ML20082G218
| ML20082G218 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 11/22/1983 |
| From: | Hukill H GENERAL PUBLIC UTILITIES CORP. |
| To: | Stolz J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM 5211-83-328, NUDOCS 8311300117 | |
| Download: ML20082G218 (32) | |
Text
.
GPU Nuclear Corporation
- Nuclear
- Snze8s48o 8
Middletown. Pennsylvania 17057 717 944-7621 TELEX 84-2386 Writer's Direct Dial Number:
November 22, 1983 5211-83-328 Office of Nuclear Reactor Regulation Attn:
J. F. Stolz, Chief Operating Reactors Branch #4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Sir:
Three Mile Island Nuclear Station Unit 1 (TMI-1)
Operating License No. DPR 50 Docket No. 50-289 Post Accident Sampling (NUREG 0737 II.B.3)
In response to your letter of July 13, 1983 (Draft SER) and as a result of our telephone conversation on August 25, 1983 between meribers of our respective staffs, enclosed are responses to the six outstanding criteria.
Sincerely, I
ki 1 Vice President TMI-l HDH/kls Enclosure cc:
J. Van Vliet R. Conte 8311300117 831122 0I PDR ADOCK 05000289 i
P pyg GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation
ENO_0SURE Criterion 2 - On site' Radiological and Chemical Analysis The Licensee should provide a procedure to estimate the extent of core damage based on radionuclide concentrations and taking into consideration other physical parameters such as core temperature data and sample location.
Response.
Based on guidance attached to your letter of July 13, 1983, and in accordance with our discussion over the phone on August 25, 1983, GPUN has developed guidelines for estimating the extent of core damage for TMI-l based on gamma analysis of reactor coolant. These guidelines have been incorporated into a procedure which is provided in Attachment 1.
Criterion 5 - Chloride Analysis The Licensee shall provide for the analysis to be ccrnpleted within 4 days.
Response
Whereas our February 9, 1983 submittal stated that TMI-l chloride samples would be sent offsite for analysis, GPUN now intends to perform a chloride analysis onsite within 4 days. The analysis will be
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performed using an ion chromatograph. The undiluted sample would be stored in accordance with normal procedures for storage of high activity materials. GPUN expects to have the draft procedure available on site for inspection in January 1984 Criterion 6 - Radiation Level Methodology and Configuration The Licensee should provide a review of predicted personnel radiation levels under as-built conditions.
Response
I Attactinent 2 provides an evaluation of personnel radiation exposure and background levels in the counting room for as built conditions.
l Criterion 7 - Boron Analysis The Licensee should provide the analytical method for baron analysis and its measurement range.
Resoonse The boron sample analysis at TMI-1 employs a titration method with l
d-mannitol and sodium hydroxide to measure boron concentration over the range of 25-6000 ppm.
The mannitol titration is considered an industry standard method for analysis of Boron.
l 0856K l
2 After an initial pH adjustment, boron reacts with d-mannitol to produce an acidic complex. The free hydrogen ion released after d-mannitol addition is titrated with standard sodium hydroxide solution. For post accident samples the sample volume specified in TMI-l chemistry procedure N1904 is reduced from 5 ml to 1 ml to reduce personnel radiation exposure.
Criterion 8 - Chloride Shipping Casks The Licensee should indicate the offsite laboratory arrangements for chloride analysis and status of approval of a shipping container.
Response
As discussed in response to Criterion 5 above, GPUN intends to perfom post accident chloride analysis onsite. Therefore, no shipping container will be required.
Criterion 9 - Radiological Analysis and Radiation Background The Licensee should provide accuracy and range of radiological analysis capability and results of the present radiation background study.
Response
The range and accuracy of reactor coolant sample analysis is shown in Table 1 attached and will be confirmed through demonstrations described in the response to criterion 10 below. Results of the radiation background study are included in Attachment 2.
Criterion 10
" Cocktail" analysis Licensee should provide:
a.
The results of demonstrations to show the capability of performing the analysis in a post accident environment.
b.
Information on equipment calibration.
c.
Information on technician training.
Response
a.
GPUN will perform demonstrations of chemical analysis capability using TMI-l instrumentation and procedures on a standardized sample and provide the results of these demonstrations in February 1984.
b.
A surveillance / calibration program has been implemented to ensure a high degree of reliability for routine sampling and analysis based on manufacturers' recomendations and plant experience. The instrumentation used to fulfill post accident requirements is the same used for analyses during routine operations.
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3 l
c.
Training in post accident sampling, analysis, and transport-are included in the GPUN Chemistry / Rad Con Technical Training Program which provides for initial training of technicians as well as refresher training in a biannual requalification training program.
