ML081000627

Massachusetts Institute of Technology Research Reactor, Response to Request for Additional Information 2/22/08

ML081000627
Person / Time
Site:	MIT Nuclear Research Reactor
Issue date:	02/22/2008
From:	Bernard J, Hu L Massachusetts Institute of Technology (MIT)
To:	Pierce S NRC/NRR/ADRA/DPR/PRTB
PIERCE S, NRR/DPR/PRTA 415-2261
References
TAC MA6084
Download: ML081000627 (115)

Text

MIT NUCLEAR REACTOR LABORATORY AN MIT INTERDEPARTMENTAL CENTER John A. Bernard Mail Stop: NW12-208a Phone: 617 253-4202 Director of Reactor Operations 138 Albany Street Fax: 617 253-7300 Cambridge, MA 02139 Email: bemardj@mnit.edu February 22, 2008 U.S. Nuclear Regulatory Commission Attn: Stephen Pierce,,012-G15 Research and Test Reactors Branch B Division of Policy and Rulemaking Office of Nuclear Reactor Regulation Washington, DC 20555 Re: Request for Additional Information; License No. R-37; TAC No. MA 6084

Dear Mr. Pierce:

Enclosed is the information requested pursuant to the NRC letter of 30 November 2007:

a) The redone Chapter 13 calculations are documented in a file memo dated 29 April, '2003 (copy enclosed.)

b) The dose calculations were performed as part of a MS thesis, "Estimate of Radiation Release During Design Basis Accident," by Qing Li. A copy is enclosed.

Sincerely, Lin-Wen Hu

/§hnABernard I declare under penalty of rJury that the foregoing is true and correct to the best of our knowledge.

Executed ~

Date Signature cc: Senior Project Manager (without enclosures)

Document Control Room (without enclosures)

JAB/koc 7 JT Enclosures

A-ppendix D Additional Information for Response to Question 92(b)

-File Memo dated 29 April 03, "Loss of Primary Flow Transient Analysis" 47

NUCLEAR REACTOR LABORATORY AN INTERDEPARTMENTAL CENTER OF *6 V UMASSACHUJSETTS INSTITUTE OF TECHN'OLOGY P,

LIN-WEN HUL 138 Albany Street, Cambridge, MA. 02139-4296 Activation Analysis Reactor Relicensing Engineer Telefax No. (617)253-7300 Coolant Chenmistry Telephone No. (617)258-5860 Nuclear Medicine Email: lwhu@iniit.edu Reactor Engineering MEMORANDUM TO: IvITR Files FROM: Lin-Wen Hu 6ý DATE: April 29, 2003 RE: Loss of Primary Flow Transient Analysis (2)

1. The loss of primary flow transient analysis was originally performed using initial conditions of reactor power 6.1 MWf, primary flow 2000 gpm, coolant outlet temperature of 55 'C, and coolant height at 10 ft (LOF case# 1). This analysis was repeated using the LSSS as the initial conditions (LOF case#2). The LSSS for the MITR-Ill are: reactor power 7.4 MW, primary flow 1800 gpm, coolant outlet temperature 60 *C, and coolant height at 10 ft. The MULCH-II code was used for both analyses. All other assumptions are the same for both analyses.

2. Figures 1 and 2 are comparisons of the coolant outlet temperatur es of the average and hot channels for the two cases. Note that the initial coolant temperatures are higher in Figure 2 because of the higher initial power (7.4 MW. v.s. 6.1 MW) and lower initial flow rate (1800 gpm v~s. 2000 gpm). The peak hot channel outlet coolant temperattires, which occur around 1.5 s into the transient, are 105.2 *C for case#2 and 97.0 'C for case#l.

Note that the coolant temperature then decreases rapidly in both cases because of reactor scram. Both analyses showed that the hot channel coolant outlet temperature would reach saturation after about 15 to 20 seconds. Figure 3 shows the calculated fuel temperatures at the average and hot channel outlet assuming the initial conditions of the LOF transient are LSSS. The calculated fuel temperatures are well below the cladding softening point of 450 0 C.

