Text

Forwards Questions & Answers for Hearing Backup Book Re TMI Incident,To Be Provided to Chairman
	ML19220C489
Person / Time
Site:	Crane
Issue date:	04/09/1979
From:	Miraglia F; Office of Nuclear Reactor Regulation
To:	Rehm T; NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO)
References
	NUDOCS 7905100335
	Download: ML19220C489 (24)
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Eisenhut/Mi 9 Job T 4/8/79 ACR$ CCNCERNS TMI -2 INCIDENT 1.

What assurance do we have that the TMI event will not hacoen at another 8?.W coactor tomorrow?

The initiating event (loss of condensate and feedwater pumps) is an anticipated transient, i.e., it is expected to occur and may occur tomorrow.

However, the severity of the consequences in the TMI-2 incident was caused by multiple circumstances and actions which are addressed by the April 5, 1979 IE Bulletin 79-05A.

The purpose of tnat Bulletin is to prevent recurrence of the contributing circumstances and actions thereby preventing recurrence of the incident.

The Bulletin requires licensees to.

review their procedures and operator actions and determine that they are adequate to prevent a similar incident particularly with regard to termination of HPI flow and tripping of RCS pumps and with regard to reliance placed on pressurizer level indicators in determining operator actions; review containment isolation singals to determine that proper isolation will be provided; and assure that adequate auxiliary feedwater flow will be provided by observing specific requirements provided in the bulletin regarding auxiliary feedwater systems cperability and availability when the plant is at power.

2.

There has been much discussion of this accident as a B&W orcblem.

What makes tnis acticent unique to B&W PWRs?

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The " loop seal" design of the connection between the pressurizer and the hot leg makes the plant susceptible to the so-called "monometer effect." When an extended blowdown through the pressurizer occurs, there is the possibility that the pressurizer will cease to be the highest temperature volume in the primary system.

Higher temperature and higher steam pressurizer in the core then generates a steam volume which could fill part of the core and the hot leg and prevent water in the pressurizer from flowing downward into the core.

Since the operator would likely assume that adequate water level in the pressurizer assures that the core is covered with water, this " loop seal" design could lead to incorrect operator action (throttling or securing water flow to the core).

' hat is, he would believe the core to be covered when i is not.

The loop seal design is present on B&W plants, but is not used on CE plants and recent design Westinghouse plants.

Also, the once-through steam generator design used on E&W plants 'as a smaller secondary water reservoir, and it is a more efficient des,

for providing heat transfer to the secondary water.

Therefore, for both of these reatons, it is more prone to quick dryout upon loss of secondary water (feedwate).

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2. Is there anythino unicue about C&W Contair. ment Isolation Featuras?

TheTMI-2containmentisolatesonlyonhighpressure(fkpsig)inthe X

containment. Many other plants (B&W z.nd other vendors) also isolate on ECCS initiation signals.

Such a cantainment isolation signal, if co

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4 it had been present on TMI-2, would probably have greatly reduced releases from containment during early stages of the TMI-2 incident.

Gus Lainas will provide addition <l information regarding this question.

4.

Are all auxiliary feedEater systems redundant and diverse All auxiliary feedwater systems FW) in PWRs are redundant.

None have only electric motor driven AFW pumps.

Most are diverse and have both electric motor and steam turbine driven pumps.

Scme have electric motor and diesel driven pumps. 8 cme do not have diversity within the Y

AFW, but are diverse to the mainfeedwater and condensate system.

These systems rely on steam turbine driven AFW pumps alone.

5.

Is there anything about the PORVs at THI-2? On B&W plants?

3 B&W plants (including TMI-2) utilize a de3ign where an electrical signal activates a solenoid which opens a pilot valve which in turn allows pressure to be introduced under the main valve seat, which th"n opens (Dresser pilot operated valve).

Although some PWRs also util#ze other valve designs many PWRs supplied by other vendors quite likely utilize similar or identical PORVs (review of records is continuing in this area).

6.

What are the cresent recuirements for coerability of auxiliary feedwater systems?

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e The relevant license conditions, or Technical Specifications for TMI-2 are attached.

It requires the reactor to be shutdown in 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if one train is inoperative and 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if both trains are inoperative.

The Bulletin 79-05A issued April 5 requires at least this same level

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of operability at all B&W designed reactors.

A copy of the Tech.

j Specs. for these plants is attached.

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The Tech. Specs for other PWRs vary somewhat depending on date of initial license and plant design.

