ML19323B540
| ML19323B540 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/04/1980 |
| From: | Disalvo R, Kelber C, Wood P NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
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| FOIA-80-301 NUDOCS 8005130503 | |
| Download: ML19323B540 (44) | |
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800513 0 DD3 A FURTHER EVALUATION OF THE RISK 0F RECRITICALITY AT TMI-2 R. DiSalvo C. N. Kelber P..M. Wood l
Office of Nuclear Regulatory Research United States Nuclear Regulatory Comission i
April 4., 1980 Os a
a en ABSTRACT This report reviews previous studies related to the probability and consequences of criticality at the damaged Three Mile Island Unit-2 reactor. !! ore detailed assessments are performed to ' confirm the adequacy of those studies rnd to provide additional insight into ways to minimize risk from criticality. The most important conclusions of this study are:
1.
The most probable mechanism for criticality, boron dilution, is a slow enough process that with appropriate instrumentation and procedures, the approach to criticality can be detected and corrected. To the extent that boron concentration in excess of 3500 ppm can be ensured, the probability of criticality is further minimized.
2.
The most likely direct radiological consequence of criticality is increased dose rates inside containment.
For the more realistic and more probable criticality events studied, off-site consequences are nonexistent. More conservative assumptions regarding the nature of the criticality, combined with multiple failures of engineered safety features are required before one calculates detectable health effects.
Even then, the consequences, as expressed in terms of the probability of latent cancer fatality, appear to be very small compared to the observed incidence of cancer death. To the extent that core g
cooling and containment integrity can be maintained, the consequences l
l of criticality can be further minimized.
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CONTENTS Page 1.
Introduction............................
1 2.
Current Status of Criticality Control at TMI-2...........
1 3.
Summary of Analyses to Date 2
3.1 Probability of Criticality..................
2 3.2 Consequences of Criticality..................
4 4.
Outstanding Questions and Further Evaluations 8
4.1 Probability of Criticality..................
8 I
4.2 Consequences of Criticality..
14 5.
Conclusions and Recommendations...... /..'.........
25 References 28 Appendix A: Summary of Criticality Analyses of TMI-2........... A-1 Appendix B: Low Baron Concentration at TMI-2............... B-1 1
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INTRODUCTION On February 28, 1980, a special task force formed by NRC's Acting Executive Director for Operations reported its findings regarding the cleanup activities at Three Mile Island.I Among its recommenda-tions were that the " staff reevaluate the potential for recriticality and ensure that adequate procedures and equipment are avail ble to prevent its occurrence."
On March 10, 1980, the Director, Probabilistic Analysis Staff of NRC's 2
Office of Nuclear Regulatory Research directed the authors to perform '
an independent assessment of the risk of criticality and to prepare this report. Specifically, we were to " review work already done on this matter by the Kemeny Commission staff, the Rogovin inquiry and NRR.
Consider the mechanisms by which boron could be lost from the core region so that recriticality might occur.
Evaluate the probability of criticality occurring, the rate at which criticality could be approached and the likely consequences of such an occurrence." Qur findings and recommenda-tions follow.
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2.
CURRENT STATUS OF CRITICALITY CONTROL AT TMI-2 As of March 31, 1980, the damaged core of TMI-2 is subcritical as verified by the single remaining excore source range neutron detector.
Believed to be maintaining subcriticality is a boron concentration (as boric acid) of 3850 parts per million (ppm) in the Reactor Coolant System (RCS) water. This is measured weekly at a location upstream from the letdown coolers approximately 200 feet from the core.
A technical specification lower limit of 3500 ppm boron has been estab-lished. There is essentially no flow of water in the RCS except during short periods associated with " burping" in the steam generators.
The pressure and average temperature in the RCS are about 280 psi and 150 F respectively.
The core is presumed to have been uncovered for up to two hours during the a'ccident of March 28, 1979.
A major portion of the Zircaloy was oxidized, and the fuel, control, and burnable poison rods experienced thermal transients beyond their design conditions. A cone of failed, oxidized fuel rods is believed to extend from the top of the core to eight feet downward.
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The control rods (silver-indium-cadium alloy) entered the core seconds into the event.
Their current status is uncertain, but melting of at least the top third should have occurred as a result of the thennal transient. 11uch of the control rod material may be retained in the outer and lower regions of the core.
Some of the boron in the fixed burnable poison rods (B C-A10 ) is probably lost since boron is known 4
23 to leach out when exposed to water in a radiation environment.
3.
SUMMARY
OF ANALYSES TO DATE 3.1 Probability of Recriticality Criticality analyses of the TMI-2 core have been made by the NRC staff (1,3,4,5)
BabcockandWilcock(6)
Brookhaven National N)
General Public Utilities (8,9) and Oak Ridge National,
Laboratory Laboratory (10)
The important models and results of these analyses are summarized in Appendix A.
A reevaluation of these analyses yields the following conclusions:
(i) - The keff of the core with 3500 ppm boron is conservatively estimated to be less than 0.90.
With at least 3500 ppm boron, the core will remain subcritical in any physically reasonable rearrangement of the fuel even in the total absence of control rods or burnable poison.t*
In a recent NRC memorandum, UI)
Marotta points out that the ORNL analysis (10) does not assume the most reactive core configuration given our current under-standing'of the core's physical condition. lie recommends using a higher reference keff (0.944 at 3000 ppm boron) for boron dilution studies. Using 100 ppm as equivalent to - 1% Ak/k, the higher reference keff yields k=0.894 at the technical specification lower limit of 3500 ppm boron.
Given the uncertainty regarding the status of control materials and burnable poisons, these analyses give no credit for their contribution to criticality control. As a result, the calculated concentrations of boron required to maintain subcriticality are overestimated, perhaps by as much as a factor of two.
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(ii) - The calculational methods and nuclear data used in the analyses are adequate. The methods used by NRC/NMSS(4),B&W(6), and BNL f ) have been tuned to experimental data through the years.
The ORNL analysis (10) includes calculations of the TMI-2 core at startup with all rods out, critical at zero power, and RCS conditions of 220 psi, 532 F and 1490 ppm boron. The ORNL calculation underpredicts criticality by about 1.5% ak/k.
This was taken into account when determining that 3500 ppm boron was adequate to prevent criticality.
(iii) - The potential for small unborated or underborated volumes of water to enter the core without the benefit of mixing must be considered as well as the more commonly addressed concern of a well-mixed gradual dilution. The affect of assuming zones of the core with lower boron concentrations was studied by Marotta b)
Introducing 1000 ppm borated water into the oute'r regions of the core would result in criticality.
Introducing a coherent mass of unborated water with 3
a volume of 3 ft into the core would also result in criticality.
This latter calculation is supported by data from the Westinghouse Reactor Evaluation Center in 1967.
(iv) - The analyses make no quantitative estimates of the probability of achieving the conditions necessary for criticality.
The major concern is the introduction of water with less than 3500 ppm into the core. The studies generally conclude that with appropriate precautions related to sampling and introducing water into the RCS, the approach to criticality is detectable and avoidable. Many recomendations designed to minimize the probability of criticality have been made with these thoughts in mind.
4 3.2 Consequ.ences of Criticality Thompson and BeckerlyO 2) have reviewed reactor accidents involving criticality or reactivity changes. A summary table from Reference 12 which includes total fissions and estimates of radiation dose is repro-duced here as Table I.
Except for NRX and SL-1, the events described resulted in little or no radiation dose.
These data must, however, be viewed cautiously if one wishes to extrapolate them to TMI.
Special consideration must be given to major differences in core design, in the initial configuration of the core, in the design and availability of engineered safety features and any other factors which are unique to TMI in its current configuration.
Another key reference in re ard to accidental criticality is the well known study by Stratton.03 The TMI-2 core is in the category of inhomogeneous water-moderated cores reviewed by Stratton. Two types of accidental criticality are reviewed. Accidents caused by the sudden insertion of reactivity (such as Bor,ax 1, the Sport tests, and SL-1)
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apppear to be limited by the rapid, almost adiabatic production of heat in the core. The power curve looks like a sharp peak.