3. Figure 4 is the calculated reactor decay power assuming equilibrium reactor power was at 7.4 MWf before scram. The reactor decay heat at 16 seconds after reactor scram is about 325 kW. As shown in SAR section 4.6.6.3, the best-estimate dry-out condition is 468 kW.

4. .The MiULCH-il output file for LOF case#2 is attached to this memo.

110 1

.- L -- - - - - - -

100 --- - . . . . . . . - -- ----

50 0 10 20 30 40 50 Time (s)

Figure 1. Coolant outlet temperatures of average and hot channels during a loss of primary flow transient, The initial conditions used for this analysis are reactor power at 6.1 MW, primary flow 2000 gpm, coolant outlet temperature 55 "C, and coolant height at 10 ft.

110 10 . . . . . - - -- - - - - - - - - - - - - - - -- - - - - - - -

60 '-- ----- - - ------------- ----

60 - -Average channel 50 --- Hot channel 0 10 20 30 40 50 Time (s)

Figure 2. Coolant outlet temperatures of average and hot channels during a loss of primary flow transient. The initial conditions used for this analysis are reactor power at 7.4 MW, primary flow 1800 gpm, coolant outlet temperature 60 'C, and coolant height at 10 ft.

140 I 12 ----- :.... -

60 -- -------

Average channel fuel temperature 40 10 - - -Hot channel fuel temperature 0 0 20 30 40 50 Time (s)

Figure 3. Fuel temperatures of average and hot channels at outlet during a loss of primary flow transient. The initial conditions used for this analysis are reactor power at 7.4 MWf, primary flow 1800 gpm, coolant outlet temperature 60 'C, and coolant height at 10 ft.

2 00 - ----- ....... I---- ----------------------------

0 10 20 30 40 50 60 70.

Timne (s)

Figure 4 Reactor decay power calculated using DKPOWR assuming equilib rium power of 7.4 MW before reactor scram.

1of multi-channel Analysis code, MULCH-II MIT Nuclear Reactor Laboratory 7/15/1996 LOSS OF FLOW PREDICTION FOR t4ITR-1iI best estimate Reactor Power (kw)= 7400.00 cooling Tower outlet Temp Cc)= 13.00 Primary Flow (kg)=- 111.00 Secondary Flow (kg)= 103.00 cooling Tower Efficiency= .80 Reference Temp(ýC)= 50.00 coolant height from air/water interface to top of flow guide (m)= 2.31

• • simulated case is -- > LOSS OF PRIMARY FLOW steady-state operation before shutdown for ****** hours Time step (S)=.100E+00 Total Simulation Time (s)= 50.00 instrument Delay Time (s)= 1.00 80% Blade Insertion Time (s)= 1.00 Pump coastdown curve:

(exp(-1.870+ .410*t/10+a 2.950*exp(t/10)+ -.680*exp(-Ct/10)A2))- .514)/C 1.492- .514)

Loop Component Geometries:

I Aflow~mA2) Vol CmA3) De(m) .dz(m) Kform Nchan 1 .320E-01 .427E+00 .203E+00 -7.08 4.58 1 2 .389E-04 .16BE-03 .704E-02 00 7.30 17.70 3 .320E-01 .468E+00 .203E+00 6.97 2.17 1 4 .339E+00 .413E+00 .180E+00 -1.22 .00 1 5 .111E+00 .760E-01 .630E-01 - .69 .30 1 6 .440E-02 .160E-01 .220E+00 -.01 .18 1 7 . 290E-01 . 180E-01 . 400E-01 -.61 .00 1 8 .125E-03 .824E-04 .219E-02 .66 2.05 345 9 .130E+00 .990E-01 .387E+00 .76 .00 1 10 .923E+00 .192E+01 .108E+01 1.22 .00 11 .320E-01 .427E+00 .203E+00 -7.08 4.58 177 12 .900E-04 .389E-03 .301E-02 .00 7.30 1 13 .320E-01 .468E+00 .203E+00 6.97 2.17

• Anti-siphon and Natural convection Valve Geometries:

