Text

Response to Aslab 820714 Order,Consisting of Status Rept on Restart Mods & Description of Short & long-term Emergency Feedwater Mods,Per NUREG-0737,Item II.E.1.1.Affidavit & Certificate of Svc Encl
	ML20062H703
Person / Time
Site:	Crane
Issue date:	08/12/1982
From:	Baxter T; METROPOLITAN EDISON CO., SHAW, PITTMAN, POTTS & TROWBRIDGE
To:	;
References
	RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.1, TASK-TM NUDOCS 8208160139
	Download: ML20062H703 (35)
	v • d • e

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' Ch!kO SECRETARY COCs ET.4C & SERVICE 3 RANCH UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING APPEAL BOARD In the Natter of

)

METROPOLITAN EDISON COMPANY

)-

Docket No. 50-289

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(Restart)

(Three Mile Island Nuclear

)

Station, Unit No. 1)

)

LICENSEE'S RESPONSE TO THE ATOMIC SAFETY AND LICENSING APPEAL BOARD'S i

ORDER OF JULY 14, 1982 i

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8208160139 82081205000289 PDR ADOCK PDR DSd3 i

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United States of America Nuclear Regulatory Commission BEFORE THE ATOMIC SAFETi AND LICENSING APPEAL B'OARD~

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In the Matter of

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METROPOLITAN EDISON COMPANY Docke't'No.5-35

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-(Resta rt')

(Three Mile Island Nuclear

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Station, Unit No.1)

AFFIDAVIT OF ROBERT W. KEATEN County of Morris

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State of New Jersey

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ROBERTW.KEATEN,beingdulyswornaccording'tolaw,dehoses and states as follows:

1.

I am Directorof the Systems Engineering Department of GPU Nuclear Corporation and have presented testimony before the 6

Atomic Safety and. Licensing Board in.this proceeding on several l

occasions.

A statement of my professional qualifications is set forth in the evidentiary record of this proceeding following Tr. 4588.

2.

The information provided in Licensee's responses, dated

/'

August 12, 1982, to the Atomic Safety and Licensing Appeal Board's questions contained in its July 14, 1982 Order was prepared by me or under my supervision by employces of GPU Huclear Corporation and is true and accurate to the best of my knowledge and belief.

Yf ROBERT W. KEATEU Subscribed to -and sworn before ne thin

/o7il day of August, 1982.

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Question I.

Update of Restart Requirements The Appeal Board requested that Licensee provide a report on the status of certain of the restart modifications listed in Appendix A to the Board's July 14, 1982 Order.

The attached chart presents the current status of these modifica-tions.

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UPDATE OF RESTART REQUIREMENTS ESTIMATED PERCENT COMPLETION ITEM COMPLETE DATE NOTE Short Term Order Iten 1 EFW Reliability b!

la-3 Auto EFW Inad to Diesels 100%

la-4 IMJ Technical Specification 100%

TSCR 103 (5/18/81), Rev.1 (11/13/81), Rev. 2 (5/20/82) pending NTC approval.

la Mditional Itens 1 CwST Invel Alann 95%

9/1/82 tkxxl to replace level gauge.

6 EFW Initiation Independent of AC (UFSG IcVel Innlication independent of ICS) 100% 2/

7 IMI Operability in Steam Envirorrnent 10%

Cycle 6 Iong Tenn - UW Safety-Grade Startup Modification 8 Cross-Tie Break 100%

Weld inspection. Results to be sent to NRC.

Order Itan 2 IE Bulletins 79-05B-3 PORV Set Point 100%

79-05B-5 Anticipatory Reactor Trip (Safety Grade) 100%

ESTIMATED PERCENT COMPLETION ITEM COMPLETE DATE 110TE Order Item 4 Separation of 'IMI-l & 2 la)

Liquid Radwaste 100%

(b)

Gaseous Radwaste 100% -!

(c)

Solid Radwaste 100%

(d)

Sanpling Syston 100%

Order Itan 8 Iessons Icarned - Stort Term 2.1.1 anergency Power Supply

- Pressurizer lieaters 100% /

2 2.1.3a Valve Position Indication 99%

9/1/82 2.1.3b Inadequate Core Cooling

- Existing Instrumentation & Saturation Meter 100%

2.1.4 Contairinent Isolation 80%

11/1/82 2.1.5c Install Redanbiner 99%

9/1/82 2.1.8c Iodine Instrunentation 60%

10/1/82 Innq Term (LT) (NURIX;-0737 numbers)

MM 9/1/82 Ilf-l (II.K.2.9) ICS FMEA Modifications 98% '

ESTIMATED PERCENT COMPLETION ITEM COMPLETE DATE NOTE L!1'3 Lessons Icarnod Category D fran NURH3-0578 75% /.