The need for more frequent training will be assessed in conjunction
~ ith critiques from quarterly emergency drills and annual emergency w
exercises.
r i
r r
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Table 1
~
REACTOR COOLANT SAbPLE ANALYSIS (1)
Analysis Ranoe Accuracy Gross Activity Note'2 Note 2 Gamma Spectrurn Range of gamma energies Within a factor of 2 across covering I131, 1133 the entire spectrum Ba140
- Csl37, Rul03, Csl34,l29m
, and Te doron 25 to 6000 ppm (3)
Within + 5% above 1000 ppm and +
50 ppm below 1000 ppm (1 ml sample)
.1 to 20 ppm (4)
Within + 10% between.5 and 20 ppm and witIiin +.05 ppm below.5 ppm Dissolved H2 or 4 to 2000 cc(STP)/kg(4)
Within + 20% between 50 and 2000 Total Gas cc/kg.and within + 5.0 cc/kg below 50 cc/kg Dissolved 02 Note 5 Note 5 pH (6) 1 to 13 Within +.3 pH units between pH 5 and 9 and within +.5 pH units for
~
all other ranges NOTES:
(1)
~~
Table 1 addresses Regulatory Guide 1.97 Revision 2 which as stated in m C of October 7, 1982 is recognized as a recemmendation and not a requirement.
(2)
Radiation dose rate to perst 'nel as det' ermined using a hand held radiac instrument in combination with information provided by spectral analysis should be sufficient to preclude the need for determination of gross activity. Therefore, to minimize personnel exposure, GPUN does not intend to measure the gross activity of the post accident sample.
(3)
While Regulatory Guide 1.97 Revision 2 specifies a range of 0-6000 ppm Baron content, Boron concentration below 25 ppm would not be expected for a post accident sample at TMI. Therefore, even though further analysis may indicate a lower range minimum, 25-6000 ppm Baron is judged to be a suitable range for post accident sample analysis at TMI-1. (See. Note 4)
.(4)
In.very chemical analysis technique there exists a lower limit of detectability.
Therefore, the zero minimum range shown' for Chloride and other primary coolant and sump analyses in Regulatory Guide 1.97, Revision 2 is not interpreted to be exact. The lower limits of detectability for chloride and dissolved H2 using the ion chromatograph are known to be.1 ppm and 4 cc (STP)/kg respectively.
(5)
E C letter of July 13, 1983 states that measurement,of either total dissolved gases or H2 gas in reactor coolant samples,is adequate and that measurement of dissolved 02 concentration is recommended but not mandatory.
Therefore, in order to minimize personnel exposure, GPUlI does,not intend to measure dissolved
^
02 in the post accident sample.
A residue of 10 cc/kg H2 is evidence of less than 0.1 ppm 0 -
2 (6)
Equipment for measuring pH of the undiluted post accident sample will be available in December 1983.
x
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PROCEDLRE FOR ESTIMATING CORE DAMT
\\
s 0856K t
i Al-0 4
._.________._m
':21.33 ORAFT 1.0 PURPOSE The curpose of'this calculation is to provice'an initial gross estimate to the exten; of core camage by gamma spectrum analysis following accident condi: ions.
2.0 REFERENCES
. 2.1.
TDR-431, Rev. O. Method for Estimating Extent of Core Damage Under Severe Accident Conditions.
3.0 PROCEDURE 3.1 Basis For Estimating Core Damage 3.1.1 The following NRC matrix shall be used in reporting the estimated degree of core damage utilizing specific radionuclide data and other plant parameters.
Degree ~of Minor Intermediate Major Degradation
(<
10%)
(10% - 50%)
(> 50%)
No Fuel Damage 1
1 1
G Cladding Failure 2
3 4
F Fuel Overheat 5
6 7
~
M Fuel Melt 8
9 10 The matrix consists of four general classes of damage and three degrees of damage within each of the classes except for the "NO FUEL DAMAGE" class.
3.2 Calculations 3.2.1 Gamma Scan Results I-131 uCi/.ml I-133 pCi/mi Ba-140 pCi/mi Cs-134' pCi/ml Os-137~ -
pCi/ml Ru-103 pCi/ml Te-129M -
pCi/ml t
Al-1
.5*
- *8 s rn..'. M.;
- r.*t s< e...., j
,s
.,e.
e.
rian- :arame ers EIS-TaVE =
'~ at es-'.ma ec ti.me cf faliere.
K-Densi:;. Cor ection Fac:cr # c-TABLE I assume sampie at 90*F.