Acont~mA2) Aref(mA2) Vball~mA3) Rball(kg/mA3) KUp Kdown NV ASV .;178E-02 .384E-02 .1059E-03 2715.00 7.90 6.90 2 NCV .271E-02 .811E-02 .2040E-03 2715.00 4

.Fraction of coolant cooling the fueled region= .920 SIX Fouling factor (c mA2/W)= .3500E-03 Fraction of energy deposited in fuel= .910 coolant= .054 D2o= 021 Graphite= .015 Hot channel Factor= 2.000 bottom ----------------------------------------------- >top 2 3 4 5 6 7 8 9 10 11 shape...avg 1.000 1.900 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 shapejiot 1.000 1.000 1.000 1.000 1.000.1.000 1.000 1.000 1.000 1.000 Peak...avg 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 Peak-hot 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 1.000 minimum flow distribution in flow channel= .8640 Engineering Factors for:

Reactor Power= 1.000 Hot Channel Flow Rate= 1.000 Heat Transfer Coef= 1.000 Hot Spot Heat flux-- 1.000 Min CHF ratio= 1.500 Min DNB ratio= 1.500

---

END OF INPUT--------------------

---

---- START OF OUTPUT -------------------

DP...core= 41528.180000 DPrati 01= -2.104 542 E-01 Ssteady-state temperatures for each components

• 1 2 3 4 5 6 7 8 9 10 11 12 13 Page 1

-iof 59.0 43.7 43.7 43.7 43.7 43.7 43.7 59.0 59.0 59.0 13.0 29.5* 29.5

• • steady-state temperatures for core region

• Tw-hot=coolant temperature at the hot channel Tw..avg=coolant temperature at the average channel Tc-hot=clad temperature at the hot channel Tc...avg=clad temperature at the average channel Tfjlot=fuel temperature at the hot channel Tf..avg=fuel temperature at the average channel 1 2 3 4 5 6 7 8 9 10 11 12 Tw..hot 43.7 47.5 51.4 55.2 59.1 63.0 66.8 70.7 74.5 78.4 82.2 82.2 TW..avg 43.7 45.3 47.0 48.7 50.3 52.0 53.6 55.3 57.0 58.6 60.3 60.3 Tc.Jiot 43.8 68.7 71.8 75.1 78.2 81.6 85.1 88.6 92.2 9S.8 99.4 82.4 Tc...avg 43.8 55.2 56.7 58.2 59.7 61.3 62.9 64.4 65.9 67.4 69.0 60.4 Tf-.hot 43.7 77.0 80.2 83.4 86.8 90.1 93.6 97.1 100.7 104.2 107.9 82.2 Tf..avg 43.7 59.1 60.6 62.1 63.6 65.2 66.7 68.2 69.8 71.4 72.9 60.3.

Qfluxji .1000E-04 .3234E+06 .3234E+06 .3234E+06 .3234E+06 *.3234E+06 .3234E+06 .3234E+06 .3234E+06

.3234E+i06 .3234E+06 .1000E-04

• • cladding Tern erature at ONB and CHFR

• 13 4 5 6 7 8 9 10 11 12 TONS-hot 107.5 114.3 114.3 114.3 114.3 114.3 114.3 114.3 114.3 114.3 114.3 107.5 TONB~avg 107.5,112.4 112.4 112.4 112.4 112. 112.4 112.4 112.4 112.4 112.4 107.5 CHFR 999.9' 9.1 9.1 9.1 9.1 9.1 9.1 9.11 9.1 9.1 9.1 999.9 LSSS OK!

Safety Limits OK!