10/1/82 l

Modifica-Environnental qualification of 2.1.3b (II.P. 2. 3)

IT Instnzrentation tim Env. Qual.

this nodification will be accom-10%

plished in accordance with tle Final Rule.

-Backup Incore Thenrocouples (safety grade)

(see note) 2.1.5a (II.E.4.1)

Dedicated 11 Penetrations 2

- Install 100%

2.1.6b (II.B. 2)

Plant Shielding 100%

MT: mtor Control Center MT DIIRS cle 6 OllRS: Decay Heat Rmoval System

- Plant Modifications 30%

Startup 2.1.7a (II. E.1. 2)

EIM Auto Initiation

- Safety Grade 100%

2.1.7b (II.E.1.2)

EIM Flow Indication 1/ 2/

- Safety Grade 100% - '~

2.1.8a (II.B.3)

Post-Accident Sampling Short-term (Category A) nodification;

- Modifications (long-term Category B) 75%

12/1/82 co mlete 2.1.8b (II.F.1)

Radiation Monitors

- Effluent Monitors 75%

1/1/83 m nitors have been sent to Batelle for calibration

- Iodine / Particulate Monitors 75%

10/1/82 Additional Itans

1 (II.F.1) Containment Pressure 90%

10/1/82 (sa fety-control grade cmplete

2 (II.F.1)

Containment Water Invel grade) 95%

10/1/82

3 (II.F.1) Containrent Ilydrogen

, _70%

12/1/82

ESTIMATED PERCENT COMPLETION ITEM COMPLETE DATE NOTE LT-4 Drcrgency Preparedness Bnergency Comunications

- Install control room anergency teleplene 100%

- Connect anergency teleplone equipnent to vital power 100%

Bnergency Facilities

- Install liigh radiation monitoring alann systan 90%

9/1/82 Board ImIosed Requiranents (Deconber 14, 1981 PID)

Plant Design, Fbdification and Procedures Findings II.E. Pressurizer IIcaters Test data to be forwarded

- Danonstrate ICS pressure control w/IIPI 100%

to NI1C by 9/1/32 II.K Canputer

- Incore thennoccuple backup display (not safety grade) 75%

10/1/82 II.M Safety Systan Status Panel

- Systan Status hinunistrative Controls 100%

ESTIMATED PERCENT COMPLETION ITEM COMPLETE DATE NOTE short-tem II.N Control Rom Design 97%

10/1/82 ong-term Cycle 6

- Correct tUREG-0752 deficiencies 10%

Startup II.P Systans Classification

- Upgrade Pressurizer IcVel Instrument Power Supplies 95%

9/1/82 II.Q EFW Reliability (see detailed question on long-tem order Itan B.2.1.7a)

Cycle 6

- Safety grade automtic EEW control 10%

Startup

- Install following long-tem EFW modifications 100% 1/

(a)

EFW cavitating venturis (b) CWST level alarm (safety grade) 10%

Cycle 6 Startup (c)

OPSG high level alam 10%

Cycle 6 Startup (d)

Safety grade isolation of MFW on OPSG overfill 0%

Cycle 6 Startup (e)

Upgrade noin steam rupture detection system to safety grade 0%

Cycle 6 Startup IKTIES - l_/ OJnstructim Ocuplete awaiting plant acceptance 2/ Construction Qmplete awaiting testing during hot functional testing or power escalation testing.

Question II.A:

In letters dated April 22, 1982 and May 13, 1982, the licensee notified this Board that certain steam and water tests exhibited valve instability that resulted in damage to the safety relief valve.

Throughout the hearing, licensee maintained that the feed and bleed mode of forced core cooling relied upon these valves to provide a release pathway for excess coolant.

In light of these tests results, how does the licensee plan to ensure that safety relief valves are capable of performing their function during feed and bleed when they may be called upon to open and close frequently with both steam and water flow mixtures?

LICENSEE RESPONSE Prior to addressing the actions Licensee will take in response to the results of the EPRI valve testing program, Licensee believes that it would be appropriate to briefly review the situations in which the feed and bleed mode of core cooling might be utilized.

Feed and bleed cooling is not required except when postulating events which are beyond the plant design basis, i.e.,

an extended loss of all main and emergency feedwater or certain accident conditions in conjunc-tion with an extended loss of all feedwater.

See Jones, ff.

Tr. 4588, at 3; Tr. 5201 (Jones).

