to where the infor-mation can be found. Parties are reminded that any contast with the NRC Staff

~

must be through counsel for the Staff. Signed affidavits identifying the individuals who prepared the responses and verifying them which are not attached will~be sent at a later date.

4 8003200 M 4

._i

y

. '5-8

- INTERR0GATORY:

Provide'a complete and accurate description of the present storage of radioactive water from the Tt41-2 accident, specifying where all that water is being stored, and providing an estimate of the radio-activity levels of the water.

3 RESPONSE:-

The following is a tank-by-tank summary of the status, as of turch 4, 1980, of storage of radioactive water from the T!41-2 accident. The source for the data is the Met Ed/GPU Rad. taste Daily Report, dated liarch 4.1980. As' stated in Section 4 of the SER for Licensee Compli-ance with the NRC Order dated August 9, 1979 (January 11,1980),there is no T!41-2 water now in Tt41-1 tanks.

TMI-2 Inventory Radioactivity Coi,contration Storage Tank (1)

(Gallons)

_(uCi/ml,grossaci.ivity)_

WDL-T-9A(2) --

WDL-T-98 --

WDL-T-11A 1,333 No data WDL-T-11B 60 No data WDL-T-2 16,324 50 WDL-T-8A WDL-T-8B WDL-T-1A 19,946 50 WDL-T-1B.

77,250 50 WDL-T-1C 77,250 50 WDL-T-5 2,970 20 WDS-T-2 9,000 30 Auxiliary Building Sump 3) 4,124-30 Tank Farm. Upper (4) 50,376 30 Tank Farm, Lower (

42,792 30 (1)In addition to-the listed storage tanks, an estimated'550,000 gallons'

~

of water remains in the containment building. This water contains an estimated 270 uCi/ml of radioactive material in solution, k2)Foradescriptionoftanks,'see'FSAR.-

(3)Four 15,000 gallon tanks located in the Tt41-2 Spent Fuel Pool.

(4)Two'25,000 gallon tanks located in the TMI-2 Spent Fuel Pool.

+

g'

+

a spe +e alp.

e

'5-9 INTERROGATORY

~

Provide a' detailed description'of how and where the contaminated water presently in the-containment of TMI-2. will be. stored when it is reaioved _.

from containment during the clean-up of TMI-2.

i

~

. RESP 0!iSE The storage.'of contaminated water from TMI-2 containment, af ter removal and processing, has not been resolved as to fin,al details.

It has been

. proposed that this water, after being treated through a system of the f

general design of EPICOR-2, would be stored in two or more 500,000 gal tanks; two such tanks are now being constructed. The system details and final design description will be contained in the Programmatic Inpact Statement, scheduled-to be published-in draft form in June 1980 and in final form in' September 1980.

4 D**D

• D lY@

oth.1UN L wo e

.I '

t 4

~~

~, '. -.

.l..

~ '

~.

ar

_4

~

10 INTERR0GATORY

~

' Provide'a detailed summary of the p1ans and timetable for the dispo-

'sition'of-the contaminate'd water from THI-2.

1

RESPONSE

Details of the plans and timetable for disposition of the contaminated water from the TMI-2 containment-have not as yet been resolved. The system details and final design description relative to treatment and-disposition of these liquids will.be-contained in the Progrannatic

. Impact Statement scheduled to be publis!1ed in draft form in June 1980, and in final form in September 1980.

GPU has made an agrec...cnt with the City of Lancaster that none of this. water would be discharged to-

{heLSusquehanna River through at least 1982. Tentative plans of GPU are to store the processed and decontaminated water at TMI for re-use in the plant.

D *

• 3D

~

3D %~ X Q

~

M o M J a hL e -

1 L

.,y.

-g-.

m

~

-g 3

_p r

e

5-11

- INTERROGATORY 2

-Identify any and all storage tanks that are presently being used to

. hold TMI-2 water, or that.will be used to hold TMI-2 water, and that ;__

are eithe'r' designed for, have been used for, or might need to be used to store water from TMI-1.

RESPONSE

The TMI facility was designed such that there are a nuc.ber of inter-

- connections between the liquid systems of TMI-l and TMI-2.

In the Ccmmission's Order and Notice of Hearing, dated August 9, 1979, Crit.arion-No. 4 established separation and isolation criteria to pre-vent transfers of liquids between plants. Section 4 of the 1MI-1 Restart Safety Evaluation Report details the procedures for the re-quired separation and isolation. With compliance with Criterion p

No. 4 in effect, the fact that certain storage tanks at TMI-2 might l'

have been designed for, have been used for, or might need to be used 'to store water fran Unit 1 becomes moot.

=i

}D

}D

' 3 Y %"u o

D o

Ju 2...\\.fr o o Ju 1

e s

g m -

g eh-@e all>+

eaM-g g.

g.*

4 e

ig.94@*M q

..w

..,4.

s ;,.

' 5-12.

INTERROGATORY Identify any and all sequences of events at TMI-1 that could lead to the generation of sufficient quantities of contaminated water that sould require use of tanks that presently _ hold, or will in'the future hold, water from TMI-2.

R_ESPONSE E

We have reviewed the. safety analyses provided by the licensee in the TMI-1-FSAR and the safety analyses provided by the staff in the SER-for TMI-1 and find no sequences of events at TMI-1 that could lead to-the generation of sufficient quantities of contaminated water that would require the use of tanks'that presently hold, or are calculated to hold in the future, water from TMI-2.

The design basis for storage

(

~

.of process liquids outside of containment is the calculated volume needed to accommodate liquid inputs from normal operation and antici-pated operational occurrences. The types of accidents which have been postulated in accident analyses demonstrate the potential for releases of large quantities of water inside the containment but do not result in leakage of_large volumes of water outside of containment.

i l

m I

v k_.