V. - V 9A RCS (V.cs at 580 F is approxima:eiy 91,000 gai)
Power level at Failure (P )
1
- Corrected to operating T...
1, NOTE:
If power level changed by more than 10% in last 22-cays, record the following:
Old Power Level (P )
1 2
Time to make Change (Tc) hr + 2=
hr
, median time to make the change Time since completion of change to when core camage is expected to have occurred (f).
(f) hr t=
Tc
+f
+ 24 2
t - median time to make power change plus the time.
after the power change until damage is exoec-ted to have occurred, in cays.
l 3.2.3 Activity Correction Aet.:
(A,.i33Vi + A..,33V:
- A,.,33V.)(K)=
uCi/ml 33 =
+
Vecs As,.
3, (A. 23,Vi,A,.,3:V:
... + A. 33.V.)(K)= _
pCi/ml
=
v.cs i
As,..i.e (A.., eV
. A...eV 3
.. - A s...c V. ) (K).
uCi/mi
=
i v.es Al-2
':04.33 R; -
ATTACHMENT 3 (Cor.:'d)
A, - Ecuivaien: Nuclice Act vity i
A. - Activity of Each Sam:le ferrec:ec fcr density V. = Volume (Gal) of eacn Comconent Sam:lec corrected at operating temperature K = Volume Correction Factor 3.2.4 Inventory Correction a.
If Steady State Power Level is less than 100 percent X
100 X..i33 X,_isi = X...i.o
% Power b.
If transient condition > 10% existed in the past 22 days calculate X for each nuclide.
Xi.i33 100
=
-0.796t
-0.796t-Pie
+
Pe 3
X,_isi 100
=
-0.0864t
-0.0864t Pie
+
Pe 3
X...i.o
=
100
-0.0542t
-0.0542t Pie
+
Pe 3
X = Correcting Factor 3.2.5 Calculate Damage to Core a.
Gap Activity G = Percent of Rods with Ruptured Cladding Releasing Gap Activity G = [(1.863 ' x 10-2 )( Ae, i 3 3 )(X, _ i 3. )] - [(8. 31 x 10-')( Ae, _ i 3 3 )( X,. 3 3 )]
G=
Al-3
1004.22 DRAFT ATTACHMENT 3 (Cont'd) b.
Fuel Failure F = Percent of Rocs Overneated with Fuel Releasing Activity F ~ = [(1.684 x 10-')( Ae. i 3 3 )(X, - 13 3 )] - [(3. 54 x 10- 5 )( Ae,. 3 3 )(X. i 3, )]
F-luel M'elting c.
M = (0.002)(Ae...i..)(X....c)
M = Percent of Rods with Molten Fuel M=
=
==-
NOTE:
The absence of Ruthenium and Tellurium activity in the RCS and/or normal operating Cesium activity levels indicate that no fuel melting has occurred.
--==--
~w.
4 6=
1 Al-4
- g w----
g-,,
1004.33 DRAFT ATTACHMENT 3 (Cont'd)
TABLE 1 DENSITY. CORRECTION FACTORS FOR RCS EQUIVALENT VOLUME CALCULATION DENSITY CORRECTION FACTOR, K
- Reactor Coolant System RCS SAMPLE TEMPERATURE, *F
- Temperature at Time of Accident *F' 80 90 100 100
.996
.998 1
150
.983
.985
.987
-200
.966
.968
.970 250
.945
.947
.949 300
.921
.923
.924
~350
.894
.895
.897 400
.862
.864
.865 450
.827
.828
.830 500
.787
.788
.790 550
.739
.740
.741 560
.728
.729
.731 570
.717
.718
.719 580
.706
.708
.709 590
.693
.694
.695
- 600
.680
.681
.683 DENSITY CORRECTION FACTOR. K NOTE:
Normal RCS System Sample temperature is approximately 90*F.
Use this temperture if no other information is avaiable.
Al-5
ESTIMATED CCRE DIFAE GJ1DELINES INTRODUCTION l.0 The purpose of this report is to establjsh the basis for a prccecure to estimate the extent of ccre carage. The esticate of ccre carage thcule be as realistic as possible to dif ferentiate between nC Carage, ClaCCing failure, fuel cverheating, and ccre melt. Therefore, all musual events, such as high reatings from incore temperature thertocouples, ccntainment hydrogen mcnitcrs anc/cr high level raciatien mmitcrs must be docurented anc correlated with the racionuclice cata (athered when core damage is suspect.
2.0 Method for Estimating Core Dama;e 2.1 Care damage' estimates at best are cnly descriptive of core ccnditions and are not cefinitive cr finite.