Typically 18 5 x 10 fissions occur, corresponding to a production of 158 MJ.
For a large core, this might be an order of magnitude larger.
Some accidents have involved slow approaches to criticality in which the reactor does not go prompt critical.
One such example is the NRX accident of December 12, 1957.. Af ter the reactor attained criticality it would rise in power until either of two conditions was met:
(i) the reactor became unstable and eventually overheated through loss of cooling; or (ii) the available reactivity was used up and the reactor operated at a steady power.
An inexorable increase in reactivity through continued removal or boron would probably lead to unstable boiling since all but the more optimistic evaluations of reactivity indicate considerable po.ential for added insertion.
It is hardly conceivable that such an increase would occur except in the absence of all precautions plus deliberate dilution of the cooling water.
Nevertheless, the advantage of early warning of reactivity increase from neutron detectors as well as effective monitoring of boron concentration is that, even if dilution does occur, the reactivity increase can be stopped and reversed before unstable core performance leads to more fuel melting.
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TABLE I.
REACTOR ACCIDENTS INVOLVING CRITICAUTY OR REACTIVITY CHANGES '(FRO'i REFERENCE 12) j l F-1 T ui l Cocee (C)
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2 33 Sph ere, 5.4e1018 '(C) Meen1 withdre=el of 2 cen-P-It R-2.S r gemme None
'S u 3.2 Dec.1949 LASL 1 ha U83 8. UO tNO 1
( ster boiler) in 13.6 liter II 0 cephite l
2 trol rode crit.1944,1950 reHector i(Q) Empeaeios andriu of nutros l
tem peret.re 12 Dec.1952 Chest Riur, Cuade ! Natural useien red.
Red lettice.
0.6 e 10 (C) Cutrel red mal-operation, P-nou, ucept in Cue badly damesed. See. 3.3 23 NRX,cr tieelity 1947, ' 13 0-cooled, po phite setety ciremit feil.re-cieu p; many P remend. replaced 2
0 0-moderated reflected comptes got..11 aoen.
felt pc.er May 1918 2
l(Q) D.mp of D 0 moderater highus 17r, meet 2
'less thee 3.9r H O reflector, ' 4.68 = 10" ((C) Eetimate of espected P-n on Cue dutroyed See. 3.4 23s 22 Jely 1954 lNRU Idebe BOR AX I { 93% enriched U 2
(treeniest teste 1954) in U-Al plates NTR e.imming pool encersion 1 -
R-mone type),
es cursion
!(Q) Stum void eisenembly II 0-modereted rea ctor I
3 Oct.1954 'liesford Production j Netarel ernism rode Proceae tabe Leen j(C) Weier Iu's changed renter P-none Some ful element.
' Sec. 3.5 2
j pattue, ehort period eccurred R-oose failed er wue Reacter H 0-coled, type-Isrge ev n.
y First ou critical sr*Phite-mod"ated rePhite Antin6 4Q) Control rod chu6es thee damaged h
scrae
. September 1944 ree r;or t
.jFul
'(C) B;ocTege of cooling eter is 'P-oone Graphite chusel Sec. 3.5 4 Jan.1955 lien ford - K7
.... eame.
... same.
Rea ctor j f eilere, precen tebe - initial start.
3-none remond by hele l local up power decrease noted -
.(Q) Scram on eur-cet is shield
' meltie 6 1
roJo wi6 drawn l
pre neure Jan.1955 jllenford Production l... same...
{... esme.. iNo over-(C) M., estimate of p.instrsmente jPaone
} Nue Se c. 3.5 1Reettor l
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l power i(Q) Rod run is by operator in-sone i
U1 0
red e.
Compact core ; 4.7 = 10 (C) Estimate of espected res it.
P-non e Core melted, little Sec. 3.6 29 Nu.1955 lNRTS Idake EBR.I 0.5 in. U235 p'eri-!!
NeK. cooled fut Nat. U blanket j low - enriint scre's attempte R-minor
- . er conteniention
'Operatione le 1951 reactor I
not effective j
i (Q) Sauteff by second scram; i
ful 1,owina e factor e
9 Oct.1957 (Eastead Windecele
!Netur L uranium rode B-eided steek j Cre phite- !(C) 5.gur energy releue, U-iP-nose enieu i Sevue cue d. image ' See. 3.7
?Operetione is sir. cooled, of rephite
' renium boraieg tringered by necteer H.= id e e,we a d reactor not rel.Ut July 1950 graphite-a.eduated 50 50 = 25 ft ' fire ove rh e atin a re4ioacu=ite, milk 25 ft ful (Q) I'loodied "ith 31 0 ovu 200 m 2 ares 2
ch onnels l
Je e tra t e d I
18 Nov.1958 t NRU -IITRE.3 l Enriched aruism fiorian tal Notknow. '(C) System en auto-control with P-een.
Core mehed, basic Sec. 3.8 syetem undemaged.
crit. October 1958 au-ceeled, eelid cylindu fulty intrumentaties R-ume site
.i.-
j modnotee (Q) Mehde=n efemp ead/or serem contem;eaties
-4, 24 hly 1959,Saue Susene Cel.
2.8% U 38 elege in.
Pendo.
(2x10" I(C) Coolent cheneel blecinge by P-eene 12 of 43 elemente Sec. 3.9 3
- SRE crit.1957 SS-clad rode
- cylinder, (inlast impurities everheating.
R releen wee melted, core
!st pe=u May 1958 Ne-cooled F8Phite minste) perhope Isel bowing 0.3% of core remend ud replaced araphite-modereted re flector i
,(Q) Ven.nl scra:e
.actieity inventory i
3 April IMO felt: Wil, Pe.
93% enriched U23s Peendo.
Overhe et i(C) Undercooled, perhape faulty 1 element melted,
'Sec. 3.10 (P-none l
f.el-negatin esto control R-minor rolesee 810* to clean up ETR U-Al plates-cylinder
- cylinder, of one
- crit.1959 Il 0-cooled-modneted
=stu element rio site dames.
j ) responu I
2 JQ Vanual serem 6
re Gec te d 13P.all fatal e s troy ed.
.Sec. 3.11
!R > 800r/hr in bids ] Core 3 Jan.1961 ;NRTS 93% enriched Usis
' Pendo.
' t.5.10H !(C) Manuel -ithdre..! of centret
, ldQ) Espusion, boitias, core
, veneel rose 9 ft,
- SI -I Al U plates, boiling, cylinder control rod e recovery 14P got reactor diamentled Il 0 cooled-moderated 5 rode, B-Al
- crit. Anguet 1958 2
i
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5 Nov.1962
'N RM 93% enriched U23s IPseedo-IP-sone Core duireyed Sec. 3.12 l(C) Pinned upuimeet ructivity ISPERTI A1.U plate type
! cylinder with tres ient as p!.nned, energy IR-miner site (destrwetion plassed Il 0-modereted bras sient reinu effecte eiere destrec-l contamination on thie tut or aut)
(destructive tute) 2 rod open tiu thee planned I
'(Q) Espansion, etu weid, so f ni l
melting inutved te s
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. A best estimate of the stable power that can be reached at or near atmos-pheric pressure can be made by extrapolation from the past natural cir-culation boiling water experiments, such as EBWR. A 20% average void in the core corresponds to about 100% quality in the hot channel exit. A steam velocity of about 0.3 m/s is commonly observed, and we can conser-vatively estimate a mean bubble rise of 2 m.
An energy balance then yields an estimate of 2/3 Mw, or since this is an order of magnitude estimate, about 1 Mw. This is in accord with the experience cited by Thompson and Beckerly(I2) and by Stratton.(13)
At higher system pressure, higher powers can be attained.
As a best estimate,' assuming two-thirds of the control rod material is effective, there might be 5% excess reactivity if all boron were removed. Assuming 2% of ak for Doppler and temperature defect, this would yield about 12%
average void at high pressurd (2200 psi) or on the order of 100 MW.