Secondly, it should be noted j

that, while the analyses of feed and bleed cooling capability t

have assumed the use of the safety valves for the bleeding function, the PORV may be utilized to perform this function if it is available.

Keaten and Jones, ff. Tr. 4588, at 7-8; Tr.

8761 (Jones).

The EPRI steam and water tests, referred to in our l

letters dated April 22, 1982 and May 13, 1982, in which the _

test valve exhibited instabilities, were performed on a long inlet (loop seal) configuration.

This configuration is representative of the TMI-1 plant specific inlet piping.

The EPRI test results for safety valves and subcooled fluid discharge have shown that the safety valves exhibited stable performance for all fluid inlet conditions when tested on a short inlet configuration.

Based on these results and a review to ensure that the EPRI test conditions bound the TMI-l specific requirements, Licensee believes that the TMI-1 safety valves will perform in a stable manner if they are on a short' inlet.

Therefore, Licensee is presently planning to modify, by restart, the inlet piping to eliminate the loop seal and move the valves into a short inlet configuration at the nozzles on

~he pressurizer.

t Upon completion of these modifications, the safety valves will be capable of performing their function during the feed and bleed mode of core cooling when they may be called upon to open and close frequently with both steam and water l

flow mixtures.

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Question II.B:

The status list indicates that the installation of the Emergency Feedwater (EFW) automatic initiation is completed as control grade equipment (Item A.8.2.1.7a) but that further modifications up to safety grade will be partially completed by August 1982, and a footnote indicates that addi-tional long term modifications are scheduled for the first refueling after restart.

During the hearing, the staff testified that emergency feedwater modifications should be completed by late 1982 (Ross, Tr. 15,577).

1.

Which, if any, of the modifications discussed in paragraphs 1028-1034 of the partial initial decisiom (PID)

LBP-81-59, 14 NRC 1211 (1981), will not be completed before restart?

2.

What are the reasons for the delay beyond the completion date estimated by the staff during the hearing?

LICENSEE RESPONSE The EFW modifications described in I.D.,

1028-1034 are all short-term modifications undertaken in accordance with the terms of the Commission's August 9, 1979 Order and Notice of Hearing, CLI-79-8, 10 N.R.C.

141, 144.

See Staff Ex. 1 at Cl-1 through Cl-12 and C8-34 through C8-40.

Each of these modifications, described below, will be fully implemented prior to restart.

1.

Safety-grade, automatic initiation of EFW on loss of all four reactor coolant pumps (4 RCPS) or loss of l

both feedwater pumps (2 FWPS) has been installed.

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Redundant, safety-grade flow indicators for EFW flow.

to each steam generator have been installed.

3.

The EFW flow control valves have been modified to fail open on loss of instrument air.

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Operator control of EFW flow to each steam generator independant of the ICS has been provided.

5.

A redundant two hour air supply to furnish instrument air to the EFW control valves and related systems has been installed.

6.

Alarms, signifying a 20 minute supply of water remaining in the condensate storage tanks, have been provided.

7.

Redundant, safety-grade steam generator level indication, used in conjunction with icem 4 above, has been provided in the control room.

In addition to the short-term modifications discussed above there are certain long-term EFW modifications associated with Item II.E.1.1 of NUREG-0737 which are being undertaken.

I.D.,

11 1037-1038, 14 N.R.C.

1211, 1364.

These are the modifications referred to by Dr. Ross at the transcript page cited by the Appeal Board.

As Dr. Ross testified, the NUREG-0737 implementation date for these long-term modifica-tions was January 1, 1982, but it was thought that procurement and design problems might result in a delay in implementing certain of the design modifications until the Cycle 6 refueling outage (i.e.,

approximately 1 year after restart).

Tr. 15,577 (Ross); see also I.D.,

1 1038, 14 N.R.C.

1211, 1364.

The current status of the long-term modifications is set forth below:

1.

Cavitating venturis, one per steam generator, have been installed.

2.

Safety-grade low level alarms with the same setpoint l

as short-term item 6 above will be installed during the cycle 6 refueling outage.

3.

Safety-grade steam generator high level alarms will be installed during the cycle 6 refueling outage.

4.

Safety-grade isolation of main feedwater on overfill of a steam generator (hi-hi level in downcomer) will be installed during cycle 6 refueling outage.

5.

The main steam rupture detection system will be upgraded to safety-grade during the cycle 6 refueling outage.

6.

An additional safety-grade signal, based upon steam generator low-low level, will be provided for EFW initiation.

l In conjunction with these six long-term modifica-tions, as noted by the Appeal Board in Question II.B.,

Licensee will further upgrade the EFW system by providing safety-grade automatic control of EFW flow to the steam generators.