ý** iSSS and Safety Limits Index

• 0: below limit 1: limit exceeded 1 2 3 4 6 7 8 9 10 12 LSSS 0 0 0 0 0 0 0 0 0 0 0 0 SL 0 0 0 0 0 0 0 0 0 0

0. OOOOOOE+00 102.175500 0. OODOOOE+00 *0.000000E+00 scram signal sent at: 1.OOOOOOE-01(s)

Blades 80% inserted at: 2.100000(s) 1.OOOOOOE-01 98.917020 0.OOOOOOE+00 0. OOOOOOE+00 2.OOOOOOE-01 92.767810 0.OOOOOOE+00 0.OOOOOOE+00

3. OOOOOOE-01 87.031840 0. OOOOOOE+00 .0.OOOOOOE+00

4. OOOOOOE-01 81.677700 0 .OOOOOOE+00 0. O0OOOOE+00

5. OOOOOOE-01 76.676730 0.000000 E+00 0. OOOOOOE+00

6. OOOOOOE-01 72.002690 0. OOOOOOE+00 0. OOOOOEi-00

7. 000000E-01. 67.631540 0. OOOOOOE+00 0.600000E+00 8.00000O1E-01 63.541280 0. 000000E+00 0. OOOOOOE+00 9.OOOO1E-01 59.711720 0.000000OE+00 0.000000OE+00 1.000000 56.124330 0.OOOOOOE+00 0.OOOOOOE+00 1.100000 52.762060 0.OOOOOOE+00 0. OOOOOOE+00 1.200000 49.609260 0.OOOOOOE+00 0.OOOOOOE-t00 1.300000 46.651530 0.000000E+00 0. OOOOOOE+00 1.400000 43.875560 0.OOOOOOE+00 0. OOOOOOE+00

1. 500000 41.269100 0.OOOOOOE+00 0.OOOOOOE+00 1.600000 38.820880 0.OOOOOOEi-00 0.OOOOOOE+00 1.700000 36.520440 0.OOOOOOE+00 0.OOOOOOE+00 1.800000 34.358120 0.OOOOOOE+00 0.OOOOOOE.00 1.900000 32.325000 0.OOOOOOEi-00 0.OOOOOOE+00 2.000000 30.412800 0.OOOOOOE+00 0.OOOOOOE-e00 2.100000 28.613860 0.OOOOOOE+00 0.OOOOOOE+00 2.200000 26.921060 0.OOOOOOE+00 0.OOOOOOE+00 2.300000 25.327790 0.OOOOOOE+00 0. OOOOOOE+00 2.400000 23.827910 0.OOOOOOE+00 0. OOOOOOE+00 2.500000 22.415700 0.OOOOOOE+00 0.OOOOOOE+00 2.600000 21.085840 .0.OOOOOE+O0 0.OOOOOOE+00 2.700000 19.833370 0.OOOOOOE+00 0.OOOOOOE+00 2.799999 18.653680 0.OOOOOOE+00 0. OOOOOOE+00 2.899999 17.542420 0.OOOOOOE+00 0.OOOOOOE+00 2.999999 16.495570 0.OOOOOOE+00 0. OOOOOOE+00 3.099999 15.509360 0.OOOOOOE+00 0.OOOOOOE+00 3.199999 14.580240 0.OOOOOOE+00 0. OOOOOOE+00 3.299999 13.704900 0.OOOOOOE+00 0. OOOOOOE+00 3.399999 12.880240 0.OOOOOOE+00 0.OOOOOOE+00 3.499999 12.103330 0.OOOOOOE+00 0.OOOOOOE+00 3.599999 11.371440 0.OOOOOOE+00 0. OOOOOOE+00 3.699999 10.682000 0.OOOOOOE+00 0.OOOOOOE-i00 3.799999 10.032570 0.OOOOOOE+00 0.000000E+00 3.899998 9.420874 0.OOOOOOE+00 0. OOOOOOE+00 3.999998 8.844763 0.OOOOOOE+00 0.000000E+00 Page 2

lot 4.099998 8. 302208 0 .OOOOOOE+00 0 .OOOOOOE+00 4 .199998 7.015791 0 .000000E+00 2 .003114E-01 4.299998 6.504248 1. 587088E-01 1. 942458E-01