At the current TMI-2 system pressure,(280 psi), the power level would be about 15 MW. This is quite approximate; each 1% Ak beyond the 2% ak to reach temperature represents about 30 MW.
Appendix 3 in NRC's Task Force Report " Evaluation of the Cleanup Activities at Three Mile Island"(I) attempted to bound the radiological consequences of a recriticality event by comparing it to the WASH-1400 sequence TKQ.
In this sequence, a transient occurs while the reactor
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is critical, followed by failure of the reactor protection system and by failure of the subsequently opened relief valve to close.
This results in core melt. Containment engineered safety features operate to remove heat and radioactivity from the containment atmosphere.
The fission product inventory assumed for these calculations is the current one at TMI-2.
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The results are reproduced here as Figure 1.
It shows the probability per year of a person at a given distance from the reactor site suffering a latent cancer fatality assuming the event, in this case the TKQ sequence, has occurred.* TKQ is presumed to be bounded by the curve labeled " CASE 3 TMI-2 + 1 YR" if the containment is unisolated.
If isolation is accomplished, the curve labeled CASE 2 TMI-2" is more representative of the consequences.
In either case, the probability of latent fatality to people more than five miles from the site appears negligible compared to more common causes of acciden-tal death.
For individuals at the site, the probability of latent fatality is one to two orders of magnitude higher.
The authors of Reference 1 believed the statistical uncertainty in the predictions of nuclear accident risk in Figure 1 to be no more than a factor of 100.
For the sake of later comparisons, we have modified Figure 1 to include the normal incidence of cancer fatality,and the mortality rate from all causes of death.
4.
OUTSTANDING QUESTIONS AND FURTliER EVALUATIONS This section focuses on unresolved standing questions or newly identified questions related to risk-from criticality and contains additional analyses we performed relevant to their resolution.
4.1 Probability of Criticality Though boron dilution is viewed as the most probable cause of criticality, there are other ways in which soluble boron might be lost from the core region.
Figure 2 is a simplified logic tree indicating mechanisms by which such losses might occur. We made no attempt to quantify this tree, i.e., to evaluate the quantitative probability of criticality occurring.
Rather, it The radiological source terms were not large enough to result in any acute fatalities.
No estimates of land contamination or psychological effects were attempted.
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Figure 1.
Conditional risk of latent cancer fatality as a fuhction of ~ distance for a spectrum of accident sequences.
(After Reference,I, Appendix 3, Figure 3-3).'
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' Figure 2.
Simplified logic tree indicating mechanisms by which boron could be lost from the core region so that recriticality might occur.
60 Rod co0tE0T24 Tied N RCS DROPS BELow vnicimu.m WHICH ENstARES CRIT 1cAuTY (C < C swacemca)
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Core is in optimum configuration for criticality.
No credit for control rods or burnable poison.
C is the technichl specification lower limit of 3500 ppm. subcritical
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s was believed that by taking proper precautions, the conditions necessary and sufficient for recriticality to occur could be precluded. Evaluations were performed and recommendations developed to maximize assurance that this was the case.
The mechanisms of Figure 2 fall into four areas which are discussed further below:
- concentration effects
- temperature effects
- pH effects
- chemical reactions.
4.1.1 Concentration Effects There are at least three potential sources of water with lower than desirable concentrations of boron which might ente.r the core. There are stagnant pockets within the current RCS boundary; water volumes interfacing directly with but isolated from the RCS boundary; and water volumes which could enter the RCS through suitable connections.
Examples of stagnant pockets could be letdown lines, the pressurizer, portions of the RCS drain system and other regions which are outside the natural circulation flow path. There is no way to measure the boron concentration in these locations, though it is presumed that the entire RCS has the same boron concentration as that measured near the letdown coolers.
Since these stagnant regions were originally borated, since they represent a small fraction of the RCS inventory, and since they would have the opportunity to mix with the RCS inventory before entering the core, there is no reason to suspect that they pose a problem.
An example of a water volume interfacing directly with, but isolated from the current RCS boundary is the pipe run in the low pressure injection system between the c.'eck valve nearest the reactor vessel and the motor operated isolation valve outside containment. The volume here is substantial (approx-
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11 imately 950 gallons of water), and inducing flow in the line would deliver this water to the downcomer and into the core region with.little opportunity 4
for mixing. This water.is normally borated (typically 1500-2200 ppm) since it is part of the low pressure injection system.
There are many examples of water volumes which might be potentially aligned for one reason or another to deliver into the RCS.
Examples are the refueling water storage tank, the containment sump water and less obvious sources such as fire hoses.
Of particular interest at this writing is a mini decay heat removal system (MDHRS) having a design capacity of 200 kw which is scheduled
' to be put into service in mid-April,1980.
Its design pressure is 235 psi.and it has been hydro-tested to 350 psi. When operating, this i
system will induce a flow of 150 gallon's per minute in the primary system. The MDHRS will tap into the existing residual heat removal system at the motor operated isolation valve outside containment.
j The flow will pass through the two check valves in the low pressure injection line and enter the RCS near the downcomer.
The system will receive flow from the RHR outlet in a hot leg.
It will contain j
a water sampling port approximately 50 feet from the core.
Plans 3
are being developed for monitoring the boron concentration from this location.
i The MDHRS is a closed cooling loop containing approximately 200 i
gallons. If it were assumed that water in the MDHRS containing no boron and water.in the pipe run to the RCS containing 1500 ppm boron were added to and mixed with the 30,000 galle,ns of water in the pressure vessel, the boron concentration would decrease from 3850 ppm to 3750 ppm, still well above the technical specification lower limit.
(See Appendix B for a more detailed analysis).
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4.1.2 Temperature effects The solubility of boric acid in water decreases as water temperature decreases.(I4) At the current RCS water temperature of 150 the solubility limit is 24000 ppm.
Temperature inside the containment is typically ~75 F,
and water in thermal equilibrium there can sustain 8900 ppm in solution.
At 32 the solubility limit is 4400 ppm.
Therefore, decreases in soluble boron concentration resulting from temperature decreases do not appear to pose a problem.
4.1.3 pH effects Boric acid (H 80 ) is a weak acid in water. The solubility of boric acid 3 3 in water is affected by the hydrogen ion concentration.
Additions of base, such as Na0H, to the RCS water would increase the solubility of baron. Add-itions of strong acids, such as nitric acid (H@3) w uld decrease the boron solubility. However, large amounts of strong acid would have to be added before significant decreases in soluble boron concentration were observed.
At this time there are no foreseeable circumstanc'es under which such additions would occur.
4.1.4 Chemical reactions Borate compounds are among the most soluble of all salts.
Exceptions are the borate salts formed by the alkali metals calcium and magnesium.
Large additions of aqueous solutions of these cations could precipitate boron out of solution.
At this time, there are no foresecable circumstances under which such additions would occur. However, to be prudent, any chemical additives contemplated for introduction into the RCS with the core in place should be tested for their compatibility with soluble boron.
4.1.5 Approach to criticality The rate at which criticality is approached is determined primarily by the rate of decrease of boron in solution. Of the mechanisms described above, the con-centration effects, i.e., boron dilution, appear to dominate the probability that criticality will occur,
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0 Excore source monitoring provides a direct measure of the approach to criticality. At this time only one such instrument is available and its ability to continue functioning in the severe environment it has endured l
is uncertain.
It is prudent to restore the neutron monitoring capability t
close to the core, but this requires access to the reactor head area.
l An interim measure might be to monitor the radioactivity level of the I
reactor coolant as it circulates through the MDHRS once that system is in service.
An alternative but indirect measure of the approach to criticality is the boron concentration.