It is this long-term modification which is referred to by the Staff in footnote 3, p.

6 of Appendix B to the Appeal Board's Order of July 14, 1982.

To clarify, the TMI-1 EFW system at restart j

will have safety-grade automatic initiation (i.e.,

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starting of the EFW pumps) as described in snort-term item 1 l t

above, but will not have safety-grade automatic control.

See I.D.,

11 1036, 14 N.R.C.

1211, 1363.

Redundant, safety-grade automatic control of EFW'to each steam generator, based upon steam generator level, will be installed during the cycle 6 refueling outage.

At the time that testimony was presented on the TMI-l EFW system, it wne thought that restart would occur in late 1981, and that most of the long-term modifications could be accomplished during the cycle 6 refueling outage, then sched-uled for late 1982.

However, it must be realized that the provision of safety-grade automatic EFW flow control and long-term modifications 3, 4 and 5 above required the design and procurement of an entirely new four channel safety-grade system.

The design engineering for this system required the performance of additional analyses beyond those originally projected, thereby resulting in a delay in the original imple-mentation schedule.1/

Further delays have been created by the long lead time for delivery of properly qualified hardware.

In view of the time and labor required for installation, the 1/

The additional engineering analyses were required due to unanticipated complexities inherent in attempting to integrate the new system with existing plant systems, i.e.,

assuring that there are no unacceptable interactions with existing non-safety-grade systems and resolving human factors considerations with respect to consistency of displays.

Additionally, engi-neering work on the long-term modifications was delayed approx-imately six months by the need to concentrate engineering resources on resolving the TMI-1 steam generator problems.,

modifications will require an extended outage and will therefore not be completely implemented until the cycle 6 refueling.

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Question II.E:

During the hearing, the licensee indicated that the high point vents were planned to be installed prior to restart (Tr. 16, 580).

NUREG-0737 requires the installation to be complete by July 1, 1982.

The status list indicates that the completion date is "to be determined."

What progress has been made in complying with the requirements of NUREG-0737 for the installation of high point vents?

Are the vents and their controls fully safety-grade?

If the high point vents will not be installed prior to restart, what is the justification for allowing operation TMI-1 before the vents are installed?

LICENSEE RESPONSE Licensee's system for providing the capability to vent noncondensible gases from the Reactor Coolant System (RCS) is described in Section 2.1.2.2 of the Restart Report (Lic.

Ex. 1).

The RCS Venting System-will consist of three separate sub-systems:

vents from the top of the pressurizer, discharg-ing to the Reactor Coolant Drain Tank (RCDT); and vents from the top of both hot legs and from the top of the Reactor Vessel Head, which will discharge directly to the containment.

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l pressurizer vent has been installed and will be operable at restart.

The design of the balance of the RCS venting system has progressed through the production of flow diagrams, piping l

drawings, pipe support drawings and electrical and instrumenta-tion details.

The entire RCS venting system will be safety-grade.

See Lic. Ex.

1, 55 2.1.2.2.1, 2.1.2.2.6.

The schedule for implementation of this modification as set forth in NUREG-0737 has been superceded by a recent l

revision 2/ to 10 C.F.R. 5 50.44(c)(3)(iii), which requires 2/

See Final Rule, Interim Requirements Related to Hydrogen Control, 46 Fed. Reg. 58,484 (Dec.

2, 1981).

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installation of these vents "by the end of the first scheduled outage beginning after July 1, 1982 and of sufficient duration to permit required modifications..." (emphasis added).

In accordance with this requirement, Licensee plans to install the balance of the RCS venting system during the first refueling outage following restart (i.e.,

the cycle 6 refueling outage).3/

The Appeal Board has also requested a justification for allowing operation of TMI-1 prior to the installation of the high point vents.

In that'the TMI-1 vents will be installed in accord with the schedule for all operating reactors, Licensee does not believe there is a need to provide special justification for permitting TMI-1 to restart.

However, it should be noted that the high point vents are solely a back-up which will be provided to mitigate a beyond design basis event -- the generation of noncondensible gases --

l which is not expected to occur in the future.

Tr. 4991-93 (Jensen).

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Licensee notes that the installation of the high point vents was never a pre-restart commitment, although installation by restart was previously thought to be possible.

See Tr.

16,580 (Keaten).

Question III.B:

In Paragraph 771 of its PID, the Licensing Board directed the staff to verify that procedures to connect the pressurizer heaters to the diesels include provisions to assure that the heaters would not be reconnected to onsite power until stabilization of the event that caused their disconnection.