A table summarizing the operability requirements is attached as are the individual plant specifications, q/

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SUMMARY

OF OPERABILITY REQUIREMENTS FOR PWR AUXILIARY FEEDWATER SYSTEMS (TIME TO SHUTDOWN)

B&W Designed PWRs (Bulletin 79-05A - 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months & 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

Plant One Train Incoerative Two Trains Inocerative Three Mile Is -1 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0 hours Three Mile Is -2 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Davis-Besse-1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Crystal River-3 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Oconee 1, 2 and 3 indefinite Indefinite Arkansas-1 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0 hours Rancho Seco 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months 0 hours CE Desianed PWRs Palisades indefinite (with fire pump)

O hours Ft. Calhoun 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 0 hours Maine Yankee 148 hours0.00171 days 0.0411 hours 2.44709e-4 weeks 5.6314e-5 months 0 hours Calvert Cliff 1'and 2 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Millstone 2 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months I hour 26 155

e CE Designed PWRs (continued)

St. Lucie 1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Arkansas 2 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour W Designed PWRs Beaver Valley 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour D.C. Cook 1 and 2 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Farley 1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Ginna indefinite 0 hours0 days 0 hours 0 weeks 0 months Haddam Neck indefinite 0 hours0 days 0 hours 0 weeks 0 months Indian Point 2 indefinite 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Indian Point 3 indefinite 0 hours0 days 0 hours 0 weeks 0 months Kewaunee indefinite 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months North Anna 1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I hour Point Beach I and 2 indefinite 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Praire Island 1 and 2 indefinite 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Robinson 2 indefinite 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Salem 1 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I hour San Onofre 1 indefinite 0 hours0 days 0 hours 0 weeks 0 months Surry 1 and 2 indefinite 0 hours0 days 0 hours 0 weeks 0 months q'

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W Designed PWRs (continued)

Trojan 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1 hour Turkey Point 3 and 4 indefinite 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Yankee Rowe 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months No second train Zion 1 and 2 indefinite 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

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o 7.

Is there anything unicue about the TMI canta:nment isolaton features?

Five other operating S&W designed plants have similar designs a: TMI, however, the actions described by the Bulletin will preclude a similar occurrence.

The large majority of other operating plants have containment isolation systems that by design would have prevented flooding of tile Auxiliary Building (i.e., loss of containment).

Most plants utilize safety injection as a signal to initiate containment isolation in addition to containment pressure.

SI was initiated 2 minutes for these other operating plants at that time, rather than at 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months as was the case at TMI. Therefore, little radioactivity would have been released.

8.

What is the sinale basic difference in the olant desians of other coerating olants that micht by itself orecluce a similar incicenTas at TMI?

We currently believe that the single most important difference in other PWR designs is related to the location of the pressurizer and routing of its surge line.

BWRs of course do not have a pressurizer.

OthertypesofPWRdesiIslocatethepressurizerandsurgelineso that core levels are directly reflected in the pressurizer where reactor system level is measured.

The TMI design requires operator interpretation of a number of instruments to properly identify reactor coolant system level.

Other designs lend themselves to more direct measurement.

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_Have we learn 2d anythina about the capability of instruments to withstand LOCAs?

To date, we have confirmed that much of the instrumentation will withstand high radiation exposures without failure.

Only one critical instrument has failed to date.

10.

Manacement Organization Q.

What are your concerns?

1.

The man in charge of the plant - Units 1 and 2 was not stationed at the plant.

The job was a corporate management job with the principal location offsite.

We felt there should be a single individual in charge of the entire plant, at the plant, to avoid any possible inter-unit conflicts, resolve priorities, handle site items such as security, emergencies, etc.

There was a lack of clarity about shift supervisors.

How were these concerns resolved?

1.

There was to be a plant superintendent, or one of the unit superintendents was to be designated as in charge of the plant.

This is acceptable.

2.b

2.

The organization was clarified.

Shift foreman (SR0s) on each unit reported to a single shif t supervisor (SRO) who had overall plant responsibility.

He was to be in charge of the site in off shifts.

This is acceptable.

11.

Secondary System Line Breaks Q.

Is there any connection between this item and the TMI-2 event?

No.

There was no steam or feed line break at TMI-2.

There are aspects of the analyses and the hardware discussed in this issue which entered into the actual event, but none of the future modifications would have had any effect on the event.