In order for this parameter to be a valid measure, one would have to be assured that the actual concentration of boron in the core region is accurately represented by the concentration measured at the l
sampling port.
j If a pocket of relatively unborated water were forced through the core by l
some unspecified mechanism, the approach to criticality could be too quick for the operator to detect and prevent. The likely result of this, however, l
would be a local criticality of short duration. As will be shown in Section l
4.2, such an event is relatively inconsequential in terms of its radiological l
i i. pac t.
l By virtue of the large volume (90,000 gal) of the RCS, the current high boron concentration, and the likely low flow rates at which water would be circulated, j
it would require from days to months to decrease the boron concentration of the j
entire RCS to below critical limits.U) This should allow ample time for the l
I operators to recognize and prevent the approach to criticality.
The probability that boron dilution is detected prior to criticality increases with boron samp-j ling frequency.
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4
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4.2 Consequences of Criticality Though all practical measures should ba taken to prevent criticality, it is assumed here that sufficient boron is lost from the core so that criticality occurs. Figure 3 is a simplified event tree which portrays a spectrum of possible outcomes shaped by the availability of key safety systems. The tree is not quantified because of insufficient data on the availability of these systems under the peculiar circumstances at TMI, though it is believed they would be operable more often than not.
Furthermore, the status of the plant, the nature of the criticality and the radiological consequences are interdependent and vary with time.
Nevertheless, the tree is useful in providing a framework for subsequent analyses and in making some qualitative judgements regarding relative probabilities and consequences of events.
In this analysis consequences are expressed in terms of the energy and fission products generated during criticality and in terms of the potential effects which the former has on dispersing the latter. We made no new calculations of radioactivity dispersion in the environment or subsequent health effects.
Rather, where appropriate, estimates were made relative to those of Figure 1.
Two modes of criticality are considered:
transient and sustained.
A transient criticality (or pJlse) might be induced by a slug of cold unborated water being pushed through the core by a column of borated water. Such an event might occur, for example, when putting a new system into operation.
Table II indicates the character of the transient critical-ity assumed in this analysis in terms of the power achieved, the fraction of the core involved and the duration of the transient.
?-
.I o
Figure 3.
SIf1PLIFIED EVENT TREE TO ASSIST IN DETERMINING CONSEQUENCES OF CRITICALITY INITIATING ELECTRIC CORE BORON SYSTEM EVENT POWER COOLING INJECTION STATUS i
1 I
maintain water inventory terminate criticality Cool Core YES maintain water inventory i
sustain criticality cool core i
l f
lose water inventory G
V terminate criticality N0 heat up core CRITICALITY l
lose water inventory terminate criticality i
heat up core e
e 1
16 Sustained criticality is the other mode of criticality considered.
This might be induced by a continuous flow through the Reactor Coolant System (RCS) of water having a low or zero concentration of boron. Once the boron concentration decreased to about 1500 ppm, the core is assumed to go critical.
Further reductions in boron concentration would increase the ultimate power level achieved.
Criticality would be detected by RCS core neutron monitoring, pressure, temperature and radiation detectors. Given zero or low forced flow or inability to restore the boron concentration, pool boiling would occur in the RCS.
Some heat would be transferred via natural circulation to surrounding structures and to secondary heat sinks.
Energy would be stored in the RCS water until the availability of a pressure relief path from the RCS allowed energy to be transferred via vaporization of water.
An equilibrium power level would be reached whose magnitude would depend primarily on boron concentration, fuel temperature and voiding in the core. (See Section 3.2).
Without makeup flow, the water level would decrease and this loss of moderator would eventually j
terminate criticality. Of course, restoration of boron could also be.'used to terminate criticality. The energy stored in the fuel and the fission product,
decay heat balanced ag'ainst available heat removal mechanisms would determine the driving force for heating the fuel and for dispersing radioactivity during and subsequent to criticality.
The approach to criticality and the course of subsequent events depend most upon those factors, including operator actions, which affect the time depen-dent concentration of boron in the core region.
For the purpose of providing quantitative indicators of consequences, three cases of sustained criticality were assumed as described in Table II. The power level and time at power are the variants. The practical basis for the assumed cases is that boron dilution goes undetected long enough to reach criticality.
Criticality is detected within minutes after it occurs and operator action halts further dilution.
The RCS and containment are successfully isolated.
It is assumed, however, that efforts to
TABLE II.
ENERGY ACCOUNTING FOR CRITICALITY EVENTS TRANSIENT SUSTAINED CRITICALITY (PULSE)
CRITICALITY CASE A CASE B CASE C Energy Generation Rate (MW) 2772 27.7 277 277 Fraction of Core Which,is Critical 0.1
.1. 0 1.0 1.0 Time at Power (MIN) 1 60 60 600 4
5 6
7 Total Energy Generated (MJ) 1.7x10 1x10 1x10 lx10 Energy Dissipated (MJ) 3 3
4 4
Q Heat Fuel to Maximum Power Level 5.1x10 9.8x10 1.7x10 1.7x10 Heat RCS Water to RHR Pressure Relief 4
4 4
4 Conditions 8.0x10 8.0x10 8.0x10 8.0x10 Heat RCS Water to RCS Pressure Relief 5
5 5
Conditions N.A.
3.9x10 3.9x10 3.9x10 4
4 Heat RPV to RCS Pressure Relief Conditions N.A.
N.A.
5.2x10 5.2x10 5
5 Vaporize Half of RCS Water Inventory N.A.
N.A.
4.7x10 4.7x10 6
6 (SUBTOTAL)
N.A.
N.A.
1.0x10 1.0x10 Energy from Criticality Remaining to be 6
Dissipated to Prevent Fuel Melting (MJ) 0 0
0 9.0x10 Energy to Heat Fuel to Melting from Equilibrium 4
4 Power Level (MJ)
N.A.
N.A.
5.4x10 5.4x10 Decay Heat Power at Shutdown (KW) (Including 164KW Prior to Recriticality) 166 174 264 1164 Time to Fuel Melt at Decay Heat Power (DAY)
N.A.
N.A.
4.8 1.0 d
e 18
/
increase the boron concentration are nonproductivo for the specified time.
Sustained criticality Case A is our realistic estimate of equilibrium power level at the current system pressure cf 280 psi.
Case B is our realistic estimate of eouilibrium power level at 2200 psi.
Case C assumes the same power level but includes a longer time for corrective actions.
The results of energy balance calculations to assess the thermal response of the system are given in Table II. The conclusions drawn from these results are:
(i) - Energy generated in the transient criticality and in sustained criticalities where corrective action is effective within an hour is consumed in raising the fuel temperature and heating the RCS water. There is insufficient energy left to uncover the core.
(ii) - Most of t'he energy generated by the sustained criticality of longer duration must be removed from the RCS in order to avoid loss of water inventory and subsequent fuel melt (i.e., the energy cannot be absorbed within the RCS itself).
(iii) - The energy to be dissipated in order to prevent core uncovery and fuel melt is within the range of heat removal capability for natural circulation through the steam generators.
(iv) - Substantial periods of time exist prior to the calculated initiation of fuel melting should core cooling be lost.
The fission product inventories generated in the criticality events analyzed are given in Table III.
The current inventory at TMI-2 is shown for comparison. The conclusions drawn from this table are:
(i) - The total inventory of fission products generated during transient criticality is insignificant relative to the current TMI inventory.
(ii) - The total inventory of fission products generated during sustained criticality is comparable to the current TMI inventory..
(iii) - All criticalities generate inventories of the volatile xenon and iodine isotopes many orders of magnitude greater than those in the current TMI inventory.
e w
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- T 9 be if e-+W
'T' 4-
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T TABLE III.