The status list attached to SECY-82-250 indicates that this item is complete.

What provisions h9ve been included in the procedures to comply with the Licensing Board's direction?

LICENSEE RESPONSE In Paragraph 771 of its PID the Licensing Board directed the Staff to " verify that the plant procedures include provisions to assure that desired pressurizer heater loads will not be reconnected to the on-site power supply after they have been automatically separated until stablization has been achieved following the event that caused their disconnection."

(Emphasis supplied.)

Licensee understands the Board's direc-tion to refer to the stabilization of electric supply to all systems connected to the diesel generator following the event which caused the disconnectionof the pressurizer heater load, rather than stabilization of the event itself.

Thus, a small break LOCA could result in an ES signal which would automati-cally disconnect the pressurizer heaters as well alaactuate the emergency core cooling systems.

Stabilization of the LOCA event itself could require a substantial period of time.

Stabilization of the electric power supply to the emergency core cooling systems or other connected loads would normally l

occur in a much shorter interval.

In other situations, l !

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however, stabilization of the event which caused the disconnection could be synonomous with stabilization of electric power supply.

Thus a fault in a pressurizer heater could cause both overcurrent and undervoltage, either of which would automatically result in disconnection of the power supply.

Maintenance of a stable power supply could not be accomplished upon reconnection without correction of the fault condition.

In this situation therefore Licensee's procedures call for a full evaluation of the cause of disconnection.

A " caution" has been added to Revision 17 of Emergency Procedure 1202-29, " Pressurizer System Failure",4/

which requires evaluating the cause of the pressurizer heater trip and verifying stabilization of electric supply to all

' systems connected to the diesel generator prior to the recon-nection of the heaters.

The procedure caution, which is applicable when the diesel generators are supplying plant load, is set forth verbatim below:

CAUTION:

Should the pressurizer heaters be tripped out as a result of an ES signal, overload j

or undervoltage condition, they are not to be reconnected until the cause of the l

trip has been fully evaluated and stabil-ization has been achieved following the event.

Stabilization shall be considered i

to be achieved when block loading is completed, voltage is at its normal value

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and the load on the diesel does not i

exceed 2850KW.

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An earlier version of this procedure (Revision 15) was admitted as Licensee Ex. 50.

The new caution has been added

'following Step L at page 12.1 of Licensee Ex. 50.

"f Question III.C:

PID Paragraph 943 listed measures that have been or will be taken at TMI-l to improve protection against small break LOCAs.

One of those measures was the improvement of the HPI systems by adding cavitating venturis and cross-connection lines.

It was also stated that the system being installed will automatically perform the balancing of HPI flow.

i How is this to be accomplished and what is the completion status of these HPI modifications?

LICENSEE RESPONSE The HPI System modification, adding cavitating venturis and cross-connection lines, has been completed.

Prior to restart, testing will be performed to demonstrate system performance.

Tr. 5605 (Jensen); Lic. Ex.

1, Supp.

1, Response to Question 36c.

A complete description of the HPI system modifica-tions and performance evaluation is contained in the Restart Report (Lic. Ex. 1) at Section 3; Supplement 1, Part 1, Questions 36b, 36c and 37; and, Supplement 1, Part 3, Questions 1,

2 and 3.

The modifications provide for assuring adequate HPI flow in the event of either a break in the HPI lines or in the event of a makeup value failing to open.

In the first case, the installed flow-limiting devices (the cavitating venturis) will limit the amount of coolant injected into the broken line, thereby limiting the fluid discharged out of the break.

In the event that one of the valves supplying HPI fluid (the MUV-16 valves) fails to open, the cross-connect devices will function and direct HPI flow to all four HPI lines.

Prior to these modifications, the control room operators were -

required to manually limit or redirect the HPI flow.

See Jensen, ff. Tr. 5496, at 7; Tr. 5605 (Jensen).

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Question III.D:

In Paragraph 1064 of its PID, the Licensing Board directed the staff to certify to the Commission that the licensee has made rea'sonable progress in initiating a program for long-term solution of the steam generator bypass logic problem.

What progress has been made by the licensee in solving this problem?

What interim methods will be used 'co ensure that plant operators are aware of the problem and the actions to be taken in the event of isolation of both steam generators?

LICENSEE RESPONSE In order to eliminate the concern raised by the Licensing Board in I.D.,

11 1060-1064 (i.e.,

isolation of all feedwater flow to both steam generators), Licensee has proposed implementing two changes to the Main Steamline Rupture Detection System (MSLRDS).

The proposed changes consist of the addition of cavitating venturis to the EFW lines and the deletion of,the MSLRDS signal to the EFW system.

