ML19330B295
| ML19330B295 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/28/1980 |
| From: | Swartz L NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | UNION OF CONCERNED SCIENTISTS |
| References | |
| ISSUANCES-SP, NUDOCS 8007310315 | |
| Download: ML19330B295 (42) | |
Text
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6 UNITED STATES OF AMERICA NUCLEr.R REGULATORY COMMISSION BEFORE THE ATOMIt SAFETY AND LICENSING BOARD 2
In the Matter of
)
METROPOLITAN EDIS0N COMPANY, ET AL.
Docket No. 50-289 i
(Three Mile Island Nuclear Station, Unit 1)
NRC STAFF RESPONSE TO UNION OF CONCERNED SCIENTISTS THIRD SET OF INTERR0GATORIES Pursuant to 10 C.F.R. 5 2.720 and 10 C.F.R. 5 2.744, the NRC Staff has responded to " Union of Concerned Scientists Third Set of Interrogatories to NRC Staff Based on New Information in the SER" dated July 7,1980.
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 Roons and guidance as to where the information can be found. Affidavits identifying the individuals who prepared the responses and verifying them which are not attached will be sent at a later date.
Respectfully submitted, d(.
Lucinda Low Swartz Counsel for NRC Staff Dated at Bethesda, Maryland this 28th day of July,1980 8007310315
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1.
The staff states the " licensee has committed to perform a functional test prior to restart to verify that all EFW pumps start on loss of feed-stater or loss of four reactor coolant pumps.
We will review and evaluate the test procedure and results prior to restart."
(. lines 22-24 on page Cl-1)
Will the test and/or review be completed prior to t he Restart hearings?
Will the inter-venors receive a copy of the staff evaluation upon its completion?
Resmuse 1.
A.
No. The functional test and corresponding Staff review is not expected to be completed prior to the restart hearing.
Yes. The results ~of the Staff's review of the test procedure and test results is expected to be available to all parties prior to restart.
B.
None.
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C.
None.
D.
None.
E.
The Staff does not intend to present exoert witnesses on tha subject matter covered in this Interrogatory as there is no conte.ition in tais area.
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UCS #2 liodification of AFW Control Valves.
7 The staff has requested that the " licensee evaluate the change in a.
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the failure mode of EFW control valves for possible adverse effects?
due to overfilling the steam generator."
(lines 7-9 on p. Cl-2)
In the Status Report, " Status Report on Evaluation of Licensee's Compliance with the NRC Order dated August 9, 1979," January 11, 1980, hereafter
" Status Report" (lir.as 11 '3 on p. Cl-3) the staff stated "to support this position, the licensee states that he will provide an analysis which will include the possible effects of transient hydraulic phenomena for the main steam line."' Does the failure to discuss such a proposed study in the SER now indicate the licensee no longer plans such an analysis?
If so, explain the basis for initially planning that study and why the staff now finds compliance with this part of the Order without that study.
b.
The staff states the Licensee has.provided a "best-estimate calci ation 1
specific to TMI-1, which indicates that ten minutes is available before the steam generator water level reaches the top of the shroud...
(and) that an 80 gallon air reservoir and nitrogen bottle storage system provides an additional two hours of control valve operability following loss of the instrument air, service air, anc backup air compressors."
(lines 12-17 on p. Cl-2) 1.
Provide the above cited calculation and bases for the ten minute and two hour time periods.
2.
Provide the staff analysis and explain in detail the bases for finding these time periods accurate and " acceptable."
(line 17 on p. Cl-2)
Describe the " options available to the operator to prevent overfill" c.
identified by the licensee.
(lines 18-19 on p. Cl-2)
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d.
The staff states in the SER that the licensee has identified the options available to the operator to prevent overfill, and we concur that operator action can be performed in the time available. We conclude that time and means would be available to prevent overfilling the steam generator subsequent to failed-open EFW control valves.
(lines 18-22 on p. Cl-2) 8&o, 0
1.
What is the exact' " time available for such operator action?
2.
What other fonctions would the operator be required to undertake under these circumstances and what would be the respective time demands on the operator from these additional functions.
3.
Describe in detail the specific "means" that would be available 7-to the operator to prevent overfilling the steam generators..."
(line 21 on Cl-2).
What commitment has the licensee made to upgrade the EFW system design e.
to safety grade in : the long term?
(lines 22-24 on p. Cl-2) What is the timetable for meeting that commitment?
f.
Describe the " adequate guidance on potential overfilling conditions" that the staff concludes has been incorporated in the licensee's revised procedures.
(line 28 on p. Cl-2)
Provide the specific procedures that have been included in the licensees revised procedures.
Provide the staff's method of reviewing these procedures.
g.
Provide the training material on overfilling conditions and preventives operator actions (lines 29-31 on p. Cl-2)
Describe the staff review process for these materials.
Response
2.
A.
a) Yes. The Staff is satisfied that the licensee has demonstrated ade-o quate means for prevention of steam generator overfill. The Staff is also satisfied that the stress analysis performed by the licensee and described in the Restart Report as part of the response to Question 2, Supplement 1, Part 2 which indicated that the flooded main steam lines could withstand the effects of the resulting dead-weight, internal pressure, and thermal expansion without adverse consequences is conservativ,e and therefore no further analysis is necessary.
In addition, it was noted that no adverse main steam line effects were detected from the Rancho Seco overfill incident.
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. b)
.l..The calculation which identifies the ten minute time period before the water level reaches the top of the steam generator shroud including the assumptions and basis is included as an attachment to the licensee's response to Question 2 Supplement 1, Part 2 of the Restart Report. The Staff has not requested copies of the calculation relating to the two hour air reservoir and nitrogen bottle storage system capacity as it is felt that the two hour design provides a large margin for conservatism.