FISSION PRODUCT INVENTORIES GENERATED DURING CRITICALITY EVENTS I
TRANSIENT SUSTAINED CRITICALITY CURRENT (PULSE)
TMI CRITICALITY CASE A CASE B CASE C INVENTORY Energy Generation Rate (MW) 2772 27.7 27.7 277 Fraction of Core Which is Critical 0.1 1.0 '
1.0 1.0 Time at Power (MIN) 1 60 60 600 4
5 6
7 Total Energy Generated (MJ) 1.7x10 lx10 lx10 lx10 Fission Products Generated (Ci)
U 3
4 5
6 5
Krypton 5.1x10 3.2x10 3.2x10 3.2x10 1.0x10 3
4 5
6 Xenon 7.1x10 4.4x10 4.4x10 4.4x10 2.3x10-3 3
4 5
6 Iodine 8.4x10 5.1x10 5.1x10 5.1x10 2.2x10-I 3
4 5
6 6
Cesium 6.4x10 3.9,x10 3.9x10 3.9x10 1.2x10 4
5 6
7 7
Others 9.0x10 5.3x10 5.3x10 5.3x10 4.0x10 5
5 6
7 7
TOTAL 1.2x10 7.0x10 7.0x10 7.0x10 4.1x10,
- Estimated disposition of current fission product inventory at TMI-2 is as follows:
4.4x10 Ci of Kr in containment 4.0x10 Ci in RCS water 5.0x10 Ci in containment sump i
4.4x10 Ci in auxiliary building storage tanks 4.0x107 C: retained primarily in fuel 1
9
6 20 To assess radiological consequences of these events, it is necessary to consider the mechanisms and the driving forces by which fission products can be transported across physical barriers on the pathway to the environment.
The normal physical barriers are the fuel matrix, the fuel rod clad, the reactor coolant system boundary and the containment building.
In this analysis, no credit is given for fuel rod clad as a physical barrier since most of the rods were presumed to have failed in the origital accident.
The principal driving forces for transport are the energy generated during and following the critical-ity and the fluid flows across these boundaries.
Table IV describes the applicable fission product transport mechanisms and some characteristics which help relate them to the estimated consequences of criticality. When combined with an understanding of the possible physical states of the plant, the conclusions drawn from this table are:
(i) - Once criticality has occurred, there is nothing that can be done to prevent significant additional amounts of radioactivity from entering the RCS water.
(ii) - Minimizing the fuel temperature during and following criticality will be most effective in preventing still much larger amounts of radioactivity from being available for transport.
(iii) - Maintaining isolation of the RCS while assuring core cooling is the earliest opportunity to limit the spread of radioactivity to the environment.
(iv) - Assuring the operability of the containment engineered safety features, e.g., the sprays, is an effective way to retain radioactivity inside the containment if core cooling is lost.
(v) - Maintaining isolation of the containment is the last opportunity to limit the spread of radioactivity to the environment.
s 8
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o
TABLE IV.
CONSIDERATION OF FISSION PRODUCT TRANSPORT f1ECHANISMS i
i MECHANISM WHERE APPLICABLE CHARACTERISTICS OF MECHANISM SOLID STATE TRANSPORT KN0CK0UT RELEASE FROM FUEL TYPICALLY 1-10% RELEASE i
MATRIX TO RCS WATER INCREASES WITH FUEL FREE SURFACE AREA i
INDEPENDENT OF FUEL TEMPERATURE RELEASE-TO-BIRTH RATIO IS LOW AND INDEPENDENT OF FP VOLATILITY
,?
DIFFUSION RELEASE FROM FUEL MATRIX TO CONTROLLED BY PRODUCT OF TIME AND RCS WATER OR STEAM FUEL TEMPERATURE RELEASE-TO-BIRTH RATIO IS HIGH AND INCREASES DIRECTLY WITH FP VOLATILITY AEROSOL WITHIN AND FROM RCS VAPORIZATION AND CONDENSATION OF S
REQUIRES FUEL TEMPERATURES >l800C l
WITHIN AND FROM CONTAINMENT CONTAINMENT SPRAYS REMOVE THEM EFFECTIVELY AGGLOMERATION AND SETTLING INCREASE WITH TIME REGARDLESS OF SPRAYS LIQUID TRANSPORT WITHIN AND FROM RCS INCREASES WITH LEAKAGE FROM RCS WITHIN AND FROM CONTAINMENT INCREASES WITH LEAKAGE FROM CONTAINMENT VAPOR TRANSPORT WITHIN AND FROM RCS INCREASES WITH STEAM FLOW IN RCS INCREASES WITH HIGH WALL TEMPERATURES INCREASES WITH LEAKAGE FROM RCS g
WITHIN AND FROM CONTAINMENT INCREASES WITH AP ACROSS CONTAINMENT INCREASES WITH HIGH WALL TEMPERATURES
?
gg Based on the material presented to this point, we have attempted to estimate the effects of criticality on dose rates inside containment and on release of radioactivity to the environment. These depend strongly on the efficacy of core cooling and on the pathways available_ for fission product transport.
The most probable sequence of events given the occurrence of criticality is that core cooling is achieved via conduction to surroundings and natural circulation through the steam generators.
If cooling is efficient enough, it is likely that the system pressure can be maintained below the relief and safety valve set points, thereby maximizing RCS integrity. Given this sequence, the principal release of fission products will be from the fuel to the RCS water during fission. Assuming a 10% release fraction character-istic of knockout, sustained criticality Case B would produce a twenty l
fold increase in the gross radioactivity level of the RCS water (currently 4
4 x 10 Ci in 90,000 gal water). Of course, much larger increases in the concentrations of short-lived isotopes such as I-131 would be observed.
With successful isolation of containment, releases to the, environment would be controlled and possibly too low to measure.
If core cooling were deficient enough to allow relief valves to open or if a lower pressure path from the RCS to containment were available, RCS fluid would leak out taking with it noble gases and some dissolved and particulate radioactive material.
Leaked water would enter the containment sump, and the noble gases would increase the radioactivity levels in containment.
Emptying the 5
entire inventory of the RCS (7.4 x 10 Ci in 90,000 gal) into the containment 5
sump (5x10 Ci in 600,000 gal) would more than double the radioactivity contained there. This source would increase dose rates in the sump region.
It would, however, little affect dose rates in the upper regions of the containment unless the containment spray recirculation system were activated.
1 a
D.
6
,.,..,w v.
.m..
23 Dose rates in the containment would increase as a result of a factor of 2.5 increase in the noble gas inventory. Naturally, these dose rates would drop as the short-lived isotopes decayed.
However, the net long term effect would be a substantial increment above the current dose rates. Subatmospheric pressure in the currently isolated containment keeps nobic gases from leaking out.
Pressure increases could negate this effect, but the driving forces associated with this event do not appear great enough to produce significant out-leakage.
In the less probable event that core cooling were deficient enough to allow uncovery of the core, quantum increases in the amounts,of radioactivity released from the fuel would be observed. This could be accompanied by a breach of the RCS boundary and dose rates in containment would certainly
' increase by an order of magnitude or more.
Reliance for consequence mitigation would be placed on the containment and its engineered safety features.
At this point comparison with the results in Figure 1 is appropriate.
The major difference between this analysis of consequences and that of Reference 1 is the assumed fission product inventory.
Reference 1 assumed the current TMI-2 inventory, i.e., no increased inventory as would be produced by any of the criticality events described in Table III.
Here we assume the more conservative energy release and inventory of sustained criticality Case C.
Comparisons are made for the following circumstances: (i) meltdown inside an essentially intact containment and (ii) meltdown inside an unisolated containment without containment heat removal or sprays.
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_-,y-y
24 6
(i) - For meltdown within an intact containment, the conditional probability of latent cancer fatality would increase by no more than a factor of eighty; from 10-9 to 8 x 10-8 at five miles. At such low probability values, such a difference is insignificant, since it is within the uncertainties of the baseline value (i.e., factor of 100). The increase is attributable to the eighty-fold increase in the noble gas inventory, which is assumed to leak at one volume percent per day. The less volatile fission products, including most of the fodine, would be retained effectively by containment sprays and natural agglomeration and settling. Melt-through of the containment base mat would not occur.
(ii) - For meltdown inside an unisolated containment without containment heat removal or sprays, the conditional probability of latent cancer fatality would increase by about a factor of three; from 10-6 to 3,x 10-6 at five miles. At such low probability values, such a diff-erence is difficult to distinguish.
The increase is attributable to the gross inventory increase generated by the criticality (i.e.,
7 7
7.0 x 10 'Ci added to the 4.1 x 10 Ci already there) and to the presence of the volatile short-lived isotopes, all of which exit containment..The potential for thyroid nodules resulting from the.r61 ease.of I-131 vould be roughly ten times that for latent cancer.