6-..

4 Interrogatory __6-8

-Identify the rate, in gallons per minute, at which radioactive water is being discharged or is leaking from the primary coolant' system of TMI-2.

Describe in detail the disposition of'any such water.

Describe

- in detail the measures that'are being taken or that are

.being taken or that_are being planned to reduce or

. eliminate any.suchl discharges or leaks.

d

[ Response Primary coolant system leakage at TMI-2 is approximately 0.2 gallons per minute'according to recent plant data. This value is estimated by knowing the volume of make-up water which must be supplied by the makeup water-system to the. primary _ coolant system to maintain a constant volume. Water leakage.

I from the primary coolant system leaks to ~either the reactor building floor-or-to the radwaste tanks-in.the auxiliary building depending on the source of~

.the leak. Certain measures can be taken to reduce these leaks such as repair of leaking valves and' pipe. connections. Specific repairs made depend upon.the radiation levels near the leaking component. Major leak repair programs are

~

not immediately necessary considering the high' radiation-fields and-the fact that the leak rate'is low.

q l

.j

..a

. ~ -.

. 4-Interrogatory 6 Describe any and all presently unused storage tanks that would be available to receive major additional releases of radioactive water from TMI-2, giving the capacity of all such storage _ tanks. Provide documentation that none of such tanks were designed for use by T.*II-1, and could not be needed for the storage of water that might be discharged from TMI-1.

FCEponse Waste water from TMI-2 can be or will be able to be stored in the following:

(1) The'450,000 gallon Unit 2 Borated Water Storage Tank (2) Approximately 400,000 gallons of waste storage capacity in the Unit 2 9

auxiliary building and fuel handling building. At the present time these tanks contain' approximately 280,000 gallons awaiting treatment in the Epicor-II processing system leaving 120,000 gallons of available capacity.

However as this water is processed in Epicor-II it will free auxiliary building' tankage for additional inputs.

~

(3) The Unit 2 spent fuel pool - approximately 230,000' gallons.

(4) Additional tanks of capacity 500,000 gallons being constructed on site specifically to old THI-2 waste water after it is processed in TMI-2 treatment systems such as Epicor-II.. The number of these_ tanks depend onithe amount of capacity deemed necessary to handle TMI-2. waste water volumes.

0 5

a

{

E

-e seg.m

• gw c ww 4

-.e,%+e.

,em

~ < m-e u g

~(5)-Approximately 220,000 gallons of waste storage capacity in the chemical cleaning building. At the present time, these tanks contain approximately 40,000 gallons of treated water from Epicor-II b

leaving approximately 180,000 gallons of available capacity.

~

The storage capacity described above is not designed for use by TMI-1 and is not needed for storage of water at TMI-1. The f:RC order for the restart of TMI-1, specifically~ Items 4 and 5, require that restoration activities at TMI-2 do not affect TMI-1 and further that the TMI-1 liquid waste system capability is adequate to-assure safe operation of TMI-1.

In addressing the criteria of Items 4 and 5 in the TMI-1 Restart Status Report dated January 11, 1980, the Staff found that the major portion of the TMI-1 liquid waste system is indeperdent of-TMI-2. The only portion that was shared prior to the accident will now be used excl.usively by TMI-1. Thus the Status Report indicates that TMI-1 waste storage capability does not rely on TMI-2 storage.

ggf w kk

~

e

~

g a-

=-

-u

7-8 INTERROGATORY Provide a thorough and detailed description of the radiation monitoring provisions for TMI, describing how such provisions will allow for com -

plete and accurate discrimination between effluents from TMI-1 and those from TMI-2.

Describe any and all meteorological and/or radio-logical conditions that would undermine or nullify the ability to discriminate between TMI-1 and TMI-2 effluents.

RE_SPONSE The radioactive effluent monitoring provisions for TMI-1 are described in detail in Section 11.4 of the TMI-1 FSAR. The radioactive effluent ~

monitoring provisions for TMI-2, and described in Section 11.4 of the TMI-2 FSAR, have been changed by the addition in May 1979, of five effluent air monitor units serving the supplemental air filtration and adsorption treatment system located on the roof of the TMI-2 Auxiliary Building. -One monitor samples plant ventilation exhaust system air at'a point in the main supply plenum just upstream of the mixing and distribution plenum and immediately prior to entering the filtration trains.

Four monitors, one on each train, are located downstream of the f.inal HEPA filters immediately upstream of the four points of discharge from the system.

Provisions allowing for complete _and accurate discrimination between effluents ~.from TMI-1 and TMI-2 are described in detail beginning on_

e page C4-10 of.the Status Report of January 11, 1980.'

~

i O

E7f8 RESPONSE-(Continued)

Meteorological conditions would not undermine or nullify the ability to discriminate between =TMI-1 and TMI-2 effluents because effluent monitoring is performed prior to release to~the environment; however, extreme radio-

~logicaliconditions, such as high levels of background radiation, could be postulated under which readouts of certain liquid or' gaseous on-line con 2 tinuous. effluent monitoring systems could produce erroneously high results.

In the case of noble gas effluents from plant vents, such conditions could result in conservative values of release, i.e., reported releases would be higher than actual releases.

In the cases of liquid effluent releases 1and of radioiodines and radioactive particulates in gaseous effluent releases, a similar condition could be observed; however, a valid reconstruction of actual releases can be made by use of laboratory analyses of the sampling media employed in the monitoring devices or of samples used to determine radioactivity content of batch-released effluents.