2.
The Staff has reviewed the steam generator overfill calculation identified in item b-)l.' above and concurs with the conservatism of the assumptions made. These include the maximum emergency feedwater (EFW) flowrate to one steam generator and no credit for reduction in water level due to creati'on of steam (i.e. the steam generator fills with saturated water).
Further, the ten minute overfill time was verified on the B&W plant simulator.
The backup air / nitrogen storage system provides a,large enough additional time period for remote EFW control valve operability on loss of the instrument air system to permit an operator to loca11y man the control valve station if necessary. This action should take no more than 30 minutes to accomplish.
c) The operator can selectively turn off emergency feedwater pumps and/or manually isolate and throttle the EF14 system valves from the control room.
These actions are included in the licensee's response to Question 2, Supplement 1, Part 2 of the Restart Report.
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d) 1.
The operator has a minimum of 10 minutes available as previously identified in the Restart SER and discussed above.
2.
It is not possible to postulate all possible operator actions
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required at the time a potential overfill situation may occur.
The necessary actions depend on the particular accident or transient situation in progress and other unforeseen complicating events. Operator functions are covered by plant procedures.
Typically, the operator will be following a plant shutdown sequence. This involv'es verifying reactor trip and actuation of backup AC power systems (i.e. diesel generator start and loading) should offsite power be lost, and verifying the establishment of natural circulation.' The time to accomplish the initial plant shutdown sequence and verification is appoxi-l o
mately one to two minutes. Once initial shutdown is verified, the operator will monitor primary system parameters (temperature and pressure) to verify that natural circulation has been established. One of the first steps taken by the operator in
.y establishing natural circulation is to verify that the proper steam' generator level has been established.
It will take approximately 15 minutes to assure that natural circulation has been established.
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5e 3.
Tripping of the motor driven EFW pumps (EP-2A&B) is accomplished by using the panel mounted switches.
Steam supply to the turbine driven EFW pump (EP-1) can be stopped by the steam isolation valve (ts-V10A&B and MS-V13A&B) switches or the controller for the steam pressure regulating valve (MS-VS).
Emergency feedwater flow from the turbine driven pump can be isolated by the switches for the pump discharge cross tie valves (EF-V2A&B). EFW flow
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can also be throttled an d isolated by the controller for the EFW control valves (EF-V30A&B). The above actions can all be taken in the control.oom.
In addition, all the above EFW pumps and valves can be contro,11ed and operated locally.
e) The licensee's commitment for modifying the EFW system to meet safety grade requirements and the corresponding time table for completion is discussed in detail on pages C8-34 thru d8-38 of the Restart SER.
B.
1.
TMI-l Restart SER, NUREG-0680 2.
TMI-l Restart Report C.
None D.
None E.
The Staff does not intend to present expert witnesses on the subject natter covered in this Interrogatory as there is 'no contention in this area.
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2f.
The general method of reviewing these procedures was to:
(1) identify the concern (EFW valves fail open on loss of air); (2) consider the T.
events which could result in a loss of air (loss of power and loss of instrument air); and (3) verify that procedures have been revised to include operator guidance in this situation (operator actions based uponinstrumentresponse). The guidance in the EP 1202-2A, " Station Blackout with Loss of Both Diesel Generators," step 2.A.3.8.a and EP 1202-368, " Loss of Instrument Air-No Backup Air Available," step B36.2.B.5 has been deemed adequate.
As stated in our letter of July 17, 1980 to the Board, the staff will not provide copies of licensee documen'ts to UCS. These documents are available fromthe licensee.
This applies also to interrogatories 2, 5, 6, 8, 10, and lla.
9 2.
The general method of reviewing these materials was to:
(1) identify 9
the concern (EFW valves fail open on loss of air); (2) verify that lesson plans address this concern (modify system design desciption);
and (3) verify that reir'ad revised emergency procedures are included in the training (EP 12-2-2A and EP 1202-36B).
The lesson plan on the EFW system covered failure modes (itea c. pages 5 and 6), manual initiation and control (item B. pages 13,14, and 15), and overcooling considerations (item B. pages 26 and 27).
In addition, the training included a review of EP 1202-2A and EP 1202-368.
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3.
In the L u t t. s Report, the staff concluded that the EFW system piping should be considered a "high energy piping system" and upgraded accordingly.
(lines 24-26 on p. C1-10)
Explain the basis for this conclusion and explain why the staff is no longer requiring the licensee to upgrade the EFW system to a high energy piping system (lines 34-41 on p.
Cl-10 and lines 1.20 on p. Cl-11 )
What is the practical effect of this change in staff position?
In other words, what additional modifications would the staff have required had it continued to require that the EFW system be upgraded to a high energy piping system?
Response
3.
A.
The basis for reviewing the emergency feedwater system design against high energy pipe break criteria is contained in Auxiliary Systems Branch Technical Position 10-1, which is attached to Standard Review Plan (SRP)
Section 10.4.9.
High energy pipe break criteria is described in SRP Sections 3.6.1 and 3.6.2.
A Staff effort is underway to evaluate all PWR
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auxiliary and emergency feedwater systems against the SRP, including those that were not originally designed to the SRP standards because of their license dates,. inorder to detennine what requirements thould be applied c
to improve EFW system reliability.
Our specific concern on TMI-l was with the possibility of fa'ilure to isolate a postulated break in the EFW pump discharge line cross tie, and the potential damage that could occur to the ERI pumps should they operate too long at near run out conditions. We are satisfied that additional aut'omatic EFW pump protection against this postulated break is not required for the reasons indicated in the Restart SER.
