ML19207A575

Responds to Request by Presidents Commission on Tmi. Forwards Internal Memos Describing Involvement in TMI Activities

ML19207A575
Person / Time
Site:	Crane
Issue date:	06/06/1979
From:	Bauer R ENERGY, DEPT. OF
To:	Ferguson R ENERGY, DEPT. OF
References
NUDOCS 7908210121
Download: ML19207A575 (16)

Text

.

G P10RGRENrM-De aartment of Energy Ch cago Operations and Regional Office 9800 South Cass Avenue Argonne, Illinois 60439 JUN 6 1979 Robert L. Ferguson, Program Director for Nuclear Energy Office of Nuclear Energy Programs, HQ INFORMATION REQUESTED BY THE PRESIDENT'S COMMITTEE ON THREE MILE ISLA'ID Attached to this memorandum are several internal memorandums from Argonne National Laboratory (AUL), which describe the input personnel had in activitics at Three Mile Island. As discussed between Mr. Feinroth and Iir. Jascewsky, of my office, by telephone, the ANL's Radiological Assistance Teau's response activities are not included since this information has already been provided to Headquarters.

Any questions concerning the materia'. provided, should be directed to Edward J. Jasceusky on (FTS) 972-2253.

Original Signed by Fred C. MattuteMet' Robert H. Bauer OES;EJJ Manader/ Regional Representat'inc 1.nclosures:

1.

Memo., Rent to Honekenp, cated 4/l.':/75 2.

Mem., Gehl to Weeks, dated 5/29/79 J.

liemo., Korbus to De.Lorenzo, date.1 3/;'.;/ 7) 4.

Memo., Frost to File, dated 5/30/7) 5.

Memo., Frost to rile, dctcu 5/31/79 6.

Memo., frost to DeLorenzo, duted 5/31/79 7.

Memo., McConnell to DeLoren..o, dated 5/31/79 8.

Memo, Cunningham to Burris, dated 6/1/79 7dO.D3 7908210IM 9

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NATIONAL

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[Q 3 U.NLR,A,.L3BOR AJOR MEMO May 29, 1979 f.'. Zl

?1U3 office o..,... - ~ n. d et y File Action T0:

C. A. DeLorenzo Director, 00S

(([

Director, Plant Systems FROM:

H. M. Korbur

SUBJECT:

Plant Systems-Reclamation Activities at Three Mile Island

REFERENCE:

Memo C. A. DeLorenzo to Distribution, subject Information Requested by the President's Commission on Three Mile Is'and Warren J. Tyrrell, Plant Systems Supervisor engaged in activities at the subjr.ct area from April 16, 1979 through April 27, 1979.

Mr. Tyrrell was assigned to the Waste Management Group under the directica of Mr. Benjamin Rusche.

His assignments were as follows:

1.

Inspection of decontaminated areas by Westingbouse and Vychem decon personal.

2.

Reviewed report on protective coatings received by Mr. Rusche from company in New Orleans. Arranged to have company come to TMI for meeting and test demonstration.

3.

Consulted with Battelle NW representatives regarding electropolishing as a method of removing radioactive contamination fron metal surfaces.

As a result of this meeting TMI obtained funding from DOE for the installation of a decon facility estimated to cost 300K.

4 Advised Health Physics as to best method ror the decontamination and disinfecting of respirators.

5.

Discussed ANL waste disposal procedures.

6.

Met with local metal fabricator to prepare estimate for fabrication of the M III bin.

7.

Worked with Health Physics regarding survey and smear results.

8.

Worked with the Waste Management Technical Support Organization Group - chart attached.

HMK:hpb Attachment cc:

J. F. Bartusek w/att.

gjg.g(<3t3 R. L. Vree 306.RF PS File

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Ext. 251, 252 o

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WASTE MANAGEMENT RADWASTE OP ORGANIZATION 1C B. Rusche (SC-ERI)

J. Seelinger 3

(Met-Ed) i%

Planning / Scheduling H. P. Contacts H. Rininger (Het-Ed)

B. Brannock

' Personnel Goordinatic i T. Hillard (Met-Ed)

B. Gunderson (GPU)

D. Fick (GPU) p_____._._._________

Construction Technical Coordination Gas Liquid Support W. Hirst (CPU)

J. McConnell (CPU)

R. L. Williams (GPU) t W. Tyrrell ( ANL)

R. Brooksbank/OR11L Lia.

I D. Hagle (UE&C)

R. Dunn (Gilbert)

M. Ross (GPU)

J. Snider (Liquid) (ORNL)

}

S. Dam (B&E)

J. Robison (Gilbert)

S. Kraft (GPU)

W. Rodger (NSA) i W. Gunn (CPU)

W. Itschner (GPU)

D. Carman (Gilbert)

J. Lieberman (USA)

(

S. Levin (CPU)

~ C. Hontgomery (CPU)

A. Larson (Gilbert)

L. King (ORNL)

J. Kindzierski (CPU)

A. Hawaz (Bechtel)

C. Edwards (Gilberts)

W. A. Shannon (ORNL)

T. Uilkema (CPU)

B. Lutz (Gilbert)

Martin (ORNL)

A. Peters (B&R)

P. Arthur (Gilbert)

B. Reinmann (GPU)

Clerical Support.