The proposed design changes were submitted for Staff approval by letter dated August 2, 1982, from H. D. Hukill to John F.

Stoltz.

This letter, which includes a safety evaluation of the proposed change, is attached hereto as Attachment A.

Licensee antici-pates implementing this design change prior to restart, subject to review and approval by the Staff.

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Question III.I:

In a footnote to Paragraph 919 of the PID, the Licensing Board indicated that the licensee planned to perform an in-plant communications study in 1981.

What is the status of that study?

If completed, please briefly summarize results and present status of implementation.

LICENSEE RESPONSE As noted by the Licensing Board, Licensee planned to begin an in-plant communications study in 1981 and to complete this study in 1982.

I.D.,

1 919, n.

109.

The performance of this study, however, received a lower priority than many of the other actions being taken by Licensee prior to res' tart and some slippage in schedule has occurred.

Proposals to perform a communications study at TMI-1 were solicited from four consult-ing firms in April, 1982.

Three firms responded in June of this year and their proposals are currently being reviewed by Licensee.

Selection of the consultant, awarding the contract and commencement of the study should be completed in the fall of 1982.

The study is expected to take six to nine months.

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Question III.K:

The Licensing Board indicated in PID Paragraph 1264 that a tunnel-like barrier for personnel passage between the Unit 1 control tower and the Unit 1 auxiliary building will be completed before restart.

What progress has been made in completing this modification?

LICENSEE RESPONSE The commitment in I.D.,

1 1264 to construct a tunnel like barrier which will provide personnel passage between the Unit 1 control tower and the Unit 1 auxiliary building, and which will also form part of a barrier that will seal the open areas between the. Unit 1 auxiliary building and the Unit 1 fuel handling building has been completed.

Licensee is currently in the process of designing a program to test the adequacy of its phase I ventilation separation program, which will be submitted to the Staff for approval.

This program will include a test of the adequacy of this barrier.

Question IV.B:

In the event that the pressuriser heaters fail to operate while the plant is operating at full power, (1) how much time would it take to achieve RHR system initiation conditions and then cold shutdown?

(2) how would pressure control be performed during cooldown to conditions allowing RHR system operation?

(3) how soon after shutdown from full power conditions does the-RHR system have sufficient decay heat removal capability?

LICENSEE RESPONSE (1)

Based upon data concerning several actual TMI-1 shutdown and cooldown events, Decay Heat Removal (DHR) Systems /

initiation will occur between 8 and 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months following initia-tion of the shutdown.

Cold shutdown conditions (less than 200*F) are normally achieved in an additional two hours.

However, in the event of a serious plant casualty, a controlled shutdown /cooldown to less than 200*F could be achieved in approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .s/

(2)

The TMI-1 normal cooldown procedure requires that the pressuriser heaters be turned off; therefore, failure of the pressuriser heaters would not adversely impact a normal plant cooldown.

RCS pressure control during a normal plant i

cooldown is achieved by use of the pressurizer sprays, to reduce pressure as necessary.

5/

The Appeal Board question refers to the RHR (residual heat removal) system.

This system is designated as the Decay Heat Removal System at TMI-1.

s/

The DHR System can be actuated at approximately 250*F and 320 psi.

See Tr. 16,556 (Colitz).

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(3)

The DHR System has sufficient capacity to remove 100% of the decay heat immediately following a controlled plant shutdown.

The DHR System is not capable of removing the maximum decay heat present at the instant following a reactor trip.

Following a reactor trip, there is an interval of approximately 160 seconds before the core decay heat level drops to the DHR System capacity.

During this time, decay heat removal is accomplished by the use of other plant systems (i.e.,

steam generators, HPI, LPI).

Dear Sir:

Three Mile Island Nuclear Station, Unit 1 (D:I-1)

Operating License No. DPR-50 Docket No. 50-289 Main Steamline Rupture Detection Systen Design Changes I

In its Partial Initial Decision (PID) on design (See PID 1060-1064) the Atomic Safety and Licensing Board (ASLB) required that GPUN investigate design changes to the Main Steamline Rupture Detection Systen (MSLRDS).

The changes are to prevent unnecessary isolation of feedvater under single fallure conditions.

A description and evaluation of the changes to the MSLRDS is attached.

This is subnitted f or NRC approval as requested by the ASLB (FID 1064).

l Sineerely, HD dill Director, TMI-1 HDE:CWS:vjf Attachnent ec:

R. C. Haynes R. Jacobs "ATIKhTir A" e

t ATTACEME2C 1 l

Main Stemmline Rupture Detection Syste= Design Changes t

1.