4. 399998 5. 977942 2 .753256E-01 1.878818E-01 4.499998 .5. 510567 3.5404 89E-01 1. 824389E-01

4. 599998 5. 095648 4 .006996E-01 1.775728E-01 4.699998 4. 726803 4. 209365E-01 1.7308 59E-01 4.799998 4. 398181 4.198132E-01 1. 688775E-01 4.899998 4. 104669 4.015948E-01 1.649104E-01 4.999998 3.841931 3. 697387E-01 1.611942 E-01 5.099998 3. 606340 3. 269660E-01 1. 577492E-01 5.199997 3. 394952 2 .753318E-01 1. 546045E-01 5.299997 3. 205383 2 .163307E-01 1. 517894E-01 5.399997 3. 035687 1. 510319E-01 1.493479E-01 S.499997 2. 884278 8.015008E-02 1.473253E-01

5. 599997 2. 749908 4.106613E-03 1. 457715E-01 5.699997 2.631436 -7. 659547E-02 1. 445610E-01 5.799997 2.526958 -1.60302 3E-01 1. 433120E-01 5.899997 2. 434266 -2 .452841E-01 1.419336E-01 5.999997 2.351322 -3 .299991E-01 1.403838E-01 6.099997 2. 276308 -4 .131188E-01 1. 386381E-01 6.199996 2. 207646 -4. 935606E-01 1. 366928E-01 6.299996 2. 144015 -5.704918E-01 1. 345587E-01

6. 399996 2. 084339 -6. 433241E-01 1. 322558E-01

6. 499996 2. 027776 -7.116890E-01 1. 298102E-01

6. 599996 1. 973688 -7 .754114E-01 1. 272496E-01 6.699996 1.921612 -8.344782 E-01 1. 246014E-01 6 .799996 1. 871231 -8 .889959E-01 1. 218944E-01 6.899996 1. 822338 -9. 391606E-01 1. 191508E-01 6.999996 1.774816 -9.852273E-01 1. 163909E-01 7.099996 1. 728600 -1.027484 1.136317E-01 7.199996 1. 683668 -1. 066240 1. 108836E-01 7.299995 1.640030 -1.101803 1. 081580E-01 7.39999S 1.597709 -1. 134467 1. 054642E-01 7.499995 1. 556729 -1.164514 1.028053E-01 7.599995 1.517125 -1. 192201 1.001875E-01 7.699995 1.478921 -1. 217760 9. 761380E-02 7.799995 1. 442143 -1.241400 9. 508589E-02 7.899995 1. 406808 -1.263309. 9.260777E-02 7.999995 1.372934 -1.283645 9.017731E-02 8.099995 1.340529 -1. 302553 8. 780032E-02 8.199995 1.309600 -1. 320155 8.,5478 34E-02 B.;299995 1. 280158 -1. 336552 8. 321S28E-02

8. 399996 1.252196 -1.351835 8.10142 2E-02 8.499996 1. 225722 -1. 366078 7.887962E-02

8. 599997 1.200733 -1.379346 7. 681540E-02 8.699997 1. 177235 -1. 391691 7.482932E-02 8.799997 1. 196599 -1. 388446 8.3932 50E-02 8.899998 1. 192712 -1. 381219 8. 550014E-02

1. 187602 -1.375722 8.5552576-02 9.099998 1.184407 -1. 372126 8. 543236E-02 9.199999 1.182682 * -1.370032 8.521260E-02 9.299999 1.182099 -1.369107 8. 491838E-02 9 .400000 1.182408 -1.369089 8 .456647E-02

9. 500000 1. 183406 -1.369776 8.417004 E-02 9.600000 1.184939, -1.371010 8. 373737E-02 9.700001 1.186888 -1.372666 8. 327780E-02

9. 800001 1.189156 -1. 374646 8.2794 54E-02

9. 900002 1.191668 * -1.376875 8.229468E-02 10.000000 1. 194368 -1.379294 8. 177859E-02 10.100000 1.197209 -1.381856' 8. 125336E-02 10.200000 1.200158 -1.384526 .8.071849E-02 10.300000 1. 203185 -1.387272 8.017738E-02 10.400000 1.206270 -1.390073 7.962 916E-02 10.500000 1.209391 -1.392913 7. 907599E-02 10.600000 1.212539 -1.395778 7.9852047E-02 10.700000 1.215703 -1.398654 7.796098E6-02 10.800000 1. 218873 -1. 401535 7.7399 30E-02 10.900010 1.222044 -1.404415 7. 683378E-02 11.000010 1.225211 -1. 407290 7. 626760E-02