In addition, I
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subsequent discussions with a pump c> pert reinforced our feeling that op'erator action could be taken to close the cross tie line motor operated isolation valves before adverse affects to the pumps would be noted.
This action can be taken to assure EFW system operation even in the event of single failure of one of the cross tie line valves to close.
It should also be noted (refer to the Restart Report, Question 4, Supple-ment 1, Part 3) that the licensee is installing cavitating venturis in the EFW pump discharge lines in the vicinity of the control valves (EF-V30A&B) to further a~ssure against possible EFW pump runout in the unlikely event of a system pipe break downstream of the venturi location.
B.
1.
Standard Review Plan Section 10.4.9 2.
TMI-1 Restart SER, NUREG-0680 3.
TMI-l Restart Status Report 4.
TMI-l Restart Report C.
None D.
The Auxiliary Systems Branch is continuing their review of all operating PWR auxiliary.feedwater systems as part of the TMI Action Plan, NUREG-0611, Item II.E.1.1.
E.
The Staff does not intend to present expert witnesses on the subject matter covered in this Interrogatory as there is no contention in this' area.
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4.
Explain the discrepancy between the statement that "the EFW pump automatic initiation signals are independent of the intergrated control system" (emphasis supplied) (lines 15 and 16 on p.
C8-35) 7.nd "the licensee has committed to provide 2
automatic initiation of the EFW system pumps." (emphasis supplied) (lines 26 and 27 on p.
cl-11).
When will such auto-matic initiation be provided?
Response
4.
A.
The statements referred to are not in disagreement. The licensee's commitment was for a two foid program of short and long term modifica-tions for EFW pump automatic initiation as described on pages C8-34 thru C8-38 of the Restart SER. Prior to restart, phe system will be modified such that all EFW pumps will start automatically on loss of both feed-water pumps or loss of four reactor coolant pumps independent of the
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integrated control system.
B.
TMI-l Restart SER, NUREG-0680 C.
None D.
None E.
The Staff does not intend to present expert witnesses on the subject matter covered in this Interrogatory as there is no contention in this area.
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a UCS #5 In the Status Report (lines 14 and 15 on p. C2-4), the licensee had
" committed to revise the TMI-l emergency procedures to provide the operator T
with guidance ont he recognition of void formation and inadequate core
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cooling." The staff stated in lines 29-30 on p. C2-4 of the Status Report that full compliance with this part of the Order would require the licensee to revise emergency procedures to provide the operators with additional guidance on " operator action required to enhance core cooling in the event such voids are formed."
(Status Report, lines 12-13 on p. C2-4)
Explain why in the SER on page C2-4 all references to this additional guidance on operator action to enhance core cooling in the event of large voids in the coolant system have been omitted.
Was such additional guidance ever provided? If so, exactly what was provided and when? Provide UCS with a copy of such guidance.
Response
Additional guidance was included in the emergency procedures subsequent to the issuance of the Status Report.
The guidance is contained in EP 1102-16, "RCS flatural, Circulation Cooling," EP 1202-6B, "Small Break LOCA," and EP1202-39, " Inadequate Core Cooling." These procedures also address the concerns of question 6 below.
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USC Interre;atorv 6 (Third Set)
, Explain exnctly what additional cperator ruidaico cs been 6.
by the licensee and approved by the staff rs cufficient te "cc.zr t'.c possiblity of RCS void formation thct.ould interrupt circulntien U.cw."
(lines 39-40, page C2-9).
When was such guid*ance provided:
Precide UCS with a copy of such guidance.
Explain cxactly how this gu' lance 7
addresses the problem of RCS void formation thct ceuld interrupt natural
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circulation.
Ec s_pgn s e_-
Fellowing a loss of forced ficw through the reactor coolant loops, natural circu-lation would be required to rerove heat from the core by anans of the steam generators.
If.all feedwater were lost, the eucretor is ins tructed to initiate HPI which will cause the FORV and/or safety valves to open
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ari provide cooling to the cerc.
These instructions are contained in the TMI-l procedures for Natural circulation (0P1102-16).
Even if feedwater were available, voids might (orm in the primary system (a) as the result of a LOCA, or (b) shrinkage of the primary system as the result of some transient or accident producing ever cooli.g by the stec generatets.
a) LOCA Analyses by the B&W and the NRC have indicated that naturel circulation has a significant effect only fer small brccks of.01 f t2 or less.
The review by the NRC staf f of the small LOCA procedures (EP 1202-6E) is described in the NEC SEE for TMI-l res art.
b)
Non-LOCA Transients or Accidents Analyses by Babcock & L'ilcox (Ecf 1) and the Les Alancs Scientific Laboratory (Ref 2) do not predict less cf naturcl circulatien f rc= non-LOCA cvents for which feodwater is availabic, including cenditf ens for which veids arc criculated to be forced in the primary systet.
In the event that netural f rculatien were blocked by sene eans that was not predicted by the analyses, i
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. this condition would be indicated by a less of the subcooled.n rgin in the Princry coolant system and the operator is re ferred to procedures for in-adequore core cooling (EP 1202-39). These procedures instruct.the operator -
to initiate HPI flow (Feed and bleed).
This action will provide cooling to :.
the reactor core.
References:
1)
B&W Report, "t.nalysis Su:. mary is Support of cn Early FC Pump Trip."
2)' Letter from C. J. E. Willcutt, LASL, to W. L. Jensen, NRC, "TMI-2 Severe Overcooling Transient with Upper Head Connectica,"
June 25, 1980.
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UCS #7 Will the analyses to " provide guidance as to the expanded systems response to anticipatory filling of the OTSG" be completed and reviewed T
by the staff prior to t he restart hearing?
