ML19305A733

Second Set of Interrogatories Directed to NRC Re Status Rept on Util Compliance w/790809 Order.Requests Info on Emergency Feedwater Pump Motors,Analysis of Small Breaks & IE Bulletin 79-05A Requirements.W/Certificate of Svc

ML19305A733
Person / Time
Site:	Three Mile Island
Issue date:	02/01/1980
From:	Weiss E SHELDON, HARMON & WEISS, UNION OF CONCERNED SCIENTISTS
To:	NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD)
References
IEB-79-05A, IEB-79-5A, NUDOCS 8002130122
Download: ML19305A733 (12)

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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD (( D i 6 91.2 6\# t g5 ,'

In the Matter of )

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METROPOLITAN EDISON COMPANY Docket No. 50-289 abl lT {s

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, (T hree Mile Island Nuclear )

S ta tion , Unit No. 1) )

UNION OF CONCERNED SCIENTISTS i

SECOND SET OF INTERROGATORIES TO NRC STAFF -

Please apoly the same instructions as in UCS's first set of interrogatories. Unless otherwise noted, the nage references are to the "S ta tus Report on the Evaluation of Licensee's Compliance with the Order , Dated August 9, 1979."

190. Page Cl-10 contains the staf f's evaluation of the environmental qualification of EFW pump motors and control valves.

a. Identify the specific regulations, Regulatory Guides , Branch Technical Positions or other documents containing the standards used , or to be used, as the basis for this evaluation.

b. Is it the staff's position that testing in a simulated adverse environment is not required for these components ? I f so , exclain the basis for that position.

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4 191. With respect to the discussion of Iten Id, "Analy-sis of Small Breaks" on op. Cl-ll to Cl-14, please answer the following : ,

a. Define " adequate core cooling" as used on line 13 of page C1-12.

i

b. Describe the operator actions needed  :

during heat removal by the steam generators and high pressure injection system, as discussed on I lines 9-13 of page C1-12.

c. What is the basis for expecting that the operator would terminate HPI before the PORV or i u

safety valves lift? (line 34 on page Cl-12 ) ,

1

d. What are the consequences if the operator fails to terminate HPI before the PORV or safety valves lift?  !

e. Has the staff analyzed a small break with loss of all main feedwater, an isolated PORV and a safety valve stuck open to determine whether this would result in uncovering the reactor _ core? ,

If so, what were the results? I f not , why not?

f. Identify the specific operator actions that are required for a small break accident, dis- i tinguishing between the immediate and follow-up ac tions . (line 24 on page Cl-13)

g. Por each of the action described in the ,

answer to subpart (f) above , identify the soecific

" circumstances" during which the action is required.

(line 25 on page Cl-13) 1 I

I

_. . ~ .

h. The staff states on line 27 of page Cl-13 that immediate operator action is required "as soon as the problem is diagnosed. " Is it the staff's position that the amount of time required to diagnose the problem has no bearing on the consequences ?

i. For each operator action identified in the answer to subpart (f), above, specify the earliest and latest times after the break during which the a operator action must be performed. In other words, give the " window" of time after the break during which the operator action must be taken, i j. Of the time available to accomplish each necessary operator action, how much of the time is l available for diagnosis and how much is available to actually accomplish the action?

k. For each of the required operator actions, specify the information available to the operator ,

to make the diagnosis and to confirm that the action has been accomplished.

i

1. In assessing whether it is reasonable to expect the operators to take the actions identi-fled and discussed above at the correct time, has the staff considered the other events occurring in

, the plant which could distract the operator or otherwise demand his. attention?

4 i

m. If the answer to subpart (1.) above is "yes,"

identify all other_ actions the operator may be  !

required to take on other systems and the alarms that can reasonably be expected to occur during that time.  ;

l 192.

With respect to IE Bulletin 79-05 A, Item #2 (p. C2-2),

what was the staff's basis for requiring review of only those i transients similar to the Davis-Besse event which occurred at l TMI-l rather than a review of all similar transients at all l

B&W facilities ?  :

193. Is it the staff's position that a desion (as distin-

[

quished from procedures or training ) which permits operators to override automatic actions of engineered safety features before i the safety function goes to completion meets the commission's ~

{ regulations. If the answer is "yes", explain the basis for that pos it io n . For example (this is only an examole ) , does the i

staff take the position that a design which cermits operators i to prevent a core cooling system from going to completion meets the regulations ?