2.2 Utilizing specific radionuclice data and other plsnt incications esticate the cegree of core camage with respect to the f cllowing bRC matrix.
l Degree of Minor Intermeciate Fajor l
l Decradaticn
(<10%)
(10% - 50%)
(>50%)
1 I
i l.Noluel Damage 1
1 1
l l Claccing Failure 2
3 4
l l Fuel Overheat 5
6 7
i l Fuel belt 8
9 10 i
l i
The matrix consists of four general classes of damage anc three cegrees of camace within each of the classes except fcr the NC Fuel Damace" class.
Ccnsequently, there are a total of 10 possible camage assessment categcries.
Fcr example, Categcry 3 woulc be descriptive of the condition where between 10 anc 50 percent of the fuel clacding has failed.
Note that the ccnditicns of mcre than one category could exist simultanecusly. The objective of the final ccre camage assessment methocclogy is to narrow ccwn, to the maximum extent possible, thcse categories which apply to the actual in-plant situation.
2.3 In orcer to conform to the above classificaticns of core carate, the follcwing assumptions are introcucec:
(a) The fuel gap fissicn prcouct inventory is assurec to be entirely released upon ruptere of fuel claccing (Class 2).
CE56K Al-6
(b) The total fission product inventory 'in the core in addition to the total fuel gap fission product inventory will be released when the fuel reaches higher temperatures indicative of fuel overheat (Class 3).
(c) _During fuel melt condition.s, 100% of the fission product inventory of the fuel gap and the fuel core, and 10% from
- me' ten fuel will be assumed to have been released into the coolant (Class 4).
2.4-The total fission product inventory in the core and fuel gaps at full reactor power for an equilibrium cycle of 930 EFPD is shown in
- Table 1 of Reference 4.4 for both TMI-II (Reference 4.1) and the extrapolated values for TMI-1._ Since the-total coolant weight at full power and temperature is around 478,423 lbm or 2.17 x 108 gms.
(Reference 4.2), then one can calculate the coolant concentrations for preliminary indications of core damage for 1% of the total fuel as shown'in Table 1.
Although, as a result of fuel melt, rare earth materials like strontium-90 and strontium-89'in addition to' barium-140 could be released into the coolant, the strontium isotopes are basically beta-emitters and connot be detected easily by gamma ray spectrometry. Therefore, in the 4
present analysis, barim -140 will be used as an indicator of fuel; melt with the asse ption that 10% of the barium is released into the coolant from molten fuel (Reference 4.4).
(,.
-Other radionuclides may be useful to aid in the evaluation of core 4 :
damage. For instance, the lac.k of Ruthenium and Tellurium-(specifically Ru-103 and Te-129M gamma emitters) is an indication that no fuel melting has occurred. A large amount of hydrogen
--re eased to conta nment indicates that possible cladding failure l
i (zircoloy oxidation) has occurred and core thermocouple readings
- above-23000F indicate that both cladding failure and fuel overheat j
has taken place (Reference 4.3).
3.0 RECOMMENDED PROCEDURAL STEPS FOR ASSESSMENT OF CORE DAMAGE (Ref. 4.4) p 3.1 Chemistry should be notified by Operations to sample the RCS for gamma spectroscopy at any time plant warning devices indicate core damage. Some examples of these are the RCS letdown high radiation
- alarm (RM-L1), the Reactor Building sump area high radiation alarm (RM-G21), or any containment area high radiation alarm. A sample should be drawn at any time core damage is suspected.
l NOTE 1:
Chemistry must implement the post accident sampling procedure and a plan of attack to protect technicians from overexposure of radiation.
c sis on the RCS for Performthenecessarygammasoectrogo'opicanal{29M, 3.2 1131, Il33, Bal40, CslM, Csl37, Ru
, and Te Al-7 0856K l-c-
,.v y,
,m.,
3.3 If the Reactor Coolant System has been breached to the Containment Building and/or Primary Auxiliary Building, samples of water from these sumps should also be obtained and counted, if possible, to approximate radionuclide inventory of these volumes.
3.4 After obtaining the results of the garr.a spectroscopy, calculate the percent cladding failure:using the following equation:
Percent of rods with cladding fa.tlure releasing Gap Activity = G G=[(1.863x10-2)(A _131) X131] - [(8.31x10-3)(A _133) X133))
I I
131 A -131 = Equivalent Core 1 Activity Concentration Where:
I A _133 = Equivalent Core 1133 Activity Concentration I
X = Nuclide Inventory. Correction Note 2:
The total inventory of 1131 and Il33 shall be calculated.by using Attachment "A" which also corrects the activity to equivalent reactor core coolant system concentrations.