A bh

~ Tl D]

• D] e*D n

6 L

,.-.=.r

+w.

e

-

T

7-9 INTERROGATORY Specify for which assumed radioactive. pathways the monitoring pro-visionswillbeabletodistinguishbetweenTMI-1andTHI-2 effluents 5 Describe any and all pathways for which the nonitoring provisions ill be unable to distinguish between TMI-1 and TMI-2 effluents.

i

RESPONSE

See response to 7-8.

j g

e 4 y

.,. ~

N 4 4 Question 8 -

' Summarize and explain the NRC Staff' position on the

~

contention. - Identify-all documents' relied on in' reach--

ing that position.

- Response The NRC St'aff has_not-yet developed a position on Contention 8.

The Staff is

-looking at' the-reports of investigating groups (Kemeny Commission report, Rogovin Inquiry report, and NUREG-0600) which indicate areas of Metropolitans Edison's management. weakness'in its review of.the subject matter of the contention.

~

i a

4 f

I 4

9 4

s d

~h++*-

so *i

=+w em..

em-b-

g F

y e

& m r

.-m-2-

-.r-1 m, -,,.w

,e 4

.<y--

.v

..r.

.. - +

4

-Question 8 '

Identify ~those aspects of the contention that NRC Staff l considers to be matters'of controversy.. For each of those aspects,. summarize briefly the opposing positions

~

'on:the controversy as perceived by the NRC Staff. Identify-

'and suimariza any and all documents in support of either positioi.

- Response

-At this' point in time, since the subject.is still under review, we know of

'no matter _.of controvarsy.

O d

O E

. I 6

4 1

4 e,we e er e a -'m e ege +

me

.M.,m, ee% e w

g i

g.e e*ee

l. '

. Question 8-3 Identify and briefly summarize any and all documents T

known to the f;RC Staff that would tend to provide evidence and or support for this contention.

Rc sponse The documents are as stated in 8-1 above.

em g

ase--

e~

m =

..m

=w m

.c,

,?.i

. Question--8-4

-Identify any and all persons known to the NRC Staff who-have knowledge or expertise-that would tend to support this contention.

For each such person,' pro--

x

- vide name, address, telephone number; qualifications.

- e and a' summary of the. nature of the support (evidence

- or expertise) that person would be capable of provid-ing for-this contention.

39slon_se.

.I know ~of'no-specific person who has stated that he -supports this contention.

a l

f 1

4

=

1 3

S J

e T

/

T w ome -

Ww., _ " m.,,'m._em, 3-%-

y

• nw m

.g..

.wm r,

,.p,.y 4

7.._.

\\

,m 17 --

+

=

Question 8-5

Identify any and all experts _that'the fRC Staff intends--

- to.have testify on the contention; state the qualifica---

tions'of each expert; and present a summary ~of the testi.

mony each expertgis expected to provide.

1

,R,esponse :

. Eri.erts have ilot yet-been identified whowill testify on Contention 8.

-Similarly, a summary of testimony on this subject is~not available at this

~

time.

I 1

i r

)

4 g' '5 I

1, s.

k-l 4

es = 4 e,

d $s9.-

, w

$r me -M

-6

,..I

,,.,., -,_.. - ~+-

,,r..

-y,..

,-..r m..

.y

Question 8-6 Identify any and all members of l4RC Staff who dissent from the overall fiRC Staff position on this contention, T

and for each of these persons, provide a summary of their dissenting position on the contention.

_ Response The I;RC Staff has not developed a staff position on this contention; it is still under review.

3 es t

Question 8-7 Identify the critical or central paramters of this contention as it is perceived and understood by NRC

~

Staff, and briefly summarize the HRC Staff's evaluation of the importance of each such parameter.

Reslonse

- The central parameters of this issue are the licensee's organization structure, lines of authority and responsibility, and the qualifications of personnel.

The Staff has not yet taken a position on the importance of each issue.

O ge

,e we

-e m a

-m

.eese = + = -

o s

,a w..

s

~

- Question 8-8~

-Provide'_a detailed summary of'any and all documents that have been prepared or commissioned by NRC Staff concerning the management ability of Licensee.

In

.particular,-for each such document, identify the

-author (s) and.their qualifications, and describe any

. evidence in such documents that point towards evidence

_of. lack of adequate management ability on behalf of

_ licensee.

1 Response To the;best'of my knowledge, the following documents have-been prepared or commissioned by the NRC Staff concerning the management ability of the license:

'1.

Report to the Commissioners an_d to the Public by Mitchell Rogovin, et al.

2.- Three Mile Island, Unit 2, Radiation Protection Program NUREG-0640.

3.

Utility Management and Technical Resources; contracted to Teknekron Research, Inc., 1483 Chain Bridge Road, McLean, Virginia, 22101, Final Report in Draft.

D 9

Question 8-9 For.any-and all aspect of Licensee's management ability T

for which Staff has found evidence.of inadequacy, describe in detail what measures have been. proposed by staff or-by Licensee to correct such inadequacy, and provide a thorough justification-as to whether staff considers such corrective

^

measures to be adequate, and if so, who.those measures can provide assurance of correction of.. Licensee's management ability.-

Response

The NRC Staff is developing new criteria for " Utility Management and Technical Conpetence." Metropolitan Edison (GPU) will be evaluated against these new criteria.to determine acceptability (or non-acceptability) of their management and~ technical capability. Such an evaluation has not yet been perfornted.

i t

I C

4 ew '. h eemumia b-

=w u.$.

.w e

,g

<. ~..

e

-Interrogatory 12-8

~ Provide a detailed explanation of staff's criteria for T

determining, from the realm of possible accidents, which accidents fall within the design basis.

If Staff's criteria is based on the assumption of single' failure (of systems or congonents), provide a full and complete justification for so limiting design basis accidents, and for excluding design basis consideration of multiple failure accidents.

If any assumptions are made concerning probabilities, provide full probabilities.