INTRODUCTION The Main Steamline Rupture Detection System (MSLRDS) is actuated on low steam I

generator pressure (below approximately 600 psig) and automatically closes the Emergency Feedwater (EN) and Main Feedwater (MW) control valves to isolate feed flow to the depressurized steam generator.

If subsequently pressure rises above 600 psig in a steam generator the EW associated with that steam generator is restored.

This MSLEDS action prevents overpressurization of containment from stem =11ne breaks in containment.

The ASLB was concerned that the MSLRDS would block all feedwater, including E W, to the steam generators in certain scenarios when it should not be blocked.

II.

SOLUTION

[

The proposed solution to the above concern ; consists of _the addition of cavitati=g venturis and the deletion of the MSLRDS signal to the Energency Feedwater System.

Low OTSG pressure, which actuates the MSLEDS, can result from either a severe l

overcooling or a mmE steamline break event.

The original design required operator j action to bypass MSLRDS to prevent a loss of heat sink if a lov OTSG pressure i

condition developed and single failure then blocked M.

The addi: ion of the l

! /

cavitating venturis to the EN System and removal of the MSLRDS from the EW valves eliminates operator action to provide m to the intac: OTSG in the event of a single failure.

Since the venturis also limit E W flow, the MSLRDS is no longer required for EW and need not Be up graded to safety grade (FID 1037e) since it is eliminated.

III.

SAFETY EVALUATION l

l Deletion of the MSLRDS from the EW valves does not affect any of the FSiR acceptanc(

criteria.

The basis for this judgment is as follows:

l The MSLRDS was installed to prevent overpressurization of the containment due to a Main Streamline Br.eak (MSLB).

Removal of the MSLRDS fro = the EF vcives vill l

make IMI-1 feedvater isolation functionally the same as TMI-2 in its response to a MSLB.

The M-2 MSLB analysis was reviewed and approved by the NRC (See M-2 i FSAR, Chap.15, Appendix B).

The TMI-2 analysis is bounding f o-T ' #c the following reasons:

r a)

The TMI-1 venturis limit total flow to a lower flow rate than the TMI-2 venturis (1150 GPM vs. 1250 G?M), and t

b)

TMI-1 cannot have a double OTSG blowdown in containmen:

(limiting pressurization accident for DG-2) because the main steam isolation valves are stop check valves for TMI-1.

i Deletion of the.MSLRDS from the E W valves does no: increase the orob-ability of occurrence of a stearline break accident.

The consequences f

of the accident, as analyze'd in the DC-2 FSAP, have not been increased, t

i, g g g.

Reactor Building ovarpraccurizction dons ntt occur and th2 rsquired hen: rc= oval capability to prevent fuel damage is provided.

Specifically, fuel danage vill not result, off-site doses will not be increased, and steam generator tube integrity will not be compromised. The conclusions are confirmed in the Restar:

l Repor^

Section 8.3.9 which ref erences the TMI-2 75/J., Chapter 15, Appendix B.

EW D is continued throughout the referenced analysis.

Addition of cavitating co the EW system limits the maximum EW flow at TMI-1 and assures that ventut_

the referenced TMI-2 analysis is bounding for TMI-1.

Further= ore, the syste=s, setpoints and/or plant conditions that are utilized in the referenced analysis are applicable to both TMI-l and TMI-2. (The ERC was also advised of the TMI-l desi n modification in Met-Ed response to IE 3ulletin 80-04 May 9, 1980 TLL 228).

F The referenced TMI-2 analysis assumed 1%AK/I shutdown margin and denenstrated that the core does not return to criticality and that the fuel rods do not violate a DNBR of 1.0.

Other assumptions made in the ref erenced analysis are more severe than those allowed by TMI-l Tech. Specs., most notably power level (2772 MR),

and RCS flow (100%).

The design peaking factor of 1.78 used in THI-2 analysis exceeds the curren design peaking factor for TMI-1.

The referenced stea=line break analysis also demonstrated acceptable offsize doses and showed ha: OTSG tube stresses resulting from the accident are acceptable.

Tube stress conditions were evaluated in BAW-1588.

The results of this evalua: ion bound :he ~2".-l EW system design with the MSLEDS signal ' deleted f ro the 75,0 valves.

Other considerations and/or questions:

Overfilling of the OTSG is an issue which has been raised and is docunented in the Restart Report, Supplement.1, Part 2, Question 2.

The analysis presented in the TMI-1 FSAR did not take credit for EW isola: ion via the MSLEDS signal.