Note 3:
The value for X'can be obtained using Attachment "B".
Use Equation No, l ~for when core damage was at a steady i state power level and Equation No. 2 for when core damage was during a transient power level.
~~ The amount of fuel cladding rtpture may also coincide with incore thermocouple temperature readings in excess of 2000oF, abnormally high concentrations of hydrogen and radioactive Xenon and Krypton in the Containment Reactor Building atmosphere if the RCS pressure boundary is breached, loose parts monitor alarm in the RCS, or f
sudden pressure changes-(hydraulic shock) in the RCS.
3.5 To calculate the percent of fuel overheating which has taken place,
-use the following equation..
Percent of fuel which has been overhtsted = F F = [(1.684x10-4)(A _133) X _133] - [(3.64xlO-5)(A _131) X _131)]
I I
I I
133 A _133 = Equivalent Core 1 Activity Concentration When:
I l31 A _131 = Equivalent Core I Activity Concentration I
X = Core Nuclide Inventory Correction See Notes 2 and 3 above.
Fuel overheating will also coincide with core thermocouple temperature readings in excess of 23000F.
Al-8 0856K
3.6 To calculate the' percent of core melting which has taken place, use the following equation:
Percent of core melt = M M = f(0.002)(ABa-140)) X Where:
ABa-140 = Total Activity of Released Barium-140 X = Core Nuclide Inventory Correction Note 4:
The total inventory of Barium-140 can be calculated using Attachment "A" which also corrects the activity to equivalent Reactor Core Coolant System concentration.
See Note 3 In addition to Barium-140, other gamma emitting nuclides such as Cesium-134, -137, Ruthenium-103 and Tellurium-129M may be released in various concentrations into the coolant during core melting.
Thus, the lack of Ruthenium and Telluri s in the Reactor Coolant and/or normal operating Cesi s activity levels indicate that no fuel melting has' occurred.
~,
1 Al-9 1
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4.0 ' REFERENCES 4.1 "Three Mile Island Nuclear Station Unit II, Final Safety Analysis Report", Appendix 15A.
'4.2 "Three Mile Island Nuclear Station Unit I, Final Safety Analysis Report" _ (Updated Version), Voltrae 2, Chapter 4.
4.3 Rogovin Report, Part 2, Vol. II.
4.4 TDR-431'.Rev. O, Method for Estimating Extent of Core Damage Under
. Severe Accident Conditions.
e
.n.
~~
l l
Al-10 0856K
TfeLE 1 CDOLANT CGNCENTRATIONS FCR FRELIMINFRY ltOICATION OF CCRE D/FAGE I
l-l l
li of Fuel 1% of GAF l
Half Garrrna Ray Inventcry Invmtcry I
1 Nuclice Life Intensity (kev)
(u ti/ ara)
(u ti/cm) l I
I I
I-131 6.04d 364 2533 59.42 I
I l
I-133 20.8h 530 6574 12.65 I
I I
I-13 5 6.59h 1132 5964 4.06 I
.I l
l Xe-133-5.25d 81 SSCO 3E4.34 I
I
- I KI-65 2.64h 156 2722 4.00 I
1 i Ba-140 12.ec 53 7 500 I
I
~-
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TABLE 2 DENSITY CORRECTION FACTORS FOR RCS EQUIVALENT-VOLtNE CALCULATION DENSITY CORRECTION FACTOR, K l
I l
-l Temperature at Time of I l
l Accident 0F l
80 90 100
'l i
I I
I I
I l
100 l
.9%
.998 1
l l
150 l
.983
.985
.987 l
l 200 l
.966
.968
.970 l
l 250 l
.945
.947
.949 il I
300 l
.921
.923
.924 l
l 350
.I
.894
.895
.897 l
l 400 l
.862
.864
.865 l
l 450 1
.827
.828
.830 ' l l
500 l
.787
.788
.790 l
l 550 l
.739
.740
.741 I
l 560 l
.728
.729
.731 l
l
'~
570 l
.717
.718
.719 l
580 l
.706
.708
.709 l
590 l
.693
.694
.695 l
l 600 l
.680
.681
.683 l
l l
l l.
l Density Correction Factor, K l
l 1
i NOTE: Normal RCS System Sample temperature is approximately 90oF. Use this temperature if no other information is available.