Identify all documents relied on, and for each such document, identify the principal authors, their professional qualifications, and relevant publications.

Response

There are no-specific criteria for deteimining, from the realm of possible accidents, which ones fall within the design basis. Over the past 25 years of the regulatory program, thousands of man-years of effort have gone into research, analysis and review of reactor designs with the objective of identify-ing those failures, conditions or events which, while in many instances remote, must be postulated to occur with concommitant threat of release of radioactive materials. From this effort, design basis accidents of a governing nature have evolved and are postulated to occur. Generally, out of a set of similar events ha~ving varying degrees of potential severity, those events which would bound the conditional consequences of the set are selected for detailed analysis.

- These design basis accidents establish the conditions for which a plant and its accident mitigation systems must be designed.

Such accidents are identified for the most-part in 10 C.F.R. 50 and its appendices, in the Standard Review P.lan and in Regulatory Guides. Clearly, identification of variations in design basis accidents-is an on-going process which must take into account the specific-design features of plants, th~e type of reactors involved and accumulating research D

~

r y

'information'and operating experience. While fundamental classes of potential

. accidents have long been identified, such-variations must be accounted for.

[

'This is1 illustrated by the experience at TMI which has caused some changes

~in approach to occur.

The' selection of a design basis accident is not limited by the assumption of a single failure. :In fact multiple failures are postulated if it is. reason-able under the circumstances to do so.

There is"some misunderstanding of the way in which the single failure criterion, er: ployed by the Staff, is applied.

In summary, an initiating event or initiating failure is postulated to occur and then any additional failures that result as

-.a consequence of the event are accounted for. All of these: failures are assumed

-to occur prior to applic'ation of the single failure criterion. -At this point l

the single failure criterion is applied by postulating an additional failure, one at.a time, in each.of the systems employed to mitigate the accident. The

~

sole ^ objective 1is.to enforce sufficient redundancy in the design of these systems to'~ assure functional capability when;needed.

An information paper, SECY-77-439 on the single failure criterion,.gives a more detailed explanation.

Probability assumptions'are not presently used in judging whether an accident should be considered-in -the design ~ envelope.

In lieu of that, decisions are made on the basis of' detailed engineering review'and engineering judgment.

Some exceptions are made in connection with events which may occur off-site--

D*F 0 PD Tl&

S Num -

.gg.

s

-r,---

v-p

+-

e e-y

,A 24-(e.g., railroad accidents) with potential to-affect the facility. Here

'a frequency of:10-7per year is used as'a semi-quantitative test'as to.

whether specific plant protection measures are needed.

3 It'is expected'that probabilistic techniques will be utilized in the future 1

to supplement engineering judgment where possible. The-existing data-base

.is not sufficient-to make probabilistic assessments on an absolute basis, but. techniques have su'fficiently matured to permit assessments on a relative basis and to assist in identifying important failure modes out of a large set of possible failures.

I

- i

~

aMM *- E M

• Mg ?

2d@ @ehu am e - WW sep@ ogw.

'+

%--ath gp eM gle e4*

W ey a

y 3

w e-y p.p

.c

. Interrogatory 12-9 Identify any and all known documents that challenge Staff's

' justification for refusing to consider mu'ltiple failure accidents in ~ developing design basis criteria for nuclear power plant operation.

For each such document, identify the principal author (s), their professional qualifications, and relevant publications. Provide a brief, but detailed summary of the arguments advocated in each such document.

12.sppnse_

The Staff is not aware of any instances in which it refused to consider multiple

. failure accidents in developing design basis criteria for nuclear peuer plant operation.

There have been staff discussions about the possible need to consider

' failure modes different from those usually postulated but these have not been in the context of multiple failures.

Examples of such discussions are contained in NUREG-0138 (Issue #7 - Passive Failures Following a loss-of-Coolant Accident) and in NUREG-0153 (Issue #17 - Passive Mechanical Valve Failures and Issue #22 -

Systematic Review of Normal Plant Operation and Control System Failures).

The' Staff does acknowledge that it underestimated the impact of human errors in aggravating the TMI-2 accident (e.g., closed-AFW valves, failure to recognize

-a stuck-open PORV, and throttling HPI pumps at low system pressure). Reports by the NRC Lessons Learned Task Force speak directly to the need to consider unusual conditions that result from multiple human or equipment failures (see Recommendation 2.1.9 in NUREG-0578 and Recommendations 8 and 9 in NUREG-0585).

The report by the President's_ Commission recommended increased attention to the possibility of multiple failures (see p. 63 and 72). Recommendation 8 in the Report by the NRC Special Inquiry Group also emphasizes the need to consider

" multiple human and equipment failures.

D*"]D *g~9~p~

Jo oau o ju A dd.,

Jnterroga_ tory 12-10 Provide a bibliography (annotated for key documents), of the literature of the probabilities and consequences of major nuclear power plant accidents.

Provide Library of Congress call numbers, whercever such call numbers are known to the NRC Staff.

Response

Attached are bibliographies on these subjects.

h 6

.o w

.esa m

,.m eme m

n.

o Interrogatory 12-11~

Identify an IRC Staff person or. persons who is thoroughly familiar with the material in the-bibliography requested

. in112-10 above, and upon whom-CEA may freely call to dis -

cuss with and consult upon the matters.in. Contention 12.

In the event there is no such' person on the f4RC Staff, so state, and provide.ti)e name of any other such person to v.hom CEA could reasonably have access for such discussion 4

cnd consultation; furthermore, in this event, provide a full and complete justification as-to why f1RC has no such

. staff erson with the above familia'rity with these matters.

R_esponse The name of the. cognizant flRC Staff person familiar with these references

' is noted at the bottom'on each bibliography. CEA is reminded that it may not " freely call" the Staff,~but must seek information through counsel-for the Staff.