194. Pages C2-5' and C2-6 contain a discussion of the .

procedures to be used to assure that safety-related system f valves are in the correct position. Does the staff take the -

i position that conformance with this part of the order will provide a degree of protection to the public equivalent or '

superior to what would be provided if the design of TMI-l met ,

Regulatory Guide 1. 47? ,

i 4

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195. The staf f notes the " extreme seriousness and consequences" of the simultaneous blocking of all auxiliary feedwater trains. (p. C2-8) Describe the extreme serious-ness and consequences" referred to.

19 6. Describe the events that could lead to "RCS void formation that could interruot natural circulation flow. "

(line 35 of page 2-9) ,

197. With respect to the discussion on oage C2-10 ,

please answer the following:

a. Describe the circumstances under which continued operation of engineered safety features would threaten reactor vessel integrity.

b. Describe the range of consequences to the public that are possible if reactor vessel integrity is lost.

c. Does the staff take the position that reliance on operator action to prevent loss of reactor vessel integrity meets the Commission's regulations? If so, reference the specific regula-tions that permit this.

198. With respect to NUREG-062 3, please answer the

following

o a. Identify the non-LOCA transients for which the

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consequences are aggravated by reactor coolant pump l trip and describe the extent - of the aggravation?

5 (NUREG-0623, p. 1) t

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l I b. Has the staff evaluated the effect of j reactor coolant pump trio during a large r

break LOCA with respect to the probability ,

of pump (and flywheel) overspeed? If so, [

i provide the adalysis and conclusions.

a r 199. NUREG-0623 states that "small break LOCA's with the t a

i pumps operational or with delayed trip can result in more severe t

consequences than when the pumps are tripped early in the I accident." (p. 1) In contrast, the staff states on cage C2-18 i of the Status Report that "the proposed logic [the automatic i pump trip] is intended to preclude pump trip during . . . very small breaks where maintenance of forced cooling is desirable."

Please answer the following:  ;

i

a. Specify the spectrum of break sizes and '

locations where maintenance of forced cooling is desirable and those where pump trip is i

desirable.  ;

b. Discuss the means by which the operator '

will be able to determine whether the pump 1

should be tripped or forced cooling should t i

be maintained. l i

200. In Section C8, the staff lists the " positions" taken in NUREG-05 78. Following each " position" is a section ,

l entitled " clarification." Is it the staff's position that compliance with the items in the " clarification" sections are necessary and sufficient to demonstrate compliance with the

" pos itions ?"

201. Page C8-6 contains seven itens of " clarification" with regard to the NUREG-05 78 position on power supply for pressurizer heaters . Please answer the follow:

a. What is the basis for item 3?

b. With respect to item 4, what is the basis for requiring change-over of the heaters to emergency power to be accomplished manually rather than automa-tically ? Identify the regulations, if any, which require this.

c. Is it the staff's position that the safety-grade circuit breakers referenced in " clarification" item d6 can be made to conform to the requirement for an isolation device set forth in Reg . Guide 1.757 202. On oace C8-14, the staff states :

1 The similarities between the instruments used at Three Mile Island tinits 1 and 2 lead us to conclude that the detection of  !

reduced coolant level or the existence of core voiding in TMI-l can be readily deter-mined with the existing plant instrumenta-tion, provided the operator is aware of the available information and interorets its correctly.

Is it the staff's position that this meets IEEE 279, 64.8, as -

l incorporated in 10 CPR 50.55 ( a )( h )? If so, specify what "the desired variables" are and explain -how the existing instrumenta-tion provides direct measures of those variables.

203. In several instances _ including for example, primary coolant saturation meters . and feedwater flow meters ,

l the staff is requiring and/or has found. acceptable the use of only two indicators.

i

9 4

a. Explain how the operator is expected to discern which is correct if the two give differing readings,

b. Is it the staff's position that such designs conform to the requirement contained in IEEE-279 that "the design shall minimize the development of conditions which would cause meters, annunciators, recorders, alarms, etc., to aive anot.alous indications confusing to the operator?"

204. On page C8-22, the staff states that it "has deter-mined that post-accident operation of the reactor coolant pumps is highly desirable." Please answer the following:

, a. Identify the particular accidents for

.I i which reactor coolant pump operation is i " highly desirable."

a

b. Describe in quantitative terms the difference in consequences for each of the accidents identified above assuming 4

first that the reactor coolant pumps are a

operating and second that they are not operating. If no detailed evaluation has been done for the case where the pumos are operating, provide you best estimate ,

c. Has the staff evaluated whether. classify-ing the reactor-coolant pumps as safety-related.

i ll 4-4

i and providing an on-site power supply (or any other means of providing forced cooling of the core following an accident ) would i provide substantial, additional protection for the public? If so, what were the

! results of the evaluation? If not, why not?