The SER discusses isolation of feed and emergency feed lines and steps taken to preclude single failure causing such isolation, but we did not consider total manual lockout of all emergency feed valves in our review.

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TMI-2 ACRS CONCERNS 12, Hydrogen Charging Line ou What is this hydrogen used for?

As a scavenging gas in the makeup tank to remove oxygen from the reactor coolant.

6 Where does the coolant water come from and where does it go from the makeup tank?

Comes frna the letdown line through various filters and other treatment.

After deoxygenation it is reinjected into the reactor coolant via the makeup pumps.

C Can hydrogen be absorbed in the water into makeup tank?

The tank is at approximately 100 F and 15 psig - very little hydrogen will be absorbed under these conditions.

c( How large is the MU tank?

600 CF total - 400 CF water, 200 CF hydrogen

c. How is the H2 admitted to the MU tank?

There are 5 " bottles" of hydrogen (1000 SCF each at 2400 psi) manifolded together.

One bottle at a time is valved through a pressure regulator into the makeup tank.

Is there any way to introduce this hydrogen into the reactor in an paccicent?

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s The makeup pumps serve as the high pressure injection pumps during an accident.

On an SI signal the valve from the BWST to the HPI pumps opens.

Procedures require that the valve from the MU tank be closed; however, even if it is not closed, equipment elevations tank charging procedures and pump alignment are such that a water seal will still be maintained in the MU tank discharge piping precluding H2 entering the pump (Met Ed had committed to close the MU tank valve automatically by the first refueling).

In addition, the SI signal will close redundant valves in the hydrogen charging line to the MU tank, limiting the amourt of hydrogen available.

g What if these equipment features and procedures do not work?

If (1) the MU tank valve is not closed and (2) the hydrogen isolation valves do not close automatically and (3) tank charging procecures are not followed, then a maximum of 1400 SCF of hydrogen (40 CF at 100 psi and.100 F) could be drawn into the RCS.

If in addition all five hydrogen bottles are valved in the pressure regulator, rather than just one, approximately 155 CFR could be drawn in.

(The sze of the bubble in the TMI-2 event was in the order of 1000 CF).

d Was there any indication that any hydrogen in the TMI-2 bubble came from tnis source?

(Still checking at site).

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DISPOSITION OF ITEMS IN ACRS LETTER ON TMI-2 1.

Augmented Startup Program to terify load follow and power transient core power characteristics.

Disposition An acceptable augmented startup program was performed on an identical core at Rancho Seco.

These results in conjunction with a closely monitored normal startup program at TMI-2 were felt to be adequate.

See p. 18-1 of SER Supplement No. 1.

2.

Asymmetric Loads on Reactor Vessel Disposition:

Based on the plant design, the probability of occurrence of the break, and the generic review in progress, we concluded that plant operation prior to final reslation woould be acceptable.

See p.182 of SER Supplement No. 1.

3.

ATWS Disposition:

We concluded that operating limitations until final resolution of the generGc ATWS issue were not necessary.

See p. 18-3 of SER Supplement No. 1.

4.

Flood Emergency Plan Disposition,'

We concluded that additional information furnished by the applicant satisfied the ACRS concern.

See p. 18-3 of SER Supplement No. 1.

5.

Fire Protection Disposition *.

We reviewed the applicants' fire hazards analysis and fire protection program reevaluation, including a visit to the plant.

Required plant modifications and an acceptable implementation schedule were identified.

The license was conditioned to require such implementation.

See: p.

18-4 of SER Supplement No. 1; p. 18-1 of SER Supplement No. 2; p. 9-1 of SER Supplement No. 2; Facility Operating License No. DPR-73.

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6.

Hermetic Seals on Instrumentaticr. in Containment Disposition *,

The applicant identified all app hcable instruments and described their construction and related maintenance procedures.

This concern was also added as an ACRS generic concern, and as such its resolution will be considered for_this plant when developed.

See p. 18-4 of SER Supplement No. 1; p. 18-1 of SER Supplement No. 2.

7.

Interference with Safety Functions of the DC System by Non-Essential Loads Disposition:

We reviewed the applicants rs',ponse to this concern, and concluded that the present design is consistent with that of other plants found acceptable.

The ACRS requrested that this matter be considered in our generic review of reliability of power supplies.

We will consider any changes identified in this review for TMI-2.

See p. 18-4 of SER Supplement No. 1; p. 18-1 of SER Supplement No. 2.

8.