)

i I

4 s

4 7-

~

4 --.

r

,-w-

-, e

,.,-.,n-

-n,-

Bibliorfaphy of Class 9 Accident Studies

. l.

Reactor Safety Study, WASH-1400, U.S.H.R.C., October 1975.

T Three liile Island--A Report to the Con.nissioners and to _ the Public,

2..

!;RC Special Inquiry Group, January 1980.

3 3.

Effect of Containment Venting on the Risk from LUR l'altdorin Accidents, HUREG/CR-0138, Battelle Columbus Laboratories, June 1978.

4.

A Value-6r. pact Assessment of Alternate Containment Concepts, I;UREG/CR-0165, Sandia Laboratories, June 1978.

5.

Generic Evaluation of Feedt ater Transients and Small Break Loss-of-Coolant Accidents in Uestinghouse-Designed Operating Plants, : UREG-0611, U.S.N.R.C., January 1980.

6.

G:-neric Es aluation of Feedt ater Transients and Small Break Loss-of-Coolant Accidents in Combustion Engineering Designed Operating Plants,

!.UREG-0635, U.S.N.R.C., January 1980.

- 7.

Analysis of the Three 14ile Island Accident and Alternative Sequences, I;UREG/CR-1219, Battelle Columbus Laboratories, January 1980.

8.

Generic Evaluation of Small Break Loss-of-Coolant Accident B:havior in Babcock and Wilcox Designed 177-FA Operating Plants, I;UREG-0565, U.S.N.R.C., January 1980.

9.

Staff Report on the Generic Assessment of Fredtater Transients in Pressurized Water Reactors Designed by the Babcock and Wilcox Cc..;3ny, i;UREG-0560, U.S.N.R.C., liay 1979.

10. - Risk Assessment Review Group Report to the U.S. Nuc1 car Regulatory Commission, NUREG/CR-040u, Ad Hoc Risk Assessment Review Group, September'1978.

11.

Techn~ical Report on Anticipated Transients Without Scram for Uater-Cooled Po.-Jer Reactors, WASH-1270, Regulatory Staff of U.S. A.E.C., September 1973.

Anticipated Transients Without Scram for Light Water Reactors, i;UREG-0460, 12.

U.S.N.R.C., April 1978.

Refer to Gordon Edison - (301) 492'-8377.

I D s

e o

e 0 sy

• M=

o-s-

.ws,m---

s -

REFERENCE LIST OF STUDIES AND IMPROVEMENTS TO CRAC Overvic.w of the Reactor Safety Study Consecuence Model, N U REG - 0 3 -10, U.S. Nuclear Regulatory Commisil51, Wasiil~n~I[tiin, DC"~2WS55, June 1917.

Aldrich, D.

C.,

P.

E.

McGrath, and N.

C.

Re s truss en, Ex a ni n,a,t i on of Of fsite Radiolog~ical Energ~ency Prote ctive Mear.u res~~for~

?!u..c l e.a r R.e.c.c. t.or A c c iTc'ni~ ~In v ol viiir~ ~Co re l'eiti,~ ~ Sin d i a Lib o rat o ri e s '

s

. Report -(to be published).

Ef f ects of Wind Shear on the Consecuence Model of t he Reactor Safet_r

Study, J.

L.

Sprung and H.

W.

Church, S AN D7 6-0 619, NUREG -017 5, Sandia Laboratories, Albucuerque, NM, 87185, J anuar y 19 7 7.

Sensitivity of the Reactor Safety Studv Consecuence Model t,o. Mix,in,g Jia i,l: hts _, J.

L.

Sprung and H.

W. Church, SAND 7 6 -0618, NUREG-0174, Sandia Laboratories, Albuquerque, NM, 87185, January 1977.

Influence of Plume Rise on the Consecuences of Radioactive Material,

,Re3caqes, A.

J.

Russo, J.

R.

Wayland, L.

T.

Ritchie, SAND 76-0534, Sandia Laboratories, Albuquerque, NM, J a r.u a r y 1977.,

.E.$ f 0 c. t. s,o f R,a i n s t o r m a nd Ru n o f f o n Co n s ec u enc e s o f Nuc l e a r _Re a.c t o r,

,Acciden.t.s_,

L.

T.

Ritchie, W.

D.

3rown, J.

R. Wsyland, SIND76-0429, Sandia Laboratories, Albuquerque, NM, October 1 9,7,6..

McGrath, P.

E.,

D.

M.

Ericson, Jr.,

and I.

B.

Un]1, "Tne Reactor Esfety Study (WASH-1400) and Its Implications for Radiological Enercency Response Planning," International Symposium on the Randling of Radiation Accidents, 2 8 Feb ru ary 1977, vienna, Aus tria, I AEA--SM-215/23 Aldrich, D.

C.,

D.

M. Ericson, Jr.,

and J. D. Johnson, Pub 1_i c, Protection Strategies for Potential Nuclear Reactor Accidents:

-~'-

Sheltering Conceots with ExistT5g Public and Private, St.ructures, SAND /7-1725, Sandia Laboratories, Albuquerque, NM (1977)T Aldrich, D.

C.

and D.

M.

Eri cson, Jr., Public Protection St,r3tegies in the Event of a Nuclear Reactor Accident:

Multi-

~

compartment Ventilation Model for Shelters, SAND 77-1555,

'S'ahdia Laboratories, Albuquerque, NM (19 ~/*[).

~~

Aldrich, D.

C.,

R. M. Blo$d, and R.

B.

Jones, A ?'odel of Public Evacuation for Atmosoberic Releases, SAND 78-0092, Sandia Lhboratories, Albuquerque, NN (19 %).