B. McCutcheon (GPU)

J. Smith (GPU)

~

K. Pastor (CPU)

J. McDonal (CPU)

B. Turpin (Duke)

Clerical Support M. Babb (Met-Ed) i B. Gunderson 7,.

4/19/79 s

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ARGONNE NATIONAL LABORATORY INTR A.L ABOR ATORY MEMO 30 May 1979 TO:

File FROM:

B. R. T. Frost, Director, MSD T. F. Kassner (ANL-MSD) participated in a meeting of nuclear fuel experts, held on April 12, 1979 at NRC-Bethesda, to update estimates of the damage to the TMI-2 core and to consider its effect on the desirability of initiating natural-convection cooling of the core. ANL-MSD has been involved for the past three years in an NRC-sponsored program to investigate the effect of steam oxida-tion on the mechanical properties and extent of embrittlement of Zircaloy clad-ding under loss-of-coolant accident conditions. A summary of the above meeting was transmitted to D. Ross, Deputy Director, Division of Project Management, Office of Nuclear Reactor Regulation by W. V. Johnston, Chief, Fuels Behavior Research Branch, Division of Reactor Safety Research, Office of Nuclear Regu-latory Research.

There have been subsequent telephone calls with NRC-RSR staff, particularly M. Pickelsheimer (Zircaloy expert and program manager for LWR cladding).

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ARGONNE NATIONAL LABORATORY INTR A.L ABOR ATORY MEMO 31 May 1979 TO:

File FROM:

B. R. T. Frost, Director, MSD

SUBJECT:

Assistance to EPRI by MSD Cer ;M_cs Croup At the request of Dr. Adrian Roberts (EPRI) Roger Poeppel of MSD conducted a quenching experiment on unirradiated UO; pellets - annealed in 100 CO2: 1 C0 for one hour at 1600*C and quenched into water. Examined pellet for cracking behavior - first time it held together in spite of cracking. Repeat of experi-ment at least twice and pellet fell apart each time into a few large pieces.

Transmitted results to Dr. Roberts on April 27.

Note:

Dr. Poeppel has been under contract to EPR; to study cracking behavior of UO2 for the past two years. The purpose of th m quench!rg experiments was probably to throw light on the likely size of UO2 particles in the upper section of the TMI-2 core.

ANL did not transmit this informat,ian to NRC.

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ARGONNE NATIONAL LABORATORY INTRA. LABORATORY MEMO 31 May 1979 TO:

C. A. DeLorenzo Director 00S FROM:

B. R. T. Frost Director, MSD

SUBJECT:

Information Relating to TMI In response to your memo of May 22. I enclose four memos that describe MSD involvements with Three Mile Island. This does not give you the information in the format that you requested but I have only just received these memos and rather than rewriting them to meet your format I am sending them as written.

In addition, Dick Weeks has accompanied John Honeka=p and others to California this week to talk to EPRI and Prof. Pigford of the Commission on key questions to be answered and what assistance ANL can provide to these two studies.

I suggest that you contact John Honekamp for information when he is at the Laboratory next Monday, June 4.

BRTF:rv ce cf Nb,. G.K c.

s ribution:

R. W. Weeks L. A. Neimark g,-l :-

'7g S. Greenberg J. Rest l7 R. B. Poeppel lNFO t A1T:O" T. F. Kassner CAD J. R. Honehamp EAW FOP !

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.ARGONNE Omcccf0;n9.:-c;L;;:i NATIONAL LABORATORY IN" A. LABOR ATORK ' MEMO

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i ADE 300 ' INFO I A",T.'"1 CID 0 May 31, 1979 EAE[1 FOP r-HB RGT TO:

C. A. DeLorenzo Director, 00S

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R. J. McConnell < [,.

Associate Director,i BE-ItiEngineer' ri'g i

T FROM:

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SUBJECT:

Three-mile Island

_..i REF:

Memo to Distribution from C. A. DeLorenzo, "Inforrra t&re ues ten by the President's Comission on Three Mile Island," dated May 22, 1979 In response to your reference request, I participated in the following meeting.