11. 100010 1.228369 -1.410153 7. 570115E-02 11.200010 1.231516 -1.413003 7. 513127E-02 11.300010 1. 234649 -1. 415838 7. 456143E-02

11. 400010 1.237769 7. 3989856-02 11.500010 1. 240870 -1.421451 7. 341693E-02

11. 600010 1.243954 -1. 424227 7.28459 56-02 11.700010 1.247016 -1.426982 7.22708 16-02 11.800010 1.250061 -1.429715 7.169686E-02 11.900010 1.253086 -1.432427 7.112198E-02 Page 3

lof 12.000010 1.256091 -1.435115 7.054621E-02 12.100010 1.259075 -1.437781 6.996840E-02 12.200010 1.262040 -1.440423 6.939100E-02 12.300010 1.264982 -1.443042 6.881388E-02 12.400010 1.267903 -1.445638 6.823530E-02 12.500010 1.270804 -1.448209 6.765780E-02 12.600010 1.273682 -1.450758 6.707720E-02 12.700010 1.276541 -1.453281 6.649942E-02 12.800010 1.279374 -1.455783 6.591803E-02 12.900010 1.282192 -1.458259 6.533635E-02 13.000010 1.284986 -1.460713 6.475499E-02 13.100010 1.287759 -1.463143 6.417334E-02 13.200010 1.290509 -1.465551 6.359024E-02 13.300010 1.293241 -1.467937 6.300624E-02 13.400010 1.295954 -1.470298 6.242431E-02 13.500020 1.298642 -1.472635 6.184042E-02 13.600020 1.301310 -1.474948 6.125543E-02 13.700020 1.303957 -1.477239 6.066981E-02 13.800020 1.306584 -1.479507 6.008346E-02 13.900020 1.309191 -1.481753 5.949731E-02 14.000020 1.311777 -1.483976 5.891071E-02 14.100020 1.314344 -1.486175 5.832172t-02 14.200020 1.316890 -1.488352 5.773330E-02 14.300020 1.319415 -1.490506 5.714568E-02 14.400020 1.321918 -1.492637 5.655696E-02 14.500020 1.324403 -1.494745 5.596639E-02 14.600020 1.326867 -1.496831 5.537626E-02 14.700020 1.329311 -1.498896 5.478402E-:02 14.800020 1.331736 -1.500939 5.419026E-02 14.900020 1.334143 -1.502960 5.359902E-02 15.000020 1.336528 -1.504958 5.300551E-02 15.100020 1.338895 -1.506934 5.241111E-02 15.200020 1.341240 -1.508889 5.181507E-02 15.300020 1.343569 -1.510821 5.121904E-02 15.400020 1.345876 -1.512732 5.062192E-02 15.500020 1.348167 -1.514621 5.002357E-02 15.600020 1.350437 -1.516489 4.942461E-02 15.700020 1.352689 -1.518334 4.882572E-02 15.800020 1.354920 -1.520159 4.822573E-02 15.900020 1.357134 -1.521961 4.762429E-02 16.000020 1.359328 -1.523743 4.702012E-02