Therefore, even for the conservative cases assumed here, the off-site con '
sequences as expressed in terms of probability of latent cancer fatality are negligible compared to the normal incidence of that health effect.
No estimates were made here of the potential for land contamination or psychological effects.
Only meltdown inside an unisolat,ed containment without containment engineered safety features would likely result in significant land contamination.
It is important to keep in mind when considering these results that the pro-bability of criticality is not_ unity as has been assumed here. Nor are the probabilities of failure of engineered safety features unity.
Precautions are taken to ensure that such probabilities are as low as practical for TMI.
t The point is that consequence analyses such as these must be taken in context with their associated probabilities.
l
^
l
25 5.
CONCLUSIONS AND RECOMtiENDAYIONS J
The conclusions of this study are as follows:
(i) - Previous studies performed independently are in substantial agreement regarding the necessary and sufficient conditions which must be met in order to achieve criticality in the THI-2 core. This study has uncovered no evidence to the contrary.
Furthermore, there has been no indication of gross inaccuracies in the findings of previous studies which would tend to underestimate the likelihood of criti-cality.
(ii) - Previous studies have assumed that the most probable cause i
of recriticality is boron dilution. This study has examined other mechanisms by which boron might be lost from the core and has reached the same conclusion.
(iii) - Previous studies have made no attempt to quantify the absolute probability that the necessary and sufficient conditions for criticali.ty will be satisfied at THI-2.
Rather, it is believed i
that the most probable mechanism for recriticality, i.e., boron dilution, is a slow enough process that the approach to criticality will be detected and corrective actions taken, provided adequate instrumentation, procedures and equipment are available.
This study agrees with that approach.
It concludes that to the extent boron concentration in excess of 3500 ppm can be ensured, the probability
~
of criticality is minimi7ed.
5 (iv)-Giventheemphasisonpreventingcriticalityinpreviousstudies, little attention has been paid to poteng1 radiological consequences.
considers consequences.
l Only the most recent task force report It indicates that latent cancer risk to off-site individuals from 1
criticality is many orders of magnitude lower than the probability of fatality from common accidents and from all causes of cancer.
i This study indicates that Reference 1 may have underestimated the potential consequences of criticality but not by enough to affect the basic conclusion.
l (v)
The most probable direct radiological consequence of criticality i
is the increase in dose rates inside containment. The magnitude i
of this increase depends primarily on the efficacy of core cooling l
and the ainlity to maintain RCS integrity.
Depending on that mag-nitude, the duration of the cleanup effort could be extended sig-l nificantly.
Increased indirect consequences such as higher occupa-tional exposures and greater likelihood of key equipment failure I
might be anticipated.
(vi) - Most probably, criticality will not result in significant off-site radiological consequences.
For less probable events there are sizable variations, i.e., one or more orders of magnitude within the spectrum of off-site c'onsequences that can be calculated. The more severe con-l sequences are less probable since they involve multiple failures of independent systems. To the extent that core cooling and containment integrity can be maintained, the off-site consequences are minimized.
i t
[
26 2
o.
! Pretfious studies have presented many recomendations designed to reduce the risk of criticality. We have made no effort to review all of these recom-mendations or to inquire as to their implementation.
Rather, we present here some recommendations which occurred to us during the course of this study and propose that those responsible for the operations at TMI take them under advisement.
(i) - To minimize the potential for criticality when the mini decay heat removal system is put into service, the following recom-mendations are made:
- All MDHRS water should be borated to 3850 ppm.
- The system should be started with low flow to facilitate mixing of the MDHRS water, the water in the lead-in pipe run, and the water in the pressure vessel.
- Boron concentration should be monitored more frequently; as frequently as practical during startup and no less than once per shift afterward. The sampling port should be within the flow and as close to the core as is practical.
- The system should be instrumented with radioactivity monitoring equipment, either gamma or delayed neutron detectors, so as to provide.a diverse measure of approach to criticality.
(ii) - Review the potential for introducing unborated water into the RCS and generate a.dministrative preventive measures where appro-priate.
(iii) - Prepare procedures to guide the operators regarding corrective action should a decrease in boron concentration be detected for whatever reason.
(iv) - Prepare procedures to guide the operator in the event that instru-mentation to monitor the approach to criticality is lost.
(v) - Investigate a standby neutron poison injection system to supply back-up in the unlikely event that a pocket of low boron con-centration should be swept into the core.
Chemical compati-bility of boric acid with cadium nitrate or sulfate or gadalinium l
nitrate should be investigated as an alternate to a concentrated boric acid injection.
(vi) - Review the instrumentation available to provide direct and indirect measures of criticality and the readings likely to be observed.
(vii) - Prepare procedures to guide the operators regarding corrective action should criticality be detected for whatever reason.
(viii) - Have procedures and equipment available for ensuring and con-l
. firming heat removal through the steam generators.
(ix) - Review the capabilities and procedures for isolating the RCS in its current configuration.
(x) - Review the capabilities and procedures for operating containment engin-eered safety features and for isolating containment.
~;
6 27 (xi) - Place high priorii.y on augmencing the excore neutron monitoring capability once contdi6 ment entry has been gained.
(xii) - Repeat the review of recriticality prior to removal of the reactor vessel head to take into account new information.
O 6
'4 4
e
- e e
28
?
REFERENCES 1.
Haller, N.M., et al, " Evaluation of the Cleanup Activities at Three Mile Island, " USNRC, February 28, 1980.
2.
Memorandum, R. Bernero (NRC/RES) to R. DiSalvo (NRC/RES), "Special Assignment - THI-2 Hazard Evaluation," March 10, 1980.
3.
" Evaluation of Long-Term Post-Accident Core Cooling of Three Mile Island Unit-2," HUREG-0557, April 1979.
4.
Memorandum, C. Marotta (NRC/NMSS) to K. Kniel (NRC/NRR), "Recriticality Potential of THI-2 Core," May 14, 1979.
5.
Memorandum, H. J. Richings (NRC/NRR) to R. Mattson (NRC/NRR), "TMI-2 Event Excore Neutron Detector Readings: Cause and Significance,"
August 24, 1979.-
6.
Letter, G. Kulynych (B&W) to R. Harding (Metropolitan Edison Co.),
" Basis for Tech Spec Boron Limits," May 1, 1979.
7.
Memorandum, D. Cokinos (BNL) to D. J. Di.amond' (BNL), "Recriticality Calculations for TMI,"
Brookhaven National Laboratory, May 18, 1979.
8.
Memorandum, G. R. Bond (GPU Service Company) to B. D. Elam (GPU Service Company),
" Recommended Boron Concentration Levels in TMI-2," August 8,1979.
9.
Barr, E. W., et al, "TMI-2 Post-Accident Criticality Analysis," TDR-049, General Public Utilities Service Company, August 31, 1979.
- 10. Westfall, R. M., et al, " Criticality Analyses of Disrupted Core Models of Three Mile Island Unit 2," 0RNL/CSD/TM-106, Oak Ridge National Laboratory, December 1979.
(Prepared for the President's Commission on the Accident at Three Mile Island).
- 11. Memorandum, C. R. Marotta (NRC/NMSS) to N. M. Haller (NRC/MPA), "Nonconservative Reference Keff of TMI-2 Core as Given in Special Task Force Report ' Evaluation of the Cleanup Activities.:t Three Mile Island,' Dated February 28, 1980,"
April 1,1980.
- 12. Thompson, T.
J., and Beckerly, J.
G., (eds.), The Technology of Nuclear Reactor Safety, Vol. 1, M.I.T. P.ress, 1964,
- 13. Stratton, W., "A Review of Criticality Accidents," LA-3611, Los Alamos Scientific Laboratory, September 1967.