Subject TMI-2 Meeting - Core Danage Assessment, April 5,1979.

Location Lynchburg, Va.

Participant R. J. McConnell, ANL/EBR-II Purpose

1) Evaluate the information available to determine a best guess of the extent and nature of core damage.

2) Propose future efforts to gain more knowledge.

Other Participants See attached.

To the best of our knowledge, this represents the only EBR-II participation.

RJM:km cc:

D. W. Cissel 79&1.39

ARGONNE NATIONAL LABORATORY INTR A.L ADOR AlORY MEMO April 12, 1979 To:

J. R. Honekamp h

OTD

^

.y From:

J. Rest HSD j

Subject:

Assistance to the NRC on Hatters Related to the Accident at the Three Mile Island Nuclear Reactor On April 9 Rich Sherry of the NRC called me to request assistance in estimating the fuel temperatures for the uncovered core that occurred as a result of the accident at the Three Mile Island Nuclear Reactor #2 on March 28, 1979.

Rich said that George Marino of the NRC had done some calcula-tions to estimate the rate of heating of the uncovered core due to decay heat assuming adiabatic conditions.

George's calculations indicated heating rates of from 1-10* F/S.

(The next day George called to revise the heatipg rates down to 1-3* F/S).

In addition, measurements had been made (how the measurements were performed and vhen they were made was not clear) on the amount of released isotopes of xenon.

The measurements indicated that 22%, 24%, and 31-38% of the total core inventory of X 135, y 133m, and X,133, respectively, had been e

e released. The NRC requested that I use the CRASS-SST code to perform calcula-tions for assumed accident scenarios to estimate the fuel temperatures required to explain the release of from 20-40% of the total core inventory of fission gas.

Two suggested scenarios were used to perform the calculations.

The first scenario consisted of an irradiation at 6 W/f t (average core power rating) an average fuel temperature of 1200*F for 60 full power days followed by at (1) a relatively instantaneous reduction in power (1.2% of nominal due to reactor scram) and a fuel cooldown to 550*F occurring in about I hour's time, and (2) a heatup of the fuel at a heating rate of 1* F/S.

The second scenario uas similar to the above, but differed in that fuel from approximately 25%

of the core which was irradiated for 60 full power days at about 12 W/f t with an average temperature of 2550*F was considered.

The objective of these calculations was to determine at what fuel temperatures GRASS-SST would predict 20-40% total gas rele.,se.

The result.s of the analysis, which were transmitted vy bally by me to the IGC (George Marino) on April ll, subsequent to separate d cussions I had with Steve Gehl and yourself, are as follows:

1.

Predictions made with GRASS-SST for the 6 W/f fuel indicate that a maximum of from 5-10% total gas releas would occur at fuel temperatures between 4700*F and the fu melting point (s5160*F).

This result was obtained assur ng that no extensive grain-boundary separation occurs in the uel.

A substantial amount of gas release from the grains as calculated to occur as a result of the heatup, but this s was trapped on the grain surfaces and edges, and hence was ot released to the exterior

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J. R. Honekamp April 12, 1979 of the fuel.

(The predicted fuel swelling due to the retained gas on the grain surfaces and edges was too small to cause appreciable long-range interlinkage of the porosity).

2.

For the 12 KW/f t. fuel 20% and 40% total gas release was predicted to occur at fuel temperatures on the order of 4500* and 4800*F, respectively. Again, this result was obtained assuming that exten-sive grain-boundary separation did not occur.

3.

Assuming that extensive grain-boundary separation does occur, CRASS-SST results indicate that for the 6 KW/ft fuel 50% or more total gas release could occur at fuel temperatures on the order of 2400-1700*F. Results for the 12 KW/ft fuel indicate that 50%

ss release could occur under these conditions at fuel temperatures on.the order of 4500*F.

The calculations indicated that the substan-tially greater fractional release of fission gas predicted to be released from the lower rating fuel than from the higher rating fuel at temperatures of about 2400-2700*F in the event of extensive grain-boundary separation was due to the fact that the lower opera-ting te=peratures of the lower rating fuel resulted in much smaller (and hence more mobile) bubbles being generated within its grains.

4.

Experience with fuel from the H. B. Robinson Reactor (30,000 mwd / con burnup compared to the %1000 mwd / ton burnup of the Three Mile Island fuel) during Direct Electrical Heating (DEH) transient tests at heating rates substantially higher than those calculated to occur at Three Mile Island indicates that %20% and 40% release occurs in fuel regions where the temperatures reached 2750 and 3650*F', respectively.