Ericson,.D. M.' Jr.,-Accident De s c ri pt_i_on s,fo r, fney,.p n ey, Pl anne rs,

Sandia Laboratories' Report (to be pubitshed).,

Eufer to Roger Blond - (301) 492'-8333, D**

'y0 a t..]

Q

+

-30 Emergency P1anning Research 1:3 thin RES/ PAS Since t he RSS

~

The foilowing reports have been published:

r 1.

McGra t h, P.

E.,

D. M. ' Eri cson, Jr., and I.

B.

Wall, "The React or Safety Study (NASil-14 0 0) and I t.s Irpiications for Radiologic 31 Emergency Response Planning," Inter national Symnosium on t.he;

'.H A.A.d,1,i.n g, p,f_ ytq,i a t. i on _Acc_i de n t g, 28 February 1977, vienna, i,

Austria, IAEA-SM-215/23.

[chnsor, Publ,ic, I'

2_.

Aldrich, D.

C.,

D.

M.

Ericson, Jr.,

and J.

D.

P r o t e c t i o n S t r a t o o,.i_ e_.s__.. f_o.r_..P_o_t_.e.-.n.t..i a. l. N u c.l.e a. r. _ _R. e a.c.t o..r

.A..c c..i.d..e_n..t..s..:.

She l t e ri ng Concep't s wi t h E x i s ti ng "AY1sug u'6W[ie' ~~ f'f '( l 9'f7

~

Publi c and Pri va t e St ruct.ures,

S ATiD 7 7 ~1'/2 5, Sandl a Labora toric's,

. ' ~ ~ ~ ' ~

~

s 3.

Aldrich, D.

C.

and D.

M.

Ericson, Jr.,

Public Trot ectica Stratecies in t h e E_ve._n t. o. f a N uc i c a r Ec E6t'o r 'Xc h i d 6n' t s':

Multi-

~

~

~

comoartment Ventilation Model for Shelters, SAND 77-1555, I

5Fn'di a I. abo r a t oil'c's, Albuquerque, NM (1977).

4*

Aldrich, D.

C.,

R.

M.

Blond, and R.

B.

Jones, A Model of Public E.V^Fla t.ipp fo r.A t mospher i c _Rp_lppp,p_s_, S AN D 7 8 - 0 0'9'2', fandia' l'

~" '~~

Lesbo ra t orie s, Albuquerque, NM (19 78)'.

5.

i

Ericson, D.

M.

Jr., Accident Dascriptions for E;;.e t genc,y ResponS9, Exercise Scenarios, S'!.S D 7 8 - O'2 697'S a nMa "Il6o r'a't o r i e s, Al buq ue r t;ue,

~~

~

~

iiciHFxL&o.'~.1XKiBE _

6.

Aldrich, D.

C.,

P.

E.

McGrath, and N.

C.

Ra s..uns en, Exa:n i n a,t i o n, of Offsite:Radioloolcal Eme rgency, P rot ec_, i ve, gg s ures_for j

t

uclear Reactor Accidents Involvino Core Melt, EAND78-0454, t

SandiTTEEora t ori es, A IEuq ue rq ue,'~$M (1978).

I 7.

Cohen,.A.

F.,

and B.

L. Cohen, Infiltration of Part,iculate Matter into Buildings, SAND 79-2079, Sandia Laboratories, Albuquerque, NM (1979).

- 8.

Aldrich, D.

C.,

D.

M.

Erikson, N.

C.

Fi nl ey, N.

R.

Ortiz, J.

L.

Sprung, and J.

M. Taylor, Energ,ency_ Response Scenarios for Transoortation Accidents Involving Radioactive i*aterials, SAND 79-2017, Sandia - Laboratories, AlbugtErgue, !!!! ( OI t65er 19 79 ).

~

Refer 'to Roger Blond

.'(301)-492-8388.

b k

M J JutlL g

I q

g t

w.

' 31-Interrogatory 12-11 (2nd) gu_e's tion Specify, in'nucerical. probability. terms per operating reactor year, the probability of an accident below which probability the accident is not considered credible by the !!RC staff.- Provide a full and complete justification for the I:P,C's selection of that probability 1cvel as the cutoff point for accident

credibility. _ Identify any and all. documents relief upon in this answer.

t-

._?.psponse Probability assumptions are not presently used in judging uhether an accident A,

should be considered in.the design envelope.

In lieu of that, decisions are hade on the basis of detailed engineering review and engineering judgment.Some.

exceptions are made in connection with events which may occur off-site o

(e.g., railroad accidents) with potential to affect the facility.

Here a fraquency of 10-7 per year is used as' a serai-quantitative test as to whether specific plant protection measures are-needed.

f o g o g 9 3-U_o o M o M 11 m

O

?

• e I'

4 P -

+

l= :g L

' Interrogatory 12-12 Question

~ Describe in detail the process relied upon by NRC staff in computing the probability of any given accident sequence.

Provide detailed and thorough

[

justification for this computational method, paying particular attention to the'~

respective determination of the role of human (operator) error as compared to roles of equipnent and instrumentation-information malfunction.

For each of the above identified three components of error (operator, equipment, and instrumentation-information) demonstrate the extent to which the assumed probabilities are based on past experience or on other estimates. Where past experience is used as a basis for probability estimates, state whether the

I estimate is based on past experience with identical reactors and control room designs to TMI-1; if not, state what methods are relied upon to take into account the specific reactor and control room-dependent characteristics of TMI-1, as well as the quality and effectiveness of operator training and emergency procedures at TMI-1; if there is no such consideration of TMI specific parameters of probability in determining accident probability, provide a full and thorough justification why such consideration is not needed.

If estimates other than those based on past experience are utilized in calculating probabilities, provide a detailed justification and basis for those estimates.