The EW flow rate assumed was 1500 GPM to one (1) OTSG at 600 PSIG (the MSLEDS set noint), this assumed flow is 2-b time the flov ra:e available te one (1)

OTSG from the TMI-l EW system with cavitating ven:uris installed.

Filling of the OTSG from the 50% operating range took 6.6 minutes using these assu=ptions.

Therefore, the operator would have (with the venturis installed and a fully opened control valve) approximately 16 minutes te terzinate an overfill condition due to EW flow.

The revised design therefore allows suf ficient time for the operator to terminate EW.

As discussed above, deletion of the MSLRDS signal to the EW valves does not introduce any accident or malfunctions not previously evaluated, nor does it increase the 14ka14 hood of occurrence or consequences of any acciden: analyzed in the TMI-1 FSAR.

In conclusion, this modification does not introduce any acciden: er ralfunctions not previously evaluated, nor does it increase the likelihood of occurrence or consequences of any accident as analyzed in the TMI-l TSAR.

No safety margins vill be reduced as a result of the codification.

Further: tore, the revised cesign OTSG improves the reliability of the EW Systen to deliver flov te the in:ac:

and vill not create a containment overpressurization or CTSG overfill condition.

"A'ITAODENT A"

4 4

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

)

METROPOLITAN EDISON COMPANY

)

Docket No. 50-289

)

(Restart)

(Three Mile Island Nuclear

)

Station, Unit No. 1)

)

CERTIFICATE OF. SERVICE I hereby certify that copies of " Licensee's Response to the Atomic Safety and Licensing Appeal Board's Order of July 14, 1982" were served this 12th day of August, 1982, by hand delivery upon the parties identified by one asterisk and by deposit in the U.S. mail, first class, postage prepaid, to the other parties on the attached Service List.

~=m Thomas A.

Baxter, P.C.

l 1

l l

~

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

)

METROPOLITAN EDISON COMPANY

)

Docket No. 50-289

)

(Restart)

(Three Mile Island Nuclear

)

Station, Unit No. 1)

)

SERVICE LIST

Gary J. Edles, Esquire Y James M. Cutchin, IV, Esquire Chairman Office of the Executive legal Director Atcmic Safety and Licensing Appeal U.S. Nuclear Regulatory Ccunmission Board Washington, D.C.

20555 U.S. Nuclear Regulatory Cocmission Washington, D.C.

20555 Docketing and Service Section Office of the Secretary

Dr. John H. Buck U.S. Nuclear Regulatory Cmmission Atcmic Safety and Licensing Appeal Washington, D.C.

20555 Board Panel U.S. Nuclear Regulatory Cmmission John A. Ievin, Esquire Washington, D.C.

20555 Assistant Counsel Pennsylvania Public Utility Ccrimission

Dr. Reginald L. Gotchy P.O. Box 3265 Atcmic Safety and Licensing Appeal Harrisburg, Pennsylvania 17120 Board Panel U.S. Nuclear Regulatory Cottmission Robert Adler, Esquire Washington, D.C.

20555 Assistant Attorney General 505 Executive House Ivan W. Smith, Esquire P.O. Box 2357 Chairman Harrisburg, Pennsylvania 17120 Atcmic Safety and Licensing Board f

U.S. Nuclear Regulatory Conmission Ms. Icuise Bradford Washingtcn, D.C.

20555

'IMI AIERT 1011 Green Street Dr. Walter H. Jordan Harrisburg, Pennsylvania 17102 Atcznic Safety and Licensing Board Panel

<Ellyn R. Weiss, Esquire 881 West Outer Drive Harnen & Weiss Oak Ridge, Tennessee 37830 1725 Eye Street, N.W., Suite 506 Washington, D.C.

20006 Dr. Linda W. Little Atcmic Safety and Lice tsing Board Steven C. Sholly Panel Union of Concerned Scientists 5000 Hennitage Drive 1346 Connecticut Avenue, N.W., Suite 1101 Raleigh, North Carolina 27612 Washington, D.C.

20036

Jordan D. Cunningh e, Esquire 2320 North Secord Street prrishrg, Pennsylvania 17110 Gail B. Phelps ANGRY 245 West Philadelphia Street York, Pennsylvania 17404 Willi e S. Jordan, III, Esquire Harmon & Weiss 1725 Eye Street, N.W., Suite 506 Washington, D.C.

20006 Chauncey Kepford Judith H. Johnsrud Envirornental Coalition on Nuclear Power 433 Orlando Avenue State College, Pennsylvania 16801 Marjorie M. Aamodt R. D. 5 Coatesville, Pennsylvania 19320 e

/

1

_~e.

.y

._ __

ML20062H703

Text

Dear Sir:

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