Release of Hydrogen Frcm Hydrogen Charging Line Disposition:

The hydrogen charging line, which runs only through a portion of the auxiliary building, was rerouted and redesigned to preclude effects on safety-related equipment.

See p. 18-5 of SER Supplement No. 1; p.

18-2 of SER Supplement No. 2.

9.

Instrument Line Failure Affecting Plant Controllability Disposition:

Applicant analysis showed that no instrument line breaks presented plant controllability problems of significance to publis safety.

Our.

generic review of the plant interactions will also consider instrument line failing.

See p. 18-5 of SER Supplement No. 1; and p. 18-2 of SER Supplement No. 2.

10.

Management Orcanization Disposition:

Additional informotion on this matter was subsequently submitted by the applicant which satisfied our concerns.

See p. 18-5 of SER Supplement No.1; p.18-3 of SER Supplement No. 2, p.13-1 of SER Supplement No. 2.

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Secondary System Lire Breaks Disposition:

This ACRS item refers to our open issue on this subject described en

p. 15-2 of the SER.

The applicant has not completed its analyses of a spectrum of steam line breaks and we had not yet completed our review of the recently reviewed feed line break analysis.

We had required modifications to assure that single failure could not pt event isolation of the feed and emergency feed systems.

Subsequently, our review of secondary system breaks was completed.

We concluded that modifications to the secondary system were required to assure that consequences of breaks were mitigated by safety grade equipment, and that operation until such modifications were implemented (by the end of the first refueling outage) was acceptable.

License conditions were imposed to assure such implementation.

See p. 15-1 of SER Supplement No. 2.

12.

Additional Means to Follow the Course of an Accident Disposition:

This issue was considered generic in nature and as such would be dealt with on TMI-2 when a generic solution was developed.

See p. 18-6 of SER Supplement No. 1.

13.

Sabotage Disposition:

We performed additional review on particular structures and concluded that their design provided an acceptable degree of security.

We also required submittal of 6n amended security plan in compliance with 10 CFR Part 73.55.

See p. 18-6 of SER Supplement No. 1.

The amended security plan was approved and made part of the licanse on February 23, 1979.

14.

Generic Items Dispositicn:

We prepared Appendix C to the SER which noted the disposition and status of appropriate generic items.

See p. 18-6 of SER Supplement No. 1; p. C-1 of SER Supplement No. 2.

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Questions and Answers - Exemptions General

1. et What decision criteria are used in decidina whether to issue an Exemption (Confirmation Order) o-to issue an order requiring a snutacwn.

Public health and safety is paramount and must be assured befo.e any exemption is issued.

Two recent cases can be used as examples.

The shutdown of the five reactors because of the error in the seismic design methods was determined to be required because no analyses were available to assure that design requirements could be met.

In the case of the small break ECCS analysis error, analyses were available and compensating actions could be and were taken to maintain the required level of safety.

D Why is the reliability of power included as a consideration in the ECCS exemptions?

Cur regulation: (10 CFR 50.12) provide that the public interest be served if an exception is to be issued.

0-Small Break ECCS Error Exemotions (Orders)

)< cl Why was Exemption fer TMI-l issued after the accident at TMI-2?

The Exemption for TMI-1 was issued prior to the accident at TM1-2, March 16, 1979.

It was published in Federal Register on March 30, 1979, two days after the accident.

7.b Does the deficiency covered in the Exemotion relate to the THI-2

, accident in any way?

If not, why not?

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Not directly.

Although the Exemption relates to allowing operator action to adjust high pressure injection system ficw, the event is a small pipe break in a particular location of a particular section of the primary coolant system.

The adjustment is for HPI flow out of the break.

The high pressure injection system was actuated in te TMI-2 event and the operator stopped HPI.

However, the action of the type specifically called for in the Exemption was not required.

% C Was there an Exemotion on this subject for TMI-9?

Yes, although it was in the form of an Order for Modification of License.

/.d-Do all operating B&W type plants _have a similar orcblom?

All except Davis Passe which, b'.cause of its raised loop configura-tion, does net ha"e this problem.

[.0 What is the status of the corrective action for all B&W tyce olants?_,

Arkansas Unit 1 - approved modification will be completed prior to startup frcm current refueling outage.

Crystal River 3 - Modification under review - Startup from refueling is mid-June 1979.

Schedule for completion not firm yet.

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Oconee 1/2/3 - Modification approved - modifications to be completed at next refueling -

Unit 3 - July 1979 Unit 2 - first suitable outage after December 13, 1978, Unit 1 - no later than Fall 1979 reload.