Al-12 0856K 1
. ATTACENENT A CALC 1)LAT10h OF EOJIVALENT REACTOR CORE COOLANT $YSTER CONLENTRAT10h5 The equations cerivec in the previcus section receires that the nuclice activities be known at full pcwer and temperature ccnciticns. In the case of any 1 css of coolant frce the core, it is essential tc calculste the total ccre
--inventcry of the nuclice activity by utilizing the fellcwing ecuivalent concentration relationship:
V AVii+AV22+AV3 3 +
+Ai i b_-
(K)
%CS M1ere :
Ae = Equivalent nuclide activity Ai = Activity of each sample Vi = Vclume of each component sampled correctec for temperature (Table 3)
VRCS = Volume of reactor coolant system at full power and temperature
. K = Volume correction e
To correct for the vclume of each sample, Table 2 coulc be utilizec to calculate the appropriate volume under full power anc temperature conditions at the time of an accicent.
EXAMPLE a.
Ccre damage is suspected after a large reactor coolant leak has tillec containment sump only.
b.
RCS temperature at time of accicent 450T.
c.
Sample temperature 50c K = 0.628, Table 2 c.
Volume RCS - 70,000 gal.
(Activity 1131 = 2.3 x 102 uCi/ml) i e.
Vclume Centainment Stop - 8,660 gal.
( Activity 1131 = 1.8 x 101 uti/c1)
Al1+AV22 (K450)
Ae=
VRCS Ae =
[(2.3x102 uti)(70,000 gal)(3765 ml )) + [(1.8x101 uti)(6,660 gal)(37E5 al )3(.62E) g2
- c. al r
ca I-(70,000 gal)(37ES
)
Ae =
(6.09x1010 + (5,goxic8)
(.826) 2.65 x 106 2
Ae = 2.32 x 10 pCi/ml for 1-131) - (.E2E) 2 Ae = 1.52 x 10 uC1/ml CE36K Al-13 L
ATTACH 4ENT B hUCLIDE INVENICRY CORRECTION FOR REDUCED FGER 0PERAT10N 1
Steacy State Concitjen To correct for Nuclice Inventory activity if fuel camage is stspectet to have occurrec during times of reouced power levels, (except 0%) where the level has not chmoed creater thm + 10% of the recucec level witnin the _last 22 cays, then'the' activity level coulc be ccrrectec by multiplying it by the follcwing factor.
100 X=
% Full Power at Time of Failure Example: : Plant operating at 50% pcwer for 30 cays Ccrrection Factor = X X = ICO = 2 50 2.
Transient Condition On the other hmd, in croer to correct for nuclice inventory activity if fuel camage is suspected to have occurred at tires other than those which fit situation #1 abcve, use the follcwing equation.
100 x
Old Pcwer level in % (e-AI) + new power level in % (1-e-At) hhere:
Old pcwer level =
% of full pcwer befcre the power change few power level =
% of full power after the power change at which tirre the fuel failure has occtrrec cecay constant (eousls 0.0E64 cay-1 fcr 1 =
1-13 1 0.7% cay-1 fcr 1-133 mc 0.0542 cay-1,for Ea-140 t=
mecian time to take power change plus the time af ter the pcwer change tntil camage is suspected to have cccurred, in cays EX/41PLE:
Reactcr pcwer was recuced from f ull pcwer tc 10%. This pC%er change took four (4) hcurs, anc eicht (6) hcurs later ccre camage is suspectec cbe to an alarm to RM-L1, I
4 6)+ 24 tole:
t=(
+
2 t = 0.42 LOG y
(ICO)e-(.0EM) (. 42)
(y e)[ y _e -(.0Es M. 42)]
113 1 X 1133 = 1.03 OE56K Al-14
POST-ACCIDENT SAMPLING
~~
RADIOLOGICAL ANALYSIS A2-0 0856K
I.
INTRODUCTION I
The purpose of 'his report is to evaluate TMI-l onsite radiological t
conditions during post-accident sampling and associated spectral analyses in terms of personnel radiation exposures and shielding requirements for compliance with the radiation criteria set by NUREG-0737, II.B.3.
Specifically:
1.
Prediction of personnel radiation exposures based on person-motion for sampling, transport and analysis of all required parameters.
2.
Analysis of the background levels in the Counting Room.
Reactor coolant samples are to be obtained within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (from the time a decision is made) without exceeding the limits of radiation criteria in 10 CFR Part 50, App. A, GDC 19 (i.e., 5 Rem whole body, 75 Rem extremities).
TMI-l Plant Procedure No. 1004.15, Rev. 2 further reduces these radiation limits to 3 Rem whole body and 18 3/4 Rem to the extremities. Post-accident sampling activities consist of the operation of valves and taking liquid and gas samples in the sample room, dilution of the liquid sample in the radio-chemistry laboratory, and analyzing the diluted samples in the counting room. These tasks will be performed using several technicians (5-6) to reduce individual personnel exposure.