However, it is expected that the much higher concentration of gas in the H. E. Robinson Fuel (very little gas was released during the irra-diation) facilicated the formation of the observed grain-surface and grain-edge channels, and this enabled the gas released from the grains to escape to the exterior of the fuel.

In addition, fairly extensive grain-boundary separation was observed to occur in this fuel as a result of the transient heating.

On April 12, I called George Marino to ask him if the NRC needed any additiona:

asnictance in the interpretation of the above results, which I had transmitted to him on April 11.

I again reiterated that these results were based on assu=ptions about the properties of the fuel (e.g., grain-size), assumptions on the irradiation history of the fuel rods, and on the suggested scenarios of the accident.

For example, from the above results it is clear that the irradiation history of the fuel (e.g., 6 KW/f t vs. 12 KW/ft) significantly affects the predicted gas release during the accident.

George told me that the NRC feels that significant grain-boundary separation could have occurred during the accident.

Their thinking is based on the results of a pressure transient in the PBF reactor where appreciabl.e grain-boundary separation was abserved.

In particular, George was referr'ing to a region of the

'6 YANb.

J. R. HonekamP April 12, 1979 PBF-tested fuel where almost complete separation (powdering of the fuel) was observed. Analyses indicated that this type of separation (differing by its powdery nature from the type of separations observed in DEH-tested fuel) was the result of the stresses generated as a re ult of requenching.

Based on this hypothesis, the NRC is using in their calculations, the 2700*F fue2 temperature predicted to occur in the low rating fuel during the accident under the assumption of the formation of extensive grain-boundary separation.

(Result #3, above).

George was reluce. ant to provide' me with any information on what the NRC calculations were to be used for (e.g., the calculation of metal-water reaction rates?), only saying that the work was of an urgent nature.

Oeorge and I both agreed that more detailed calculations (e.g., calculations with GRI.SS-SST and LIFE-LWR) could be useful.

I indicated that I would attempt these calculations in the near future.

Note (added later):

J. Rest is continuing telephone discussions with NRC-RSR personnel on fission gas release calculations.

O

/AF9C3C)t4 t4 EE NATIONAL i

LABORATORY INTRA. LABOR ATORY MEMO 29 May 1979 TO:

R. W. Weeks, Associate Director, MSD FROM:

Steve Gehl, Irradiation Performance Group, MSD RE:

Assistance to NRC, etc. regarding TMI As the following list indicates, most of the assistance I have provided has been indirect, e.g., with Honekamp and Rest.

I have included all informa-tion for completeness, so that you can do the editing.

1.

Week of 4/2. Discussion with J. Honekamp on the likely fission-gas be-havior in IMI prior to accident (how much fission gas release before accident?)

Provided literature references.

2.

Week of 4/9. Several discussions with J. Rest regard *.ag all aspects of fission-gas behavior prior to and during accident. Discursed the impl.dcations of GRASS calculations, which were then transmitted to NRC (C. Marino).

3.

4/12. Phone call S. Gehl to G. Marino (NRC). Discussed temperature history during accident; role of fuel microcracking and chemical reactions in fission gas release. George asked that we consider running DEH (Direct Elac-trical Heating) tests in steam atmosphere.

4.'

Week of 4/9. Phone call from EPRI, referred-by Honekamp. Provided ref-erences on fission gas behavior in LWR fuels.

5.

5/10.

Internal Memo (Gehl to Neimark) " Core Thermal Conditions and Fission-gas Release in TMI-2" (can be provided if needed).

Note:

L. Neimark, S. Gehl and J. Rest have been carr.ng out a program to study fission gas release from UO2 under transient conditions for NRC-RSR (C. Marino) for the past three years.

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L. Burris Chemical Engineering FROM:

P. T. Cunningham Chemical Engineering /ACL

SUBJECT:

Assistance to NRC on the T.M.I. Accident Attached is a report describing our assistance to NRC on the T.M.I. accident, per my conversation with Don Webster on Wednesday.

P. T. Cunningham PTC/jes Attachment cc:

D. Webster F. Cafasso R. Meyer C. E. Johnson R. Heinrich R. Larsen (RES) 78E$3C[.i g

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Assistance to NRC on the T.M.I. Accident Type of Assistance: Review of experimental data resulting from y counting of a smnple of primary coolant water to determine what, if any, inferences might be drawn from the data.

Who Provided:

Data was reviewed by R. Larsen, Radiological and Environmental Research Division, and C. E. Johnson, R. Heinrich, and P. Cunningham, Chemi al Engineering Division, Argonne National Laboratory.

To Whom Provided: Conclusions were transmitted by telephone to Bud Cherry, Vice President, Corporation Planning, General Public Utilities Corporation.