Rancho Seco - Modification approved - to be installed next refueling.

Three Mile Island 1 - Modification approved - to be installed no later than next refuelic] (April 1980) or first outage after September 30, 1979 projected te last at least 30 days.

Three Mile Island 2 - Staff has reviewed modification - modification will be completed prior to resumption at operation.

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DAVIS BESSE ACRS CONCERNS is 1.

What is te disposition of the ACRS concerns for Davis Besse l?

See letter from Denton to Ahearne, dated December 20, 1978.

DAVIS BESSE 1 ACRS CONCERNS 2.

Seismic Reevaluations atWhen is the first refueling outage scheduled?

March 1980.

If the seismic reanalysis indicates some deficiency in strength of some systems, s' 7 port on equipment what will the NRC require?

Modification of facility systems to meet acceptance criteria.

6 Is the seismic reevaluation at Davis Besse 1 the same problem as the seismic evaluation proolem oT the rive plancs walca aere soueuvoiJ No, the problem at Davis-Besse relates to the acceleration value which was used in the analysis and the method of correlating earthquakes with the values of applied acceleration.

The problems at the five plants relate to an error in a particular computer code which was used to analyze those five plant's systems.

c When will the guidelines be transmitted to the licensee?

They were transmitted January 30, 1979.

ECCS en the cer<

,7+n-ice",nno af 3.

Why dces the staff recuire more an=1veic a licensee? Why didn't the staff wait to issue the license until after it had acceoted the ECCS analvsis in total?

The kind of analysis required was only to quantify the margins of previous evaluations which show that the proposed linear heat generation limits are well in compliance with the regulations.

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4.

State of Ohio et Would this imply that there is no monitorinc crocram for radiolecical reliances from the Davis Besse Plant?

No.

The licensee is required by the Environmental Technical Specifications to have_a monitoring program.

6 What is the function of the State in this recard?

To confirm the licensee's program or as an audit of the licensee's program.

2.b

Q.

What is the status of environmental qualification considerations of the TMI-2 facility?

B.

Much of the equipment currently being relied upon in the present (418) mode of operation, (forced circulation of the RCS and heat rejection to the main condenser through a steam gererator),

are not safety grade equipment and therefore not covered by NRC requirements. All instruments important to the present mode of operation, i.e., RCS parameters, are located in the annulus between the shielding wall and the containment wall.

These instruments, based on available infonnation, wi}l function with doses up to 10' Rad.

Estimated dose rate is 10 rad / hrs.

Therefore, this dose will be reached in approximately 40 days.

Other equipment in the containment is gengrally gstimated to be able to function with doses between 10 to 10 Rad.

The dose rate in the containment is estimated to range from 800 to 2000 Rad /hr. A critical component is the surge capacitor on the reactor coolant pump motor, that is variously estimated to have a life of 18 to 125 days.

Water level in the containment is estimated to be over there above the floor. Therefore a steam generator level and RC flow instrument may be submerged.

Nine other instruments, including pressurizer level, are mounted 6 to 12 inches higher and could be near the water level.

Additional details are provided in the attached memorandum.

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UNITED STA TES NUCLEAR REGULATORY COMMiss!ON 3y f'

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WASHINGTCN. O. c. 2csss

(?$N?Lj' E gcf A?n. - 7 1979 MEMOP.ANDUM FOR: Brian Grimes, Assistant Director for Engineering Projects, DCR S. H. Hanauer, Assistant Director for Plant Systems, DS.h..b, THRU:

[

M [I FROM:

R. M. Satterfield, Chief, Instrumentation and Control Systems Branch, DSS

SUBJECT:

SUMMARY

OF RECENT STAFF EFFORTS TO ASSESS ENVIRONMENTAL QUALIFICATION OF SAFETY RELATED ELECTRICAL EQUIPMENT INSTALLED AT TMI-2 In response to the recent event at TMI-2, an effort was initiated by DSS arid 00R personnel to establish the radiation withstand capability of specific pieces of safety related el ctrical equipment installed at TMI-2.

This infonna-tion, together with criculated radiation levels inside of containment, was to be used to assess the possibility of failure of this equipment due to radiation expos 2re as a function of time. This memorandum summarizes the status of this effort.

Work thus far indicates that time of failure is diff. cult to establish because of uncertainties in the calculated dose rates at various locations inside containment.

Better predictions would be possible only when better data become available on which to base dose calculations.