Radiation-source term data used in this report are those stipulated in Regulatory Guides 1.3 and 1.4 (Table 1). Radiation dose rates are analyzed using computer code ISOSHLD-II (ref. 1), which employs a point kernel integration method.
II.
METHODS AND DATA A.
Source Term Data 1.
Source term data in this report were taken from Table 3.1-2 of Ref. 2, which is consistent with Regulatory Guides 1.3 and 1.4.
2.
All sample line pipe sources were conservatively assumed to contain zero-time decayed reactor coolant activities (Table 1).
3.
The source term for the liquid sample bottle was assumed to be depressurized (no noble gases) and 30-minute decayed.
4.
The liquid sample source term in the Counting Room - 2 cm3 of diluted liquid sample with dilution factors of 104 and 7
10.
5.
The extremity dose rate is from a syringe which contains 1 cc of 30-min decayed noble gas source.
0856K A2-1 1
I
A syringe is used to draw 1 cc of noble gases from a 300 cc bottle holding noble gases transferred from the 65 cc primary coolant liquid sample bottle.
B.
Dose Acceptance Criteria 1.
10 CFR Part 50 Appendix A, GDC 19:
a.
5 Rem Whole Body b.
75 Rem Extremities 2.
TMI-l Plant Procedure 1004.15, Rev. 2 (used in this report) a.
3 Rem Whole Body b.
18 3/4 Rem Extremities C.
Post-Accident Samo11ng Activity Scenario (Figure 2) 1.
All technicians wear Complete Breathing Apparatus for entering sample / counting rooms (Figure 1).
2.
Technician 1 spends 25 minutes in the Sample Room establishing the valve lineup (all the sample lines with no primary coolant yet). Then Technician I spends 2 minutes opening CAV-110 (Figure 2) in the vacinity of sample lines which contain primary coolant.
3.
Technician II spends 4 min. in the Sample Room taking the liquid sample, then leaves the room.
4.
Technician III takes the gas sample for 10 minutes.
Technician III is assumed to have a 30-second extremity contact exposure to the 1 cc gaseous sample. A syringe is used to extract 1 cc of gaseous sample from the 300 cc bottle containing noble gases drawn from the liquid sample.
5.
Technician IV spends 15 minutes in the Sample Room restoring the sampling system.
This part of the procedure could be completed later.
D.
Radiation Source Geometry 1.
Sample line pipe containing primary coolant is shown in Figures 2, 3, and 4.
2.
All sampling line pipe is 3/8" diameter and 0.065" wall thickness.
3.
Individual Sour.:e Description a.
S123 =
3 pipes 30" in length (see Figures 3 & 4) shielded by 4" of lead M-2 0856K
b.
S45 2 pipes 7 ft in length sheilded oy 2" of lead
=
c.
S67 Same as S45 except pipe length is 10 ft
=
d.
S89 2 pipes 3 ft in length shielded by 2" of lead
=
e.
S10 Same as S89 except pipe length is 10 ft
=
f.
S11 Cooler source with 2" lead shielding (Ref. 3)
=
E.
Airborne Activity in the Counting Room 1.
Atmospheric dilution Factor (X/Q) was based on the diffused source - point receptor configuration X/Q = 3.44 x 10-3 sec/m3 2.
Wind Speed of 0.782 m/sec and Pasquill Type "F" e
3.
Finite cloud model in the Counting Room F.
Comouter Code - ISOSHLD II (Ref. 1) which employs a point kernel integration method.
G.
Dose Rate Results 1.
Radiation Exposure to Technician I
~
a.
16.44 Rem /hr from Source S45 (See Section II.D) b.
0.895 Rem /hr from Source S123 c.
6.43 Rem /hr from Source S10 Total exposure to Technician I during 2 min. stay time in the nuclear sampling room.is (TDl):
TD1 = (16.44 + 0.895 + 6.43) x 2/60 = 0.79 Rem 2.
Radiation Exposure to Technician II a.
18.86 Rem /hr from Source S45 b.
5.09 Rem /hr from source S67 c.
1.13 Rem /hr from Source S11 Total radiation exposure to Technician II during 4 min. stay time (TD2):
TD2 = (18.86 + 5.09 + 1.13) x 4/60 = 1.67 Rem 3.
Radiation Exposure to Technician III A2-3 0856K
a.