When Provided:

Data received on Saturday, March 31, 1979, and conclusions transmitted by telephone on that afternoon.

Results: The data provided is summarized as follows. A 100-mL sample of primary coolant water was obtained and y counted to determine the fission products present. Counting results, obtained from Mr. D. Henderson (BAPL) were:

Isotope Decay Rate, d/m/mL*

1 131 1 3.0 x 10 0 132Te 4.5 x 108 1

133I 1.5 x 10 0 1

134Cs 1.4 x 108 136Cs 3.9 x 108 137Cs 6.1 x 108 140Ba 4.7 x 107 es.90Sr 1.2 x 107

• All count rates corrected to noon, 3/30/79.

The sample was reportedly reading about 1 R/mL and contained trace levels of U.

Other'information available included:

a reactor had approximately 60 full power days equivalent of operation; e reactor core 93 x 103 Kg; e primary coolant volume 4 x 103 ft ;

3 e gas (probably hydrogen) " bubble" approximately 1.2 x 103 ft3 at 1000 psig.

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2 Based on this data, and other reasonable assumptions or estimations as necessary, it was concluded that:

A.

About 3% of the core inventory of Cs was in the coolant (perhaps 25%

of the fuel was exposed to coolant).

B.

Fission product distribution was not consistent with that expected if fuel vaporization or meltdown had occurred; therefore, fuel probably still intact (at least as pellets).

(The concentration of 132Te, which, at the time of the accident, was comparable to that of 131I in the fuel, was about 102 less than that of 131I in the water.

Tellurium is a relatively volatile fission product.

The concentration of 140Ba, which, at the time of the accident, was also comparable to that of 131I in the fuel, was about a factor of 103 less than that of 131I in the water.

It is expected that 140Ba would be found in the water if fuel melting had occurred.)

C.

Hydrogen appeared to be the only reasonable major constituent for the bu bble. A water-metal reaction was postulated as the source for hydrogen and to achieve the stated volume would have required involve-ment of about 25% of the cladding inventory of Zr.

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Assistance to NRC on the T.M.I. Accident C

Type of Assistance: Data provided on the solubility of hydrogen in water at high pressures and temperatures.

I Who Provided:

D. S. Webster and M. Blander of Chemical Engineering Division, j

Argonne National Laboratory.

2 To Whom Provided: Joseph Murphy of the Nuclear Regulatory Commission p

(apparently part of a group assembled for the emergency).

q4 When Provided:

Initial telephone conversations on Sunday night, April 1.

H Data sought on Monday, calculations made and results telephoned on Tuesday q

(4/3).

Results: At the time of this inquiry by NRC, there was presumed to be a 1000 cu. ft. " bubble" of hydrogen at the top of the reactor, in solubility-4 equilibrium with recirculating water.

In order to dissipate the bubble, it U

was believed possible to draw off 20 gpm of water (through a cooler, let-down orifices, filters) and spray it into the makeup storage tank at about 1 atmosphere pressure. Dissolved hydrogen was expected to come out of solution, whereupon it would be discharged to the containment; the water would be pumped S

back into the reactor primary system. The questions were: What is the solubility of hydrogen at 1050 psia in water at 280 F, and how long will the H removal take?

2 The most probable value for the solubility of H2 at 1050 psia in water at

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280 F was judged to be 1600 ml H2 (S.T.P.)/kg water (Reactor Handbook, 2nd Ed., Vol.1, p. 851, Interscience Publishers, N.Y.C. ).

Furthermore, as the water temperature fell during passage through the cooler, the solubility would fall to about 1300 ml H (S.T.P.)/kg water at 120 F.

2 At the 1600 ml/kg value, 20 gpa of water (75 kg/ min) would carry 120 liters H

(S.T.P.)/ min, or 0.09 cu. ft./ min at 1050 psia and 280 F, to the disen-2 gagement vessel.

If it is assumed that the returned water saturates with a

l hydrogen again in the reactor, the 20 gpm purge will have to continue for about a week (_1000 cu. f t.

= 11000 min) to remove the 1000 cu. ft. bubble.

0.09 cu. ft./ min

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Af ter.the bubble is gone, purging should continue in order to remove most of in soluti n.

The relationship for this falling-the large amount of H2 j

concentration period is t = 3. In g9., where M is the mass of M

C primary system (taken as 230 000 kg), q is the pumping rate (75 kg/ min), and Co/c is the ratio of initial H2 concentration to the desired final concentra-tion (taken as 1600/10).

The time for this phase of purging is about 11 days.

disengages during spraying (?)

If, as suggested by NRC, only half the H2 both times will be doubled.

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