- 14. Cohen, P., Water Coolant Technology of Power Reactors, Gordon and Breech, 1969.
l 1
~
6 e
APPENDIX.'A
SUMMARY
OF CRITICALITY ANALYSES 0F'TMI-2 1
1 1
I L
e A-1 APPENDIX A
SUMMARY
OF CRITICALITY ANALYSES OF TMI-2 NRC Staff Report
" Evaluation of Long-Term Post-Accident Core Cooling of Three Mile Island Unit-2," NUREG-0557, April 1979.
This analysis is based primarily on calculations by the B&W naval criticality group using the KENO-IV Monte Carlo code. Calculations and cross sections have been tested against many experiments. The calculations on slumped cores assume no control rod or burnable poison material in the core.
1 The major conclusions of this analysis are:
1.
For no collapsing of " layers" the system is will subcritical at a boron concentration of 1500 ppm.
2.
For a collapse of 3 " layers", giving a combination of about 42% of the reactor fuel, criticality would be approached at 1500 ppm but it would be about 4% subcritical at 2200 ppm.
3.
For a collapse of 5 " layers", giving a combination of abcut 71% of the reactor fuel, the system would be several percent supercritical at 2200 ppm but several percent subcritical at 3000 ppm.
4.
For a complete combination of all fuel, either in a cylinder or sphere the system would be slightly subcritical at about 3000 ppm.
This last result is the basis for B&W advocating a boron level of 3000 ppm to cover the most extreme configuration.
l i
l
5 A-2 O
Memorandum, C. Marotta (NRC/NMSS) to K. Kneil (NRC/NRR), "Recriticality Potential of TMI-2 Core," May 14, 1979.
Caluclations were KENO-IV Monte Carlo. code.
Assumptions:
a.
no control rods b'.. no burnable poison c.
two zone core, 2.96% outer zone and (1.98 + 2.64)/2% = 2.31% inner zone d.
no core barrel, 2 - foot unborated water reflector.
This is 0.5% to 1% AK/K conservative.
Results:
Benchmark calculations on zero power, 530 F, clean,: all rods out, just just critical TMI-2 core with 1500 ppm boron was within +0.5 AK/K.
Results on the as-built TMI-2 lattice arid the lattice with fuel rearrangea in the most reactive pitch are shown in Table I.
Local criticality:
Keff of four assemblies, 2.96% enriched fuel..in square array by pure water (B) ppm Keff 2500 0.839 + 0.004 2000 0.866 1500 0.886 1000 0.924 500 0.953 0
1.000
==
Conclusions:==
2% of core filled with pure water will result in local criticality.
l l
The technical specification boron concentration of 3500 ppm will reduce the i
Keff of the most reactive configuration in Table A-I to less than 0.90.
i 6
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A-3
~
TABLE 1 K
of TMI-2 Core As Function.0f PPM Baron in Water eff (No Control Rods or Burnable Poisons)
(RoomTemp)
AS BUILT PITCH MOST REAC PITCH 1.44 cms 1.26 cms PPM BORON PPM B0 lion ZR-CLA0 A
B C
K ZR-CLAD A
'B C
K gff eff c
l' YES 1500 1500 1500
- 1. 04 0 -
YES 3000 3000 3000 0.944 YES 3000.
3000
'3000 0.883 YES 3000 3000 2000 0.954 NO 3000 3000 3000 0.857 YES 3000 3000 1500 0.989 YES 3000 3000 1000 0.992 N0 3000 3000 3000 0.936 NO 2500 2500 2500 0.977 NO 3000 2500 2000 1.000
- All K calc. by KENO-123 Gps, using 15,000 neutron Mstories and all within
+0.00%fnK,ff f
for 1 St.dev.
I Corf7AIN S 10100 run. Roo $; 2.31*/ E ; 33 '/, cog; 3
toff 7AtWS 11 '134 5 ve:. rod 5 1.11%E 33 % cets 3
3 3
>M*A tHS 12/28 Wet, RaOS; 2 9 t,'fg,.'349p; UHBoRATED WA f_
g (AtL-ARounO )
ff (
/u REFLECTcR.
g ---
.A
. ia rr ActsvE gtgT}/ E;3 A 5 e c L A 7 T / t-e T
E s$$* q
/
W.. ~
t l
4-P d TY1 ~I R =.10 ym R =.5 n 2c m.
COP 2 pupiTcH
(
S Kerca 4
D-A Fo R.
TAB W i, abo #
6 A-4 Letter.
G. F. Kulynych (B&W) to R. W. Harding (Metropolitan Edison Company),
" Basis for Tech Spec Boron Limits," May 1,1979.
B&W recommends a lower limit of 3000 ppm based on B&W criticality calculations:
B&W recommends an upper limit based on solubility of O
Tempe ature F 50 5000 60 6000 70 and higher 7000 The material attached to the letter gives the following data on the un-damaged core at 88.3 EFPD at cold shutdown from PDQ-7 calculations.
T, 3F Control rods Keff Boron, ppm 70*
all out
.95 2155.
70*
all out
.99 1795 70*
all in
.95 1705 70*
all in
.99 1385 280*
all out
.97 2'iOO 280*
all in
.92 2l00 135 No credit for L0mped Burnable Poison, no Xe No credit for Sm buildup, 1% AK/K conservation.
l l
i l
t
g A-5 Memorandum, D. Cokinos (BNL) to D. J. Diamond (BNL), "Recriticality Calculations for TMI," Brookhaven National Laboratory, May 18, 1978.
This analysis used the HAMf1ER multigroup, integral transport theory code.
This code, originally developed by duPont at Savannah River Laboratory and revised by EPRI-NP-565 in October 1978, has been successfully used by the nuclear industry for years.
It offers an analysis of TMI-2 criticality that is completely independent from the ORNL and NRC/NMSS Monte Carlo (KENO) calculations:
The cases considered were pellet slump with no control, rod or burnable poisons based on the average fuel enrichment of 2.6%.
The results are:
% core slumped critical Boron concentration 30%
2720 ppm 50%
2900 ppm 100%
3C50 ppm 8
a
E A-6 L
Memorandum, G. R. Bond (GPU Service Company) to B. D. Elam (GPU Service Company);
" Recommended Boron Concentration Levels in TMI-2," August 8,1979.
This internal memorandum recommends TMI-2 boron concentrations based on calculations later reported in Reference 9.
Recommendations:
Minimum Boron Concentration 3500 ppm Target Boron Concentration 3900 ppm Maximum Boron Concentration 4300 ppm The basis of these recommendations is the analysis of the foltowing cor.rigurations:
1% Shutdown Configuration Boron Concentration
- 1. Optimum Pellet Water 3400 Density Mixture
- 2. Total fuel Pellet 3470 Slump (SLAB)
- 3. Intact High Enrichment 3270 Fuel, No Discrete Poison Control 1
The target and minimim concentrations are a direct result of the current l
j evaluation. The maximum concentration is unchanged from the proposed TMI-2 Technical Specifications and remains substantially below the theore-tical boron solubility limit.
Consequently, no significant boron precip-itation is expected.
I i
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A-7 Barr, E. W., et al, "TMI-2 Post-Accident Criticality Analysis" TDR-049 GPUSC Nuclear Analysis Section, August 31, 1979.
Calculations presented in the analysis were made by two different methods.
Monte Carlo Calculations with the KENO-IV code, the same code used by B&W, ORNL and NRC/NMSS and the XPOSE computer code. XPOSE is an Exxon Nuclear Company version of the widely used Westinghouse LEOPARD code.
Both have been widely checked against experiments and approved by NRC for licensing purposes.
The conclusions of this study were previously summarized in Reference 8.
A careful review of the details of the calculations on models in this' 100 page report leads to the conclusions that the recomendations of Rdference 8 are valid.
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A-8 Westfall, R. M., et al, " Criticality Analyses of Disrupted Core Models of Three Mile Island Unit-2,'.'
ORNL/CSD/TM-106, Oak Ridge National Laboratory, December 1979.