(lines 2-3, page C2-10) Will
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a copy of the staff's review be provided to t he intervenors upon its completion?
Response
The staff expects to issue a supplement to the SER prior to the hearing.
If the analyses is submit'ted by the licensee as expected in sufficient time to permit completion of our review prior to preparation of that supplement, thee results of our review will be included.
If not, our review will be reported in a subsequent supplement.
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UCS #8 Identify the " additional operator guidance on overriding automatic safety features actuation" that has been developed by the licensee.
2 (line 3 on page C2-ll) Provide a copy.
Explain in detail the staff basis for now finding compliance with this part of the order.
Response
Additional operator guidance on overriding automatic safety features actuation was required to identify operator res onse if ESFAS were r
inadvertently initiated. This guidance was provided in step 3.3.5 of j
OP 1105-3. The licensee has now pravided-operator guidance on override for actual occurrences (see discussion in Order item ld) and inadvertent actuations.
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a n._tgry_,f_(~.: rd 5?t) m lic(nsee hm con :itted to change the I'EV sctpoint to 2'50 psis.
('ine 31, p. C2-ll) Decs the enission in tbc SIR of the sentence:
"We will iequire the licensec to report completion af this change prict to startup" (Status Repert, lines 30-31 on pcge C2-ll) i:rdicate that the licenece need not change the PORV setpoint prior to startup?
If not, c:: plain the basis for the staff's new position. Fcen.eill the FORV setpoint be changod?
Explain why the licensee is still fcund in ompliance with this part of th[e Order.
Re onse:
- he change refered to between our ststus report of January 11, 1980 and the SER for TMI-l restart dated June, 1930, is editorial. The NRC will require thet the PORV setpoint be changed to 2450 peig before startup. Note that the first paragraph of the discussien for item 3 ef IE Bulletin 79-05B on Page C2-ll.
of the SER for TM1-1 restart states " Verification of thecc setpoint changes (2430 prig PORV) will be nade by the staff prior to startup." There is no change in,the staf f's position on this' matter, s
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UCS #10 In the Status Report (lines 1 and 2 on page C2-12), the staff stated it was reviewing "the procedures under which manual reactor trip on low T
pressurizer level and main steam line isolation valve closure are appropriate."
Describe in detail the staff review and explain the basis for finding that the licensee's revised procedures now include " appropriate guidance for a manual reactor trip under such conditions" (lines 12-15 on page C2-12).
Provide the licensce'r revised procedures to UCS.
Response
The staff reviewed Procedure AB 1203-15. " Loss of Makeup."
It was determined that the loss of makeup capability, with resultant reductions i
in pressurizer level, could 16ad to the loss of RCS pressure control if the level dropped below the low level heater cutoff setpoint.
Therefore, a manual reactor trip under these circumsances was deemed appropriate. Procedure AB 1203-15 includes this requirement.
Since MSIV closure leads to a loss of heat sink and will lead to an increase in RCS pressure the staff requires a manual reactor trip to reduce the severity of this transient.
Procedure AB 1203-42,
" Inadvertent Closure of a MSIV" was issued to require a manual reactor trip upon MSIV closure.
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11.
Regarding tne u vu c....
a ci qual 112 cation of the reactor protection system (RPS) equipment, the staff maintains:
the licensee has stated that the RPS modules to be used have been qualified for use in B&W safety systems.
Also, the pressure switch sensors will be equivalent to switches used 7
in other plant safety applications.
Existing RPS power supplies, flux signals, interlock circuits, and indicators will be used as needed by the added equipment.
The require-ments for the RPS (e.g.,
cooling, power, seismic, environmental) will be the same for the system with anticipatory trips.
We have not, however, received the oocumenta-tion necessary for independent evaluation of these aspects of the design.
Upon receipt and review of this documentation, we will provide our findings in a supplement to this report.
(Status Report, lines 30-38 on p.
C2-13) a.
Has the " documentation" to which the Status Report referred been received and reviewed by the staff?
If so, provide UCS with such docu-mentation.
If not, when will it be?
b.
Describe in detail the paramete:s and the methods used for such qualification.
c.
Has the staff performed an independent review of the qualification of the RPS modules?
If so, describe that review in detail.
d.
The staff states in the SER that final design documentation for the additional equip-ment includes " electrical schematic diagrams, 6
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Justification of trip signals, and location layout drawings ? (lines 3 and 4 on page'C2-14).
However*,
in the Status Report, the staff stated it would also require " fi nal logic diagrams" and " piping and instrumentation diagrams. "
(Status Report, lines 3 and 4 on p. C2-14)
Why are these two sets of diagrams no longer required in order to find compliance with this part of the Order?
How has the review compensated for the lack of these diagrams?
Response
A.
We are in receipt of design drawings which show that the new components are functionally the same as and have like mqdel numbers to that of the existing and approved TMI-1 design. We also have statements by the licensee
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(in the Restart Report) that these components are the same as those used -
in the original TMI-1 RPS.
By utilizing additional RPS modules and pressure switches that are generically the same as devices already reviewed and approved by the staff for use in similar applications within the safety system, the licensee obviated the need for detailed staff review of the entire modifi-cation.
It has been and continues to be the staff's practice that items which have already been found acceptabl-by previous staff review are not required to be rereviewed for the same or similar applications.
There fore, submittal of such items as the parameters and methods of qualification were not requested.
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The staff has not performed an independent review of the qualification of the RPS modules at this time, because, as discussed above, these modules are part of the already reviewed and approved original TMI-l RPS.
The list of items needed for review delineated in the Status Report reflects staff requirements for a new RPS design that has not been previously r<3vi ewed.