10.4 Rem /hr from S45 b.
0.895 Rem /hr from 5123 c.
6.43 Rem /hr from slo Total 10 min. radiation exposure (TD3):
TD3 = (10.4 + 0.895 +6.43) x 10/60 = 2.95 Rem 4
Contact dose rate from the syringe containing I cc of 30-min.
decayed noble gases:
Ds = 143 Rem /hr With 30-second contact time D = 143 x 30/3600 = 1.2 Rem 5.
Contact dose rate from a 2 cm3 of liquid sample 7
(DR = 10 ).in the Counting Room = 0.21 mrem /hr e
6.
The maximum airborne activity in the Counting Room results in a dose rate 'of 160 mrem /hr.
In order to assure the + 50% accuracy of the sample analysis,'
the background radiation (160 mrem /hr) has been reduced to about one half of the 2 cm3 liquid sample dose rate (0.5 x 0.21 = 0.105 mrem /hr) by providing a shield for the GeLi
> ~-
~
detector.
A2-4 0856K i
L_
IV. -
SUMMARY
OF RESULTS 1.
Radiation expesure to each Technician during post-accident sampling activities is as follows:
Radiation Radiation Exposure
. Technician I.D.
(Rem)
Time Duration (Min.)
1 0.8 2
II 1.7 4
III 3.0 10 2.
The extremity dose received by a Technician handling the post-accident noble gas syringe for 30 seconds is approximately 1.2 Rem.
A2-5 0856K
1
.1 1
IV.
-REFERENCES-
- 1.
BNWL-236, UC-34, " User's Manual for ISOSHLD code," June 1966
- 2... GPU TDR No.~183,.Rev. 3, April _21, 1981 j-
' 3.
GPU CalCJlation No N1779-5412-005,. February 4, 1983 I
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TABLE 1 SHIELDING S0WCE TEFF.S (T=0)
Liquio(1)
Gasecus(2)
Source Source Reactor Coolant Activity Activity Ccncentration
' Isotope (C1)
(Ci)
- (uti/cc)
Br-64 7.85 + 6 3.93 + 6 2.43 + 4 Kr-83m 9.25 + 6 9.26 + 6 2.67 + 4 Kr-65m 2.19 + 7 2.19 + 7 6.79 + 4
'Kr-85 5.30 + 5 5.30 + 5 1.64 + 3 Kr-67 4.00 + 7 4.00 + 7 1.24 + 5 Kr 5.60 + 7 5.60 + 7 1.74 + 5 Rb-85' 5.64 + 5 1.75 + 3 SI-89 7.42 + 5 2.30 + 3 Sr-90 3.99 + 4 1.24 + 3 Sr-91 9.72 + 5 3.01 + 3 Sr-92 9.50 + 5 2.94 + 3 Y-90 3.96 + 4 1.23 + 3 Y-91 9.85 + 5 3.05 + 3
-Mo-99 1.28 + 6 3.97 + 3 Ru-106 2.29 + 5 7.10 + 2 Xe-131r
'4.38 + 5 4.3 8 + 5 1.3 6 + 3 Xe-133m 3.07 + 6 3.07 + 6 9.51 + 3 Xe-133 1.27 + 8 1.27 + 8 3.93 + 5 Xe-135m 3.26 + 7 3.26 + 7 1.01 + 5 Xe-135 2.09 + 7 2.09 + 7 6.48 + 4 Xe-138 1.17 + 8 1.17 + 8 3.63 + 5 I-131 3.68 + 7 -
1.84 + 7 1.14 + 5 I-132 4.31 +~7
-2.16 + 7 1.34 + 5 I-133 6.40 + 7 3.20 + 7 1.98 + 5 1-134 8.00 + 7 4.00 + 7 2.48 + 5 1-135 6.35 + 7 3.18 + 7 1.97 + 5 Cs-134 1.27 + 4 3.93 + 1 Cs-136 8.02 + 3 2.48 + 1 Cs-137 4.99 + 4 1.55 + 2 Cs-138 -
1.23 + 6 3.81 + 3 Ba-137m 4.67 + 4 1.45 + 2 3.87 + 3 Ba-140 1.25 + 6 La-140 1.27 + 6 3.93 + 3 Ce-144 7.50 + 5 2.32 + 3 Cr-51 5.20 - 3 5.80 - 4 hn-54 Mn-56 1.70 - 2 5.60 4
Fe-59 3.00 - 2 Co-58 4.00 - 3 Co-60.
Zr-95 5.00 4
(1) Basec on 100% noble gas core inventory, 50% halccen core inventory, ano L
1% of all other core inventory.
(2) Basec en 100% noble gas core inventory anc 25% nalegen core inventcry.
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