.16.100030 1.361506 -1.525504 4.641682E-02 16.200030 1.363663 -1.527243 4.581314E-02 16.300030 1.365802 -1.528961 4.520664E-02 16.400030 1.367923 -1.530659 4.459845E-02 16.500030 1.370026 -1.532336 4-398996E-02 16.600030 1.372112 -1.533991 4.338052E-02 16.700030 .1.374181 -1.535626 4.276973E-02 16.800030 1.376230 -1.537239 4.215769E-02 16.900030 1.378261 -1.538833 4.154232E-02 17.000030 1.380276 -1.540407 4 .0925S7E-02 17.100030 1.382275 -1.541960 4.030832E-02 17.200030 1.384255 -1.543492 3.969054E-02 17.300030 1.386217 -1.545004 3.906859E-02 17.400030 1.388165 -1.546495 3.844668E-02 17.500030 1.390093 -1.547966 . 3.782198E-02 17.600030 1.392005 -1.549417 3.719531E-02 17.700030 1.393901 -1.550848 3.656527E-02 17.800030 1.395782 -1.552258 3.S93502E-02 17.900030 1.397644 -1.553649 3.530069E-02 18.000030 1.399493 -1.555020 3.466428E-02 18.100030 1.401323 -1.556371 3.402504E-02 18.200030 1.403140 -1.557702 3.338385E-02 18.300030 1.404940 -1.559013 3.273974E-02 18.400030 1.406725 -1.560304 3.209186E-02 18.500030 1.408494 -1.561575 3.144047E-02 18.600030 1.410249 -1.562826 3.078729E-02 18.700040 1.411987 -1.564057 3.012876E-02 18.800040 1.413713 -1.565270 2.946629E-02 18.900040 1.415425 -1.566462 2.880011E-02 19.000040 1.417121 -1.567635 2.812902E-02 19.100040 1.418803. -1.568788 2.745483E-02 19.200040 1.420473 -1.569922 2.677305E-02 19.300040 1.422130 -1.571036 2.608772E-02 19.400040 1.423774 -1.572130 2.539635E-02

19. 500040 1.425404 -1.573204 2-469829E-02 19.600040 1.427022 -1.574259 2.399493E-02 19.700040 1.428630 -1.575293 2.328474E-02 19.800040 1.430224 -1.576308 2.256517E-02 Page 4

1of 19.900040 1.431809 -1.577302 2.183984E-02 20.000040 1.433381 -1.578276 2.110512E-02 20:100040 1.434944 -1.579229 2.036150E-02 20.200040 1.436497 -1.580162 1.960638E-02 20.300040 1.438040 -1.581075 1.884016E-02 20.400040 1.439577 -1.581966 1.806357E-02 20.500040 1.441105 -1.582836 1.727110E-02 20.600040 1.442629 -1.583684 1.646614E-02 20.700040 1.444145 -1.584510 1.564514E-02 20.800040 1.445657 -1.585314 1.480731E-02 20.900040 1.447165 -1.586094 1.395013E-02 21.000040 1.448669 -1.586850 1.307002E-02 21.100040 1.450178 -1.587582 1.216443E-02 21.200040 1:451686 -1.588288 1.123034E-02 21 300050 1.453200 -1.588966 1.026593E-02 21.400050 1.454720 -1.589617 9.263691E-03 21.500050 1.456253 -1.590237 8.220721E-03 21.600050 1.457798 -1.590823 7.126574E-03 21.700050 1.459365 -1.591373 5.975167E-03 21.800050 1,460959 -1.591881 4.752456E-03 21.900050 1.462589 -1.592344 3.442875E-03 22.000050 1.464268 -1.592751 2.025819E-03 22.100050 1.466011 -1.593090 4.705717E-04 22.200050 1.467837 -1.593352 -1.250510E-03 22.300050 1.469713 -1.593582 -3.058654E-03 22.400050 1.471583 -1.593843 -4.828255E-03 22.500050 1.473397 -1.594168 -6.474151E-03 22.600050 1.475123 -1.594566 -7.953397E-03 22.700050 1.476750 -1.595028 -9.257955E-03 22.800050 1.478272 -1.595539 -1.040088E-02 22.900050 1.479,694 -1.596079 -1.140486E-02 23.000050 1.481027 -1.596635 -1.229586E-02 23.100050 1.482276 -1.597195 -1.309329E-02 23.200050 1.483453 -1.597750 -1.381631E-02

23. 300050 1.484563 -1.598297 -1.447770E-02

• 23.400050 1.485618 -1.598832 -1.508596E-02 23.500050 1.486617 -1.599354 -1.565339E-02 23.600050 1.487573 -1.599862 -1.618138E-02

• 23.700050 1.