This list was incorporated in the staff's generic evaluation of the preliminary RPS design modification for the B&W operating reactors.
The generic evaluation did not include THI-l and therefore for consistency, the list was also included in the preliminary TMI-1 evaluation.
s Upon reviewing the THI-l proposed final des.ign modification, it became clear that many of the informational items required for a new unreviewed RPS design concept would not be necessary.
As the actual design is documented in the Restart Report and the Restart SER (NUREG-0680), I will not reiterate the detailed design here The logic used by the additional RPS trips is the very logic (i.e. the same hardware) that all the or.iginal
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RPS trips feed into.
Therefore, additional 1.ogic drawings on an already reviewed and spproved design were not necessary.
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The RPS modification consists of new sensors ccnnected to new RPS modules via new cabling and through new buffer devices.
The buffer devices and modules are located within the existing RPS cabinets.
(See Figure 2.1-1 in the TMI-l Restart Report.) Piping and instrumentation drawings were not required as there is no real instrumentation involved with the piping system.
Only the sensors are located externally to the RPS cabinets.
Sufficient drawi.ngs h* ave been supplied by the licensee to enable verification of the implementation of the design.
In summary, the RPS Modification was a simple changa to an already reviewed and approved design which'was effec.ted by using more of the same'componen,ts.
In such a situation, the acceptability of the modification is based upon the following:
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- 1) Does it meet safety system criteria,
- 2) will it perform its intended function, and _
- 3) does it not degrade or interfere with the existing portion of the design?
Based upon our review, we have concluded that the above three questions are answered in the affirm tive.
In addition, we have required submittal of an RPS check-out proced:tre (for our review and approval) which demonstrates both the operability of the new trips and the continued operability of the. previous. RPS trips.
- p. C2-14) i B.
TMI-l FSAR TMI-l Restart Report, TMI-l Restart SER (NUREG-0680)
C.
None D.
None E.
The staff ddes not intend to present expert witnesses on the subject
-T matter covered in this interrogatory as there is no contention in this area.
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UCS #12 The Status Report contains the following quote:
The licensee states that the Teedwater pumps differential pressure switches will remain control grade (they are installed in a non-seismic building) and will be connected to the safety grade initiation circuits through buffer relays.
We have reviewed this design and require additional justifica-tion for not meeting safety grade requirements for the differential (lines 1-5 on p. C8-35) pressure switches and circuitry.
The SER discussion of " Automatic Initiation of the AFS" omits any reference to these switches in their circuitry.
(p. C8-35-38)
- a. 4Ias " additional justification" provided for not meeting safety grade requirements?
b.
Will the differential pressure switches and circuitry be upgraded to safety grade?
c.
If the answer to b).is "no," explain in detail the basis for the staff conclusion that the licensee will meet the requirements of NUREG-0578, Item 2.1.7.a.
Response
t 12.
A.
a) Page C8-35 of the Restart SER describes the differential pressure switches used to provide automatic EFW pump initiation on loss of main feedwater with basically the same description as was used in the Restart Status Report. Additional justification for accepting the design as presented is indicated in the SER discussion and in item b).
B.
Till-l Restart SER, NUREG-0680 C.
None D.
None E.
The Staff does not intend to present expert witn_sses on the subject matter covered in this Interrogatory as there is no contention in this area.
12.b and c A.
Yes, the differential pressure switch circuitry will be safety grade. As noted on page C8-35 of the Restart SER the staff's only concern over the I
safety grade aspect of the design has been that the pressure switches are located in a non-seismic Category I structure.
In order to acceptably interface these pressure switches with the EFW system, some sort of isolation device was required. As noted in the Status Report, the licensee stated that safety grade buffer' relays would be used for this purpose. The use of safety grade buffer relays in an acceptable method of interfacing the non-seismic portion of the circuit with the seismic portion.
However, the staff withheld judgement on the acceptability of the des.ign pending receipt of the qualification dopumentation of the buffer rel' ys.
a Subsequently, the liensee documented within the Restart Report, that the relays to be used were the same as those used as part cf the already reviewed and approved TMI Unit 2 Engineered Safety Features Actuation System in an identical application i.e. as bu'ffer relays. With this documentation in-hand the staff was able to conclude that the circuit is safety grade e,d acceptable. Based upon the "yes" answer to part b, part c does not require an answer.
B.
TMI-1 Restart Report (Med Ed/GPU)
TMI-2 FSAR (Met Ed/GPU)
TMI-1 Status Report (NRC)
TMI-1 Restart SER (NUREG-0680)
C.
None D.
None E.
The staff does not intend to present expert witnesses on the subject matter covered in this interrogatory as there is no contention in this area.
o O
e
3:r UCS #13 In the Status Report, the staff noted that the licensee committed to complete the safety grade modification for the automatic initiation of the auxiliary feedwater system (AFWS) by June 1, 1980 (lines 33 and 34 on p.C8-34). Although the staff now estimates that this safety grade modification will not be ompleted until mid-1981, the staff still concludes that the. licensee has met this requirement (line 10 on p. C8-37).
a.
Where specifically in Amendment 12 is the licensee's proposed schedule for installation of the safety grade design?
b.
How many level indicators are there per steam generator currently?
What are their ranges?
- c.,The licensee's " Report in, Response to the NRC Staff Recommended Requirements for Restart of TMI-1" (Restart Report) indicates that two steam generator level instruments have been modified to avoid flooding (lines 33-37 on p. 2.1-23; Am.13).
In the SER, the staff refers to three level indicators (line 28 on p. C8-37). Does this discrepancy mean that only two of three level indicators have been modified to avoid f,looding?
If not, explaid what it means.
d.