This. study was prepared for the President?s Commission on the Accident at Three Mile Island. Two Monte Carlo codes were used for this analysis; KEN 0 IV and MORSE-SGC/J. A 27 group cross section library was used which is a subset of an 218-group ENDF/B-IV library. The calculational methods were checked against 7 critical experiments and the TMI-2 startup criticality tests.
In the range of water-metal ratios of interest, the calculations underpredict Keff by about 1.3% nK/K. An adjustment for this was made when using these results.
Three models of core disruption were studied. All cases had 3180 ppm boron in the water. These are:
a.
MORSE-SGC/S Three Jump Slump Core Model as shown on Figure 1.
(Note the figure and table number from the original report).
Results are g' en in Table 13. Correcting Case B for tne effect of burnable poisons (LBP) and the 1.3% AK/K bias:
0.875 + 0.006 + 0.013 = 0.894 b.
KEND-IV Displaced Fuel Slump Model as shown in Figure 2.
Results are given in Table 14.
Correcting Case B for the 1.3% AK/K bias:
Keff = 0.870 + 0.013 = 0.883 c.
KENO-IV In-Place Fuel Slump Model as shown in Figure 3.
Results are given in Table 15. Correcting the 50% swelling case which has the l
highest Keff for the 1.3% 6K/K bias:
Keff = 0.845 + 0.013 = 0.858
~
The present Technical Specifications on boron is 3500 ppm while these cal-culations were at 3180 ppm. The additional 320 boron would lower Keff by i
about 3%.
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A-9 Axial l Zone IV l Spacer
!44" l Zone I lZoneII l
Zone III Grids Level g_._.______
H20-B'(3180) g 7
t-
,_____j
/
/
1st 123.42"
/ (2.3)30. (0.635),
U 6
/
H20-8 (0.313) 102.85"
/
2nd 5
82.28"
/
3rd Swollen fuel Pins, Zr0 in Coolant 4
2 4th 61.71"_..
h Fuel. Pins Swollen 30%,
3 Nonnal Coolant 5th 2
20.57" 6th 1
7th Fig. 1.
MORSE-SGC/S Three Jump Slump Core Model*
OControl and Lumped Burnable Poison Rods from Disrupted Portion of Core Missing.
Boron in Coolant in All Zones at 3180 wppm. Core Barrel, Radial, and Axial Reflector Regions in Model.
~
d A-10 144" -
HO+B 2
120" _
U(2.57)30s - H2O + B Mixture
/
.r
~
72"_
Normal Pin-Lattice Core Boron at 3180 wppm Fig. 2.
KENO-lV Displaced Fuel Slump Model*
i
- Includes Radial and Axial Reficctors of 1120+B
~ '.
l
1 A-11 0
144" H O + B, Pin-Lattice Without 002 2
ha f
Pin-Lattice Core, Fuel Pin Volume Increased With A Constant Density and Mass of U02 Boroh'at 3180 wppm Fig. 3.
KENO-IV In-Place Fuel Slump Model hg values:
144", 114.2", 94.6", 80.8", 70.4" bIncludes Radial and Arial Reflectors of H2O 4 B
~
d 4--..
S A-12 Table 13 MORSE-SGC/S "Three Jump Slump" Disrupted Core Multiplication Case Description Factor A.
Base configuration" 0.862 0.006 B.
Case A with control rods out 0.875 0.006 C.
Case A with LBP rods removed 0.868 0.006 D.
Case A with controls rods and boronb out 1.079 i 0.012 E.
Case A with LBP rods and boronb out 1.043 0.010 F.
Case A with control rods inserted, boron out 0.988 0.011
- 13.5% of upper middle core collapsed as U308-H2O mixture; Zr02 distributed in coolant channels of lower core; intact portion of fuel pin swollen by 30%; boron in coolant at 3180 wppm.
Boron remaining in U 0 -H O mixture.
38 2 Table 14.
KENO-IV " Displaced-Fuel Slump" Disrupted Core #
Multiplication Case Description Factor A.
Base configuration 0.845 0.006 B.
Case A with control rods out 0.870 i 0.006 b
C.
Case A with boron out 1.080 i 0.006 Upper 50% of core collapsed as'U 0 -H O mixture; 38 2 corresponding portions of control and LBP rods missing; lower half of core in normal configura-tion; boron in coolant at 3180 wppa.
bBoron remaining in U 0 -H2O mixture.
38
1
.. ~
e-Table 15. KENO-IV "In-Place Fuel Slump
- Disrupted Core" A'asumptions: Fuel stays at constant density (0.925 or theoretical);
Zr clad expands at constant volume;b fuel height drops to conserve volume.
Hin. Gap Swelling Height Fuel OD Clad OD between pins KEND-IV XSDRNPHd f
(% of Max)
(cm)
(cm)
(cm)
(cm) k-eff Lattice k.
C None 365.8 0.94 1.092 0.176 0 737 0.006 0 907 255 290.0 1.056 1.179 0.132 0.807 0.006 0 980 50%
240.2 1.160 1.273 0.085 0.845 0.005 1.014 3,
75%
205.2 1.255 1.360 0.042 0.840 0.006 1.005
.L 100%
178.8 1.344 1.443 0.0 0.812io.0073 0 950
[horonat3180wppm,constantlatticepitch=1.443cm.
Constant clad volume, interior radius increases.
- Clad, control rods & LBP rods above, core as normal.
Ch.57wt5enrichedUO2 (core average).
6 o.
C a e g
s G
e APPENDIX B LOW BORON CONCENTRATION AT TMI-2 4
0.3 g
u UNITED STATES e...f p3 Keq NUCLEAR REGULATORY cOMMisslON o
WASHINGTON, D. C. 20555 t
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g
%s MAR 2 81980 Docket No. 50-320 MEMORANDUM FOR:
R. DiSalvo, Probabilistic Analysis Staff, RES FRDM:
A. J. Ignatonis, NRC/TMI Technical Support Staff
REFERENCE:
B&W Letter TM/235 LCR/171 from L.C. Rogers of B&W to Messrs.
J.C. Devine and G.R. Skillman of GPU, dated February 19, 1980
SUBJECT:
LOW BORON CONCENTRATION AT TMI-2 Per your request I performed some work regarding the existence of possible low boron concentratiun water in the stagnant loops that are connected to the Decay Heat Renoval System (DHRS).
According to the Burns and Roe drawings the volume of RCS water contained between valve DH-V4B and th'e check valve CF-V5B (discharge side of DHRS) is estimated to be approximately 950 gallons. The volume in the drop line (suction side of DHRS) located between the intersection of the hot leg and valve DH-V3 is estimated to be approximately 430 gallons.
Both of these loops are stagnant and haven't been borated since the accident.
Recent information from the licensee shows that the boron concentration in these loops ranged from 1,500 ppm to 2,250 ppm prior to the accident.
I have performed a rough evaluation to determine the overall result when adding low boron concentration water to the RCS.
For simplicity, complete mixing and 0 ppm boron concentration in the stagnant loop was assumed.
Based on the B&W figures provided in the above reference, the boron concentration of the mixture is 3,564 ppm when mixing the following volumes:
reactor vessel water (30,150 gal.
l 9 3,800 ppm) plus the DHRS water (1,122 gal
@ 2,270 ppm) plus the MDHRS water (950 gal.
+ 430 gal. 0 0 ppm)plus the two loops--suction and discharge to the (200 gal. 0 0 ppm) t tration provided in the above reference, the addition of the two unaccounted volumes of water would not decrease the boron concentration below 3,000 ppm during. injection.
Furthermore, since there may not be complete mixing in the RCS, and there may be l
some other uncertainties such as cold water in stagnant lines, for safety reasons '
plans are being made to drain and borate the DHR and the MDHR systems prior to I
MDHR system operation.
If you have any further questions, you can contact me or anyone on the TMI Technical Support Staff.
i
- p. c /g &
l Algis J IgnatVnis NRC/TMI Technical Support Staff cc:
J. Collins R. Conte
Contact:
Al Ignatoni,s, NRR M. Greenberg G. Kalman 49-29403 T. Poindexter duPC-
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