205. With respect to the discussion of isolation of the reactor coolant pump seal injection lines on pages C8-22 and C8-23, please answer the following:

a. Describe the evaluations the staff has '

done to determine whether the health and i

safety of the oublic is better protected by not automatically isolating the seal ,

injection lines or by isolating them.

b. What is the staff's judgment concerning the probability of a loss-of-reactor-coolant if the seal injection lines are isolated?

Consider in your answer the probability of loss of off-site power, the new procedure for the operators to trip the reactor coolant pumps and the addition of a means to automati . -

cally trip the reactor coolant pumps.

l

c. What information is available to the .

1 operator to indicate the need to manually l

! isolate the seal injection lines? ,

d. In approving a design which does not ,

provide for automatic isolation of the seal

l l

injection lines, did the staff consider the financial consequences of damage to the reactor coolant pumps?

206. The staff states on page B-1:

In our evaluation, each item or sub-item covered by the Order has been reviewed for conformance with the requirements of the order.

Where existing standards remain valid, they are used as the basis for assessing such conformance.

In some areas, existing standards have been judged inadequate since the TMI-2 accident; in others, formal acceptance s tandards do not exis t.

In these cases, new acceptance criteria have been or are being developed. Where these are available, they have been used. Where new criteria are not available, judgment of the staff has formed the basis for assessing conformance with the Order, considering such factors as i comparison with other plants, degree of improve-ment over previous implementation, and expert opinion. In each case, when new criteria or standards become available, we will evaluate the items against those criteria and report our findings in a supplement to this evaluation.

Please identify each " item or sub-item" as to which " existing standards have been judged inadequate" or " formal acceptance standards do not exist."

207. Provide a list of the staff members who performed the j technical review and prepared the inputs to the S ta tu s Re po r t .

208. Provide a list of the staff management personnel who reviewed and approved the inputs to the Status Report.

, 209. With respect to both groups, of persons identified in the two previous answers, provide a statement of their educational background, training, and qualifications.

21 0 . With respect to both groups, provide their time and attendance cards for the period since the TMI-2 accident until the present time. A computer printout of this data is acceptable

4 I

if accompanied b/ an explanation of the program and activity codes.  :

, 211. Provide the staff inputs to the Status Report.

212. With reference to the staff's answer to UCS ,

interrogatory #46, does the staff agree that a break in the -

too of the pressurizer (or the pipes connected thereto) could have the same effect as an unisolated, stuck-ocen PORV? If >

the answer is "yes," identify which of the short and long-term i

measures are addressed to mitigation of that oostulated event.

UNION OF CONCERNED SCIENTISTS By:  ;

~

Ell yn R . Weiss

! SHELDON, HARMON & WEISS 1725 I Street, N.W.

Suite 506 -

Washington,.D.C. 20006 i

> (202) 833-9070  ;

DATED: February 1, 1980 k

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f UNITED STATES OF AMERICA '

NUCLEAR REGULATORY COMMISSION WED CCng,q . ,,

BEFORE THE ATOMIC SAFETY AND LICENSING BOARD

)

In the Matter of )

)

METROPOLITAN EDISON ) Docket No. 50-289 COMPANY, et al., ) (Restart)

)

(Three Mile Island ) .

Nuclear Station, Unit )

No. 1) )

)

CERTIFICATE OF SERVICE I hereby certify'that a copy of " Union of Concerned Scientists Second Set of Interrogatories to NRC Staff" was mailed postcqe prepaid first class this 1st day of February 1980 to the following

, parties:

Secretary of the Commission ATTN: Chief, Docketing and Service Section -

U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Ivan W. Smith, Esquire ,, , x Atomic Safety & Licensing Board Panels ~

U.S. Nuclear Regulatory Commission ,'-

',f Washington, D.C. 20555 '

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Dr. Walter H. Jordan U- D ,a Lc I

[F 881 W. Outer Drive 5,.'k T #'

Oak Drive, Tennessee 37830 \ ,'; A b[d ,y tf p*p

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( ,7,s Dr. Linda W. Little v r ,5 '

5000 Hermitage Drive  %'-;

Raleigh, North Carolina 27612 George F. Trowbridge, Esquire Shaw, Pittman, Potts & Trowbridge 1800 "M" Street, N.W.

Washington, D.C. 20006 James Tourtellotte, Esquire

• Office of the Executive Legal Director U.S. Nuclear Regulatory Commission Washington, D.C. 20555 s, -

• hand-delivered 2/4/80 Elly h Weiss