The SER includes the following quotation:
To assure that the operator can properly control steam generator level from the new manual EFW control valve station in the event of
~
a single failure in the ICS or non-nuclear instrumentation (NNI),
the licensee has committed to modify the existing steam generator level indication in the control room prior to restart. The modi-fication will consist of providing one of the three existing steam
~
generator level instruments with a redundant battery-backed power supply independent of ICS and the normal NNI power source.
- Thus, two level -indicators will be powered from separate class IE bu-ses, thereby meeting the single failure criterion. We therefore find the proposed modification acceptable (lines 24-32 on p. C8-37).
1.
How many level instruments per steam generator will be provided with a redundant battery-backed power supply?
2.
If only one level indicator per steam generator will be so modified, explain how the proposed level instrument arrangement will meet the single-failure criterion particularly for a steam or feedwater line break.
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3.
Explain how the proposed arrangement meets the requirements of IEEE-279 regarding the development of anomalous conditions con-fusing to the operator.
e.
What criteria will be used by the staff in its review of the detailed design for the level indicator arrangement in order to determine the adequacy of the modification prior to restart?
Response
13.
A.
a) The proposed schedule -for ' installation of the safety grade EFW system modifications is shown in the response to Question 4, Supple-ment 1, Part 3 of the Restart Report.
a b)
In clarification Of the Restart SER (page C8-37), there are currently three types or ranges of level indicators per steam generator', two of which have a redundant counterpart for a total of five level indi-cators per steam generator. The operator, howev'er, is able to read only three gauges at a time and may manually select to read one of the redundant indicators as he sees fit. The level indicator arrangement and ranges are as follows:
1.
Two (2) startup range indicators with a range from 0" to 250" 4
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2.
Two (2) operating range indicatois with a range from 102" to 394" 3.
One (1) full range indicator with a range from 6" to 606" c) The Staff has been informed verbally by the licensee that three of five steam generator level transmitters per steam generator (one instrument for each of the three ranges described in b) above), rather than two as indicated in the Restart Report, have been raised to avoid flooding in the reactor building.
Thus, at least one full set of steam generator level instrumentation per steam generator is assured of operation in the event of flooding in the reactor building.
The licensee will amend the Restart Report to reflect this chan'ge.
d) 1.
As stated in the Restart SER, all the steam generator level indi-cators are po,wered from the same battery backed 1E source. One of the five indicators per steam generator will be modifi.ed to receive power from a separate redundant battery backed IE source.'
T herefore, one level instrument is assured of operation in the event of a single failure in the power supply to the other four.
A main steam or feedwater line break does not affect this capatility.
2.
Refer to d)l. above.
3.
Redundant startup and operating range instruments per steam generator will provide the operator with the capability of cross checking the level reading at his discretion. The operator may a'sb perforin a channel check on a particular level instru-c ment circuit to verify that it is functioning properly if he
L5.i suspects a problem with the indicated reading.
As an additional verification the operator will verify that primary system parameters (temperature and pressure) are within acceptable limits.
B.
1.
TMI-1 Restart Report 2.
TMI-1 Restart SER, NUREG-0680
~
C.
None D.
None.
E.
The Staff does not intend to.pr'esent expert witnesses on the subject matter covered in this Interrogatory as there is no contention in this area.
13.e A.
The criteria used to evaluate the automatic initiation of the EFW system and the new m nual EFW control station and our evaluation of same are documented in the TMI-1 Restart CER (pp. CP-34 thru C8-38).
The only item which has not been fully closed out te the detailed design modification for stea enerator level indication. The requirements, as detailed in the SER (p. C8-371, are that the redundant level indicator recieve Class 1E power from a redundant
~~
power source (other than the NNI power source which supplies the remaining four indicato~rs) and that this source'not be subject to failure of the NNI
_c;'
power source.
,These requirements are the acceptance criteria we shall apply when reviewing the details of this aspect of the design.
To reiterate, the remainder of the design has been found acceptable as documented in the Restart SER.
(See also the staff response to Item 13.d.1.)
B.
THI-l Restart Report (Met Ed/GPU)
TMI-1 Restart SER (NUREG-0680)
C.
None
...c
. D.
None E.
The s'taff does not intend to present expert witnesses on the subject matter covered in this interrogatory as there is no contention in this area.
1 UCS #14 With respect to the proposed new instrumentation for the detection of inadequatIe core cooling, the staff has concluded that the licensee is not in compliance
~
with this part of the Order because "no additional instrumentation has yet been committed to, scheduled, or conceptually addressed..." (lines 22-24 on p. C8-21).
Explain precisely what the staff will require on this issue prior to restart and for the long-term modification.
Response
The staff requirements fo,r new instrumentation to detect inadequate core cooling are presented in NUREG-0578 (1) and supporting documents (2, 3) and are reiterated in the TMI-l Restart.(SER (4).
These documents state that the licensee shall provide a descrition of any additional instrumentation or controls proposed for the plant to supplement those devices cited in the preceding section of the requirement (2.1.3.b) which give an unambiguous, easy-to-interpret indicat' ion of inadequate core cooling.
Metropolitan Edison and Babcock & Wilcox have proposed the use of existing instruments (that is, no additional instrumentation) to provide an indication of inadequate core cooling. The basis for this proposal is that there is no need for additional instrumentation, particularly a reactor vessel water level measurement, to determine inadequate core cooling.
We have reviewed the justificatio'n for no additional instrumentation and have found it unacceptable.
Prior to restart of TMI-1, we will require a commitment on the part of the licensee to provide "an unambiguous, easy-to-interpret indication of inadequate core cooling." We will require a conceptual description of, 9
e,
and implementation schedule for, the instrumentb) proposed to accomplish this task. We will not accept a discussion of the desirability of such instruments.