1.

_Cualification Data for Selected Safety Related Electrical Ecuictent Installed AT TM1-2 1.

Instrumentation Related to Ccolant Temperature, Pressure and Flow Measurements Table I lists those pressure and differential pressure transmitters and temperature sensors located inside containment which are considered important to the current mode of operation (Reactor coolant pump operating and providing

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v; ficw through one steam generator.)

This table alsc presents the elevation of the transmitters above the containment floor (282' 6" level).

All instru-ments are located in the annular region between the shielding wall and the containment wall. The table also shows the radiation dose for which the transmitter was designed and/or tested.

In an attempt to gather more infomation on those transmitters suspected to be particularly susceptable to radiation damage (i.e. both the Bailey and Foxboro transmitters), DOE-Naval Reactors and Sandia Laboratories were asked to search their files for radiation test data on these instruments.

While neither organization had made use of the specific mode' instru-ments *f concern, Sandia Laboratories discovered through discussions with the equipment manufacturer that certain Foxboro pressure transmitters installed at TMI-2 are equipped with radiation hardened amplifiers and 8

that these transmitters have survived doses as high as 2 x 10 rads.

These instruments are used to measure reactor coolant pump seal cavity pressure at IMI-2.

(The identification numbers for these instruments are RC22-PT4 through -PT8).

Should existing reactor coolant pressure transmitters fail, these hardened transmitters may continue to be available to measure reactor coolant pressure.

1.2 Allis Chalmers Reactor Coolant Pump Motors B&W has established that the surge capacitors are radiation sensive.

They estimate that capacitor failure is expected to occur between dose levels 6

6 of 3.1x10 and 4.2 x 10 Rads. The mode of failure is thermal runaway.

This N

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will lead to short-circuitry of the capacitor which will trip the supply circuit breaker.

If the capacitor remains short-circuited.

the motor cannot be restarted or operated.

If the capacitor shculd open-circuit, the motor can be started and operated without it.

Therefore, it appears worthwhile to attempt a restart in the hope that either the capacitor has opened up or that the startir.g pump will open it up.

1.3 Pressurizer Vent, Block and Relief Valves With regard to the block and relief valves, B&W estimates that the design 0

radiation level is 2 x 10 rads. This value was based on tests performed on similar valves.

The pressurizer vent valves were supplied by Burns and Roe and that organization has been contacted to establish the appropriate design radi-ation level for those valves.

1.4 36" Recombiner Isolation Valves 7

These valves were supposedly qualified to 2 x 10 Rads. However, on April 6,1979, Region III notified NRR that the valve manufacturer had filed a Part 21 notification indicating that at least one component of 5

the valve had been tested only to 4 x 10 Rads. As noted belcu, the effect of this reduced design radiation level has yet to be assessed.

This issue is being pursued with the manufacturer.

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1.5 Fan Cooler Motors O

It was established that the fan cooler motors survived exposure to 10' Rads.

1.6 Decay Heat Removal System Components Components of the decay heat removal system which may be affected by radiation have been identified. However, design radiation levels have yet to be established.

In view of current plans not to use the DHR system, there are no efforts underway to develop this information.

2.

Estimates of Time of Failure of Safety Related Eauicment Due to Radiation Excosure In an attempt to predict when critical electrical equipment installed at TMI-2 might fail due to radiation exposure, dose rates calculated for various points inside containment were used in conjunction with the design radiation levels described in the previous section to establish equipment lifetime.

The dose rate calculations were based on analyses of containment air and coolant samples taken on March 31, 1979.

(See Enclosure 2)

It was assumed that the air sample was representative of the average containment atmosphere and that the coolant sample represented the liquid on the containment floor.

It was further assumed that the contributions to the calculated dose d"e to radiation frca the reactor vessel, steam generators and pressurizer were neg.1gible, because of the shielding between these components and the equipment of concern.

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' This latter assumption was supported by calculations which showed that little of the radiation produced by ccolant within these vessels penetrated the vessel walls.

2.1 CRNL Estimate of Radiation Dose at Location of Foxboro and Bailey Transmitters.

ORNL, using a two dimensional computer calculation, estimated the dose rate delivered to those Foxboro and Bailey transmitters that are a part of the reactor coolant pressure and ficw measure system.

(See Table 1) These transmitters are located in the annular region between the containment wall and the shielding wall. They are installed on an instrument rack and are positioned several feet (See Table 1) above the containment floor. The dose to these instruments is due to radiation from both the containment atmosphere and the contaminated liquid covering the containment floor.