Response

The NRC does not rely upon calculations of the probabilities of accident

- sequences in the regulatory process. The staff-is developing methodology and making trial applications of probabilistic safety analysis techniques but these _

are not now employed in licensing nor are their quantitative results regarded

, as generally reliable. -It is anticipated that those safety insights that are

~

e h

m n

'ne

LfM=RRt

- relatively insensitive to the accuracy of the probabilistic results may become

'th' ' basis for regulatory ' action, but there are not' plans to employ numerical e

probability criteria in regulation.

7 flowever,' accident sequences identified in WAsil-1400 are examined taking accounti of specific plant differences.

Data are obtained from plant operational'

~

J experience.and from selected portions of WASH-1400 after critical review. No analyses of this type have been yet undertaken for TMI-1.

l 4

l 3

G i;..

+ a.

34-Interrogatory _12-13 Question Identify any and all exper ts who were relied upon in providing the answer to 12-12 above; for each expert provide name, address, phone number, and professional qualifications.

j

Response

Dr. Gordon Edison of the Probability Assessment Staff aided in respone to 12-12.

e e

e g

,e e

e ameme =, e wm.

=-,

=

o we-=

= - e e-m-

e

l

. Interrogatory 12-14 Question Identify and and (sic) all documents that were relied upon in providing the answer to 12-12 above; for each document, provide author (s), and their 1

professional qualifications.

R_.e.sponse h' ASH-1400 vias used in response to 12-12.

my c,v y:

3 M

cw c

m

(

Interrogatory 13-1

=

Summarize and explain the NRC Staff position on the contention.

Identify all documents relied on in reaching that position.

Resp,on se -

In order. to assure that operating personnel maintain an overall perspective of-plant conditions, it is the Staff's position that operator training instructions must emphasize the use of various plant parameter indications when evaluating plant conditions (reference IE Bulletin 79-05 A, item 4.d).

In addition,

' operators must receive training in the revised emergency procedures, which now include requirements to re-verify key parameters using alternate indications and instructions to verify adequate core cooling (reference UUREG-0578,-item 2.1.9).

Finally, the presence of the Shift Technical Advisor (reference NUREG-0578, item 2.2.1.b) will further assure that a consideration of overall plant conditions is. maintained during transient operations.

As noted in the Status Report, the licensee has complied with the Staff's requirements outlined above.

Therefore, the licensee has performed training, revised procedures, and installed a Shift Technical Advisor to provide reasonable-assurance that a "mindset" which denies information indicative of serious reactor.

problems willLnot exist.

om m

g-J a v Ju o M 11Lnis

+

• • .4 e*

_ weese. nee-e

-p

+e

~

=

y Interrogatory 13-2

/ _

Identify those aspects of the contention that fiRC Staff

~

considers to be matters of controversy.

For each of those aspects, summarize briefly.the opposing positions.

on.'the~ controversy'as perceived by the fRC Staff.

Identify and summarize any and=all documents in support of either position..

Response

We know of:no' aspect of this contention which are considered to be matters of controversy.

do]M 90 3'Y A 9 D

! D e M 2..k

a 9

)

)

i w

1 p '

e e dg e

,a w W p.

, yg, mp p w,,

g,,

'f~

,., +

i

) -

4 Interrogatory 13-3.

Identify and briefly summarize any'and all documents T

'known to~the f4RC Staff that would tend ~.to provide evidence and or support for this contention.

R 7 esponse i

We'know of no documents which provide evidence-or. support the contention.

the specific training provisions designed to address a "mindset" problem

~

are necessary prior to the restart of TMI-1.

.s 4

4 4

i s

' 6 1

Y v

A s-

...,a_...

...,,.~,-r

Interrogatory 13-4

- Identify any and all' persons known to the flRC-Staff who have. knowledge or expertise that would tend to support this contention. For each such person, provide name, address, telephone number,-qualifications, and.a summary of the nature of.the-support (evidence or expertise)'that

~

person would be capable of providing for this contention.

j Response-There are no persons known.to the NRC Staff who 'have knowledge or exportise p

,that would tend to support'this contention.

A If -

1.

f

(.

4 M

F

% M4

.m a w e

g.-

e

,p, g

m

'4e i

,p

-s0-Interrogatory 13-5 Identify any and all experts that the flRC Staff intends to have testify on the contention; state the qualifica-T tions of each expert; and present a summary of the testi-rnony each expert is expected to provide.

, Response The experts that the NRC Staff intends to have testify on this contention have not yet-been identified.

9 i

, ~......

..~.

.c

' Interrogatory 11'3-6

~

. ldentify any and all members of.NRC Staff who dissent 7

' from the overall.NRC Staff position on this contention, and for each of those persons, provide-a summary of their dissenting position on the contention.

. Response We know of no members of-the NRC' Staff who dissent from the overall Staff

- position on this contention.

d J-f 4

• IdPO*'

+'+

{ Pf m'M

N M Mw gg% mg 4 yg

-q

,4W g_q#,

e ggm

,g, 4

t-_..

I 9 4 Interrogatory 13-7 Identify the critical or-central parameters of this

~

. contention as it is perceived and understood by NRC

~

Staff, and briefly summarize the NRC Staff's evalua-

~ tion of the inportance of each such paraaeter.

Response

The centra'l parameter of this con'entiontis perceived by the NRC Staff as t

being whether operator training is conducted prior to the-restart of iMI-1 g

to assure operators maintain an open mind when evaluating plant conditions.

.The Staff. feels that the areas of operator training, procedure revision, and 4

technical support are important'in-this regard.

As noted in the response to Interrogatory 13-1,-the licensee has complied with the Staff's requirements,

in these areas.

i-

@h E

[-

l i --

j.'