In the long term, we will require design and testing details, t
implementation, and verification of installation of these instruments.
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15.
A.
a) Yes. The Staff agreed with the licensee concerning the satisfactory seismic and environmental qualification of the emergency feedwater flow devices.
However,.he Staff has been made aware of LER 80-012/0lT-0 dated
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July 11,1980, which describes certain nonconservatisms in the original Intermediate Building main steam line rupture analysis. The LER is a result of the licensee's reevaluation of environmental qualification of safety related equipment in response to IE Bulletin 79-018. As a result of this information, the environmental qualification of the transducer portion of the emergency feedwater flow devices which is located in the Intermediate Building will be rereviewed as part of the restart effort. Our concur'rence with the environmental qualifica-tion of the flow display computer is not affected by this LER as it is located in an a'rea not affected by the steam line break analysis as indicated in the Restart SER page C8-39.
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b) The Staff's original review detemined that the expected design ambient Intermediate Building and diesel generator building environ-ments are bounded by the qualification ranges for the sonic flow device transducer and computer respectively.
This conclusion will be reevaluated for the Intemediate Building as discussed in item a) 1 above.
The building ambient environments are as follows:
1.
Intermediate Building:
50*F to 323' maximum for a matter of seconds; <120*F after one half hour
, 2.
Diesel Generator Building:
50*F to 104*F In addition, the expected process line temperature range for the EPf suDply lines in the.Intemediate B'uilding to which the transducers may be exposed is from 40*F to 100*F.
o c) No. The staff is not aware of any tests that have been perfomed on the sonic flow devices. See d) below.
d) The Staff has not been informed as yet as to the specifc nature of the vendor documentation the licensee plans to submit. To fulfill our requirements, however, we require submittal of actual certified test results as part of the Restart Report.
B.
1.
TMI-1 Restart Report 2.
THI-1 LER 80-012/01T-0 dated July 11, 1980 i
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$s.
.u.
C.
None A reevaluation of the Intennediate Building environment will be perfonned D.
z by Jared Wermiel of the Staff, and a rereview of EFW system equipment qualification will be performed by the Equipme,nt Qualification Branch.
The Staff does not intend to present expert witnesses on the subject E.
matter covered in this Interrogatory as there is no contention in this area.
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3i U.ITED STATES OF AMERICA NUCLEAR REGULATORY. COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
METROPOLITAN EDIS0N COMPANY, et al.
)
Docket No. 50-289
)
(Three Mile Island, Unit 1)
)
AFFIDAVIT OF ROBERT'G. FITZPATRICK I, Robert G. Fitzpatrick, bei.ng duly sworn, do depose and state:
1.
I am a Senior member of the Power Systems Branch in the Division of Systems Integration, Office of Nuclear Reactor R.egulation of the Uni.ted States Nuclear Regulatory Commission.
I am responsible for the electrical aspec,ts of the safety review of ass.igned nuclear power plants, including Three Mile Isla'nd, Unit 1 Restart Pr.ogram.
2.
The answers to UCS Round Three Interrogatories 11,12 (b&c) and 13 (e) were prepared by me.
I certify that the answers given are true and accurate to the best of my knowledge.
s W
- A R6be'rt G. Fitzpatfic Subscribed and sworn to befora me this 25th ay of d
J_uly,.1980 i
hu L %
' Notary'Public
/
j (4y Commission expires: July 1, 1982
UNITED STATES OF ASSRICA NUCLFAR REGUIAIDRY COtt1ISSION BEFDRE THE A704IC SAFEIY AND LICENSING BOARD In the Matter of
)
METROPOLITAN EDISON RIFANY, et al.
Docket No. 50-289 (7hree Mile Island, Unit 1)
)
e*
AFpIDAVIT OF BRUCE BOGER I, Bruce Boger, being duly sworn, do depose and state:
1.
I am sployed by the Operator Licensing Branch, Office of Ibclear Reactor Regulation of the United States Nuclear Regulatory _Conmission.
2.
'Ihe answers to UCS Interrogatories 2f, 2g, 5, 8 and 10 were prepared in part or in whole by me.
I certify that the answers given are true and accurate to the best of my knowledge.
- Y Bruce A. Boger
<f Subscrib.I and sworn to before me this 18th day of July 1980 W
d icgt A u m r. 1 m
- --s,os r
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
}
In the Matter of METROPOLITAN EDISON COMPANY, ET AL.
Docket No. 50-289 (Three Mile Island Nuclear Station, Unit No. 1)
AFFIDAVIT OF HARLEY SILVER I, Harley Silver, being duly sworn, do depose and state:
1.
I am employed by the Operating Reactors Branch #4, Division of Licensing, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Comnission.
~
2.
I authored or contributed to the responses to UCS Interrogatories 2f, 2g, 5, 7, 8,10 and 14.
I certify that they are true and correct to the best of my knowledge.
~::
l ey 5 1 Subscribed and sworn to before me this 25th day of July,1980.
42 t-1)
Nota ublic
/
My Commission Expires:
July 1, 1982
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~
UNITED STATES OF AfiERICA NUCLEAR REGULATORY C0!i'11SSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
~
In the Matter of liETROPOLITAN EDIS0N C0!!PANY, et al.
Docket No. 50-289 (Three Mile Island, Unit 1)
AFFIDAVIT OF JARED S. WERMIEL I, Jared S. Wermiel, being duly sworn., do depose and state:
1.
I am a member of the Auxiliary Systems Branch in the Division of Syst' ems Integration, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
I am responsible for the auxiliacy systems aspects of the safety review of assigned nuclear power plants, including Three Mile Island, Unit 1 Restart Program.