One cause for inaccuracy of the dose rate estimates is the assumption that the coolant sample taken on March 31 is respresentative of the water on the containment floor. This sample was taken several days after the event whereas a significant portion of the water on the containment floor was deposited during the early stages of the event and should, therefore, be much less contaminated than the Maruh 31 sample.

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i From these calculations it was estimated that the transmitters are exposed 4

to a dose rate of 1 x 10 Rads /hr. The Bailey Model SY pressure transmitter 4

was tested to withstand 5 x 10 Rads with no effect on output and 4 x 10 Rad with slight shift in calibration of 4f. (BAW 10003).

Several of these trans-mitters apparently continue to function nomally.

It is therefore assumed 4

that their capability above 5 x 10 Rad is being used.

2.2 Calculated Doses to Other Ccmponents Inside Containment Other calculations were perfomed by the NRC staff to estimate the dose received by certain other important components inside containment.

In these calculations it was assumed that the only l

isotope of importance in the containment atmosphere 's Xe The l

analysis of the containment air sample showed Xe to be present in the concentration of 675 microcuries/cm3 (Table II). To be conservative l

and to account for additional Xe being vented into the containment from the pressurizer as a part of the degassing of the primary coolant, l33 the calculations were performed assuming a Xe concentration of 1000 microcuries/cc.

l The immersion dose due to Xe was estimated as a function of room size.

The roca enclosure was assumed to be spherical and to be bounded by a significant amount of shielding for the soft gamma (0.03 Mev) associated l33 with Xe decay. Dose rates at various locations in the containment

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were approximated by estimating a " characteristic radius" for the enclosed area of interest using the best available building drawings.

l 131 In addition to this air dose from Xe

, the dose from 1 dissolved in the water on the containment floor sas also calculated where applicable.

This calculation was performed using the concentration measured in the primary coolant sample (Jable II). These calculations are entirely dependent on the radioactivity concentrations assumed in the air and water. Uncertainties in these values would introduce corresponding uncertainties into the calculational results.

These calculations produce radiation dose rates as follows:

Various valves located on top of the pressurizer 800 Rad /hr Reactor Coolant pump surge capacitors 1100 Rad /hr*

Electrical penetrations 1500 Rad /hr Based on current knowledge of the design radiation level for pressurizer 5

block and relief valves, the lifetime is estimated to be 2.5 x 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

The surge capacitors are calculated to last 3000 hrs. The design radiation level for the penetrations has not yet been established.

An estimate was also made of the dose delivered to the 36" Recombiner Isolation Valves.

In this calcu'aion it was assumed that the major damage mechanism was exposure of the valve seat due to the ficw of containment

S&W estimated that the intggrated dose to the sarge capacitors as of April 4, 1979 was 1.8 x 10 Rads.

They estimate that thermal runaway could begin 13 days after the event.

The B&W dose calculation was based on consideration of contributions from both containment atmosphere and liquid at the bottom of the containment. The NRC calculations were based or the assumption that the capacitors would be shielded from the liquid at the bottcm of containment.

The difference between the NRC and B&N calculations will be investigated further.

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gas passing through the valve.

It is believed that the metal exterior of the valve will serve to shield valve ccaponents from the radiation from the con-tainment atmosphere surrounding tne valve.

The radiation dose to the valve seat is estimated to be about 2000 Rad /hr with most of this dose due to beta rays. The staff estimated the radiation 6

level capability for the valve seat mattr'al for this valve to be about 10 Rads. Therefo,e, the seat material should survive at least 21 days.

Assuming the seat is thick relative to the cenetration capability of the beta rays, the integrity of the seat would be expected to be maintained substantially beyond that time.

As noted above, the valve manufacturer has indicated that certain valve components, particularly organic material used in a solenoid valve mounted 5

on the isolation valve, were tested only to 4 x 10 Rads. The best information now available indicates that failure of the solenoid valve would prevent closure of the isolation valve. However, without more detailed knowledge of the solenoid valve design and the shielding afforded the organic material by the valve body, it is impossible to establish whether, and at what level, radiation will cause failure of the solenoid valve.