. - ~

' Interrogatory 13-8 Describe in detail any and all-screening procedures, b

known to the NRC, to detect the development or existence of an operator 'mindset',.wherein an operator is so conditioned on the basis of his experience, in conjunc-tion with prevailing mangement and operating attitudes, to sustantially rule out an interpretation of a sequence of alarms, adverse signals, and indications of abnormal transients, as indicative of a major accident with con-sequences or a partial or full core melt.

If any psycholo-gical tests are so used as screening procedures, identify such tests, identify the author of the tests, and his/her qualifications and professional background, and summarize any research that has investigated whether the tests pro-vide an accurate indication of the existence of such an operator 'Mindset'.

Response

.No screening procedures or psychological tests to detect the development or existence.of an operator "mindset" as defined above are known to the NRC.

However, as noted in Section Cl-l$ of the Status Report, the NRC intends to conduct written and oral examinations on all TMI-l licensed personnel.

These examinations will require operators to demonstrate the importance of maintaining an overall perspective of plant conditions. This will provide reasonable

-assurance that a "mindset" does not exist.

D N

..s 4 i

t 1

Interrogatory 13-9 Describe in-detail any screening procedures that will be C

used by Licenseefto detect.the development or existence

~~ S of an operator 'mindset' (as defined above). Sucmarize any research that has been conducted to evaluate the efficacy ~of such screening procedures, identifying the principal investigator (s), and providing their profes-sional qualifications.

Response

The NRC Staff is not aware that the licensee intends to use any screening pro-cedures to detect the development or existence of an operator "mindset" (as

' defined in interrogatory 13-8). However, as noted in Section 6 of the Status

. Report, the licensee intends to have an independent auditor group conduct a written and oral evaluation of all TMI-1 licensed personnel. Aniong the areas expected to'be covered is inadequate core cooling. Testing in this area will ascertain whether an operator.has been trained to recognize, in spite of. con-flicting indications, conditions which are indicative of a major accident which could lead to inadequate core cooling.-

D I'

-.-i w.....-..

r y

W --

w er-r s

v

9 W

Interrogato_ry'13-10 Identify any and all known research investigating the relationship between the development of operator ' mind-set' and the frequency of alarms o~r indications of minor abnormal transients, and/or the frequency of false alarms or other false-indications of abnormal transients. For each.such research, identify the principal investigator (s) describe their professional qualifications, and summarize the findings ~of the research.

_R.isponse We;know of no research that has been performed which investigates the relationship between the development of operator "mindset" (as defined in Interrogatory 13-8) and the frequency or alarms or indications of minor abnormal transients, and/or the frequency of false alarms or other false indications of abnormal transients.

e 1

-w e *

• m m

-'e.m,

.,ge.-

ew.,4,,,4 w,,,,,

i

e, u

^

_g 46-Interrogatory 13-11 Identify and describe any and all studies that havei developed,' commissioned, or planned by !;RC into factors associated with the development of operator mindset..

If no such studies have been prepared, commissioned, or planned by fiRC, provide full and detailed justification why such studies are not considered to be necessary.

Response

The llRC plans to review the operator training, licensing, and requalification processes as detailed in Action Plan (flVREG-0660) Items I.A.2 and I.A.3.

The improvements anticipated by.these items and those already noted as being irplemented in the response to Interrogatory 13-1 will upgrade operator aware-ness and response to abnormal situations. We feel taese changes will satis-factorily resolve the concern of operator "mindset" as defined in Interrogatory

' 13-8.

6 9

.mW.WG'

+ 4B=%

S Wu hM

+e**64 9e M-MC

7

.2-.

-47_

4 Ir t_crrogatory 13-12 n

Describe in detail any and all communication and dialog that has taken place between NRC and professionals with experience and research into operator mindset in situa-tions analogous to nuclear power plant control rooms, for example, personnel in the IJASA Aviation Psychology program.

If no such communication and dialog has taken place, provide a thorough justification as to why that has not been considered necessary or valuable.

P sponse 3

llo comunications are known to have taken place between the ftRC and professionals with experience and research into operator "mindset" in situations analogous to nuclear power plant control rooms.

This is because no such communication was recommended by any of the groups investigating the T!41-2 accident.

Respectfully submitted, A

Lucinda Low Swartz Counsel for fiRC Staff a

f fiar 8b.

9 o@N

UNITED STATES OF AMERICA

~ fiUCLEAR RECULATORY CGIG1SS10tl BEFORE THE ATOMIC SAFETY AllD LICEt4Sil:G BOARD In the Matter of 3

METROPOLITA!i EDISOfi COMPAriY, et al.

)

Docket tio. S0-289

)

(Three Mile Island, Unit 1)

)

ATFIDAVIT OF FRA!iCIS P. CARDILE I, Francis P. Cardile, being duly sworn, do depose and state:

1.

I am a !!uclcar Engineer in the Office of Standards Ccvelopm0nt of tie United States fluclear Regulatory Commission.

I am generally r.n,.onsible for star.dards related to waste management.

Previously, however, I.tas a

• vclear Engineer in the Office of f!uclear Reactor Regulation.

I was resp;nsible for review and evaluation of rad. caste treatmant sys t:.ms and for the calculation of releases of radioactivity from nuclear rm.er raciors.

My professional qualifications statement is attached.

2.

The answers to Lewis interrogatories 12,13,16,17, 22,. 28, 32c, 35; CEA interrogatories 6-8, 6-10; and A'!GRY ' interrogatory 6 were pr: pared by me.

I certify that the ansviers given are true and accurate to the t.est of my knowledge.

0

,M j.. s<M Francis P. Cardile Sobscribed and sworn to hefore me this llth_-day of

.b

/

?;otary PubTic "y Co mission expires':

' July 1,1982

,