2.
The answers to Union of Concerned Scientists Third Set of Interrogatories 1, 2a, b,'c, d, e, 3, 4, 12a, 13a, b, c, d and 15 were prepared by me.
I O
certify that the answers given are true and accurate to the best of my knowledge.
NN.Y Jafed S. Wermiel S.:bscribed ar.d sworn to before me this 23rd day of
' July 1
M
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y,980 g
~i%sNL( thoddk Nr tary P6blic
/
Fy Commission expires:
July 1, 1982
i UNITED STATES OF AMERICA TR' CLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSIllG BOARD In the Matter of METROPOLITAN EDIS0N COMPANY, ET AL.
Docket No. 50-289 (Three Mile Island Nuclear Station, Unit flo.1)
~
AFFIDAVIT OF JOHN C. V0GLEUEJE I, John C. Voglewede, being duly sworn, do depose and state:
1.
I am employed by the Core Performance Branch, Division of Licensing,s Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
~
2.
I authored or contributed to the responses to UCS Interrogatory 14.
I certify that the. answer is true and correct to the best of my knowledge.
v John C. Voglewedet.3 Subscribed and sworn to before me this;gE day of July,1980
?dh lG hota'ry PubFic ' g My Commission Expires:
/ /i
/
O w
pi UNITED STATES OF AMERICA NUCLEAR REGULATORY C0"JilSSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD _
4 In the Matter of METROPOLITAN EDISON COMPANY, Docket No. S0-289 ET AL.
(Three Mile Island, Unit 1)
CERTIFICATE OF SERVICE I hereby certify that copies of NRC STAFF RESPONSE TO UNION.0F CONCERNED SCIENTISTS THIRD SET OF INTERROGATORIES dated July.28, 1980, in the above-captioned proceeding, have bem hand-delivered to Counsel for UCS and to the Licensing Board offices this 28th day of July,1980.
ucinda Low Swartz Counsel for NRC Staff l
e i
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of f
METROPOLITAN EDIS0N COMPANY, Docket No. 50-289 ET AL.
(Ihree Mile Island, Unit 1)
[
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF RESPONSE TO UNION OF CONCERNED SCIENTISTS THIRD SET OF INTERR0GATORIES," dated July 28, 1980, in the above-captioned proceeding, have been served on the following, by deposit in the United States mail, first class, or, as indicated by an asterisk through deposit in the Nuclear Regulatory Comission's internal mail system, this 28th day of July, 1980:
i -
- Ivan W. Smith, Esq.
Mr. Steven C. Sholly Atomic Safety & Licensing L,ard Panel 304. South Market Street U.S. ?!uclear Regulatory Commission Meclianicsburg, Pennsylvania 17055 Washington, D.
C.
20555 Mr. Thomas Gerusky Dr. Walter H. Jordan Bureau of Radiation Protection 881 W. Outer Drive Dept. of Environmental Resources Oak Ridge Tennessee 37830 P.O. Box 2063 Dr. Linda W. Little 5000 Hermitage Drive Mr. Marvin I. Lewis Raleigh, North Carolina 27612 6504 Bradford Terrace Philadelphia, Pennsylvania 19149 George F. Trowbridge, Esq.
Shaw, Pittman, Potts & Trowbridge Metropolitan Edison Company 1800 M Street, N.W.
Attn: J.G. Herbein, Vice President Washington, D. C.
20006 P.O. Box 542 Reading, Pennsylvania 19603 Karin W. Carter, Esq.
505 Executive House Ms. Jane Lee P.O. Box 2357 R.D. 3; Box 3521 Harrisburg, Pennsylvania 17120 Etters, Pennsylvania 17319 j
Honorable Mark Cohen Walter W. Cohen, Consumer Advocate 512 D-3 Main Capital Building Department of Justice Harrisburg, Pennsylvania 17120 Strawberry Square,14th Floor Harrisburg, Pennsylvania 17127
?.
M Allen R. Carter, Chairman John Levin, Esq.
Joint Legislative Committee on Energy Pennsylvania Public Utilities Comm.
Post Office Box 142 Box 3265 Suite 513 Harrisburg, Pennsylvania 17120 Senate Gressette Building Columbia, South Carolina 29202 Jordan D. Cunningham, Esq.
Fox, Farr and Cunningham Robert Q. Pollard 2320 North 2nd Street 609 Montpelier Street Harrisburg, Pennsylvania 17110 Baltimore, Maryland 21218 Theodore A. Adler, Esq.
Chauncey Kepford WID0FF REAGER SELK0WITZ & ADLER Judith H. Johnsrud Post Office Box 1547 Environmental Coalition on Nuclear Power Harrisburg, Pennsylvania 17105 433 Vrlando Avenue State College, Pennsylvania 16801 Ms. Ellyn R. Weiss Harmon & Weiss Ms. Frieda Berryhill, Chairman 1725 I Street, N.W.
Coalition for Nuclear Power Plant Suite 506 Postponement Washington, D.C.
20006 2610 Grendon Drive Wilmington, Delaware 19808 Ms. Marjorie M. Aarnodt R.D. #5 Holly S. Keck Coatesville, Pennsylvania 19320 Anti-Nuclear Group Representing York 245 W. Philadelphia Street York, Pennsylvania 17404
- Atomic Safety and Licensing Appeal Board 4
7 U.S. Nuclear Regulatory Commission ggg Washington, D.C.
20555 Counsel for NRC Staff
- Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555
- Secretary U.S. Nuclear Regulatory Commission ATTN: Chief, Docketing & Service Br.
Washington, D.C.
20555