The manufacturer will provide appropriate drawings shortly.

j R'., h(. Sa tterfi el dQgp a, d. ;

m Chief, Instrumentation and Control Systems Branch, DSS cc:

V. Stello G. Lainas R. Mattson X. Kniel 26}3()

D. Eisenhut R. M. Satterfield F. Schroeder D. Tandi R. Tedesco C. F. Miller 7}

}()O S. H. Hanauer M. Srinivasan

TABLE 1 B&W Instruments Inside Containment IA - INST MOUNTED BY ITSELF IR - I.1ST MOUNTED ON RACK Parameter Inst Ident #

Mounting Tyce Rad Level Above 232.5' Design Test Level 7

SG "B" PRESS SPGB-PTl IM 13 Fox EllGH 10 2' 5" RC Flow LpA RC14A-DPT-3&4 IM-14&l5 Bailey BY 10 3' 0" 7

PRZ LEVEL RC.1-LT 1, 2&3 IR-424&425 Bailey BY 10 3' 6" SG "5" PRESS SP 68-PT2 IR-428 Fox EllGH 10 RC FLOW LpA RC14A-DPTl&2 IR-425&426 Bailey BY 10 RC FLOW LpB RCl48-DPT3&4 IR-429&430 Bailey BY 10 7

RC FLOW LpB RCl43-DPil&2 IM-12&l3 Bailey SY 10 l

l 7

SC "A" LEVEL SPlA-LT2&3 IR-426 Bailey BY 10 7

SG "A" LEVEL SPlA LT4&5 IR-426 Bailey BY 10 (SU) 7 i

SG ' B" LEVEL SPlB LT1, 2&3 IR-428 Bailey BY 10 7

SG "B" LEVEL SPlB LT4&5 IR-428 Bailey BY 10 V

(SU)

SGA LEVEL SPlA LTl IR-426 Bailey BY 10 5' 2" 7

SG "A" PRESS SP6A PTl&2 IR-426&424 Fox EllGH 10 7

RC PRESS (WR) RC3A PT3&4 IR425&427 Fox EllGH 10 7

RC PRESS (NR) RC3A PT5 ira 24 Fox E11GH 10 7

RC PRESS (WR)

RC3B PT3 IR429 Fox EllGH 10 V

RC TEMP (NRT ) RC5A TE2&4 Rosemount 177Y 10" Q

c (A LCCP)

RC TEMP (NRT ) RC58 TE2&4 0

c Rosemcunt 177Y 10 (B LOOP)

RC TEMP (WRT ) RC15A TEl h

(A LOOP)

\\gj RC TEMP (WRTH (A LOOP)

__.RC__ TEMP (WRT,,) RCISB TE1

.+

TABLE 1 (Cor.t'd)

Rad Level Above 282.5 Parameter Inst Ident #

Mounting Type Desicn Te3t Level RC TEMP (WRT ) RCISB TE2&3 Rosemount 177Y (B LOOP) c PRZ TEMP RC2 TE 1&2 RCPlA SEAL PU10-Fil EROCKS RET FLOW RCPlB SEAL MU10-FT3 BRC0KS RET FLOW RCP2A SEAL MUiO-FT2 BROOKS RET FLOW RCP2B SEAL MU10-FT4 BROOKS RET FLOW

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TABLE 1 (Cont'd)

EQUIPMENT INSIDE CONTAINMENT NECESSARY FOR CONTINUED OPERATION IN PRESENT MODE Design or Elevation Comoonent Identification No.

Type Test & Level Feet Pressurizer Vent Valve (Mov) RC-Vll7 352' 8

Block Valve (Mov) RC-V2 2.04 x 10 355' Relief Valve (SV) RC-R2 355' Spray Valve 355' Heaters 321' 9

Reactor Building Air Cooling Units 1 x 10 Containment Isolation Valves on Purge System In Core Thermocouples Reactor Coolant Pump Motor and Surge Capacitors

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A TABLE 2 Results of Analyses of Containment Air and Coolant Samples Taken l' arch 31, 1979 Constituents of Contaminated Coolant I

4 I

- 1.3 x 10 microcuries/cc l3 3

I

- 6.5 x 10 l36 2

Cs

- 1.8 x 10 137 2

Cs

- 2.8 x 10 140 Barium

- 200 Constituents of Contaminated Containment Atmosphere l33 Xe 675 microcuries/cc Xe 15 135 X

8.1 II I

.063 26 134

G', Q. Have Feedwater transients occurred at other B&W facilities.

How many? Significant?

A.

Feedwater related and similar transients have occurred at TMI-2 and other plants with B&W reactors.

These are described in the attached memorandum.

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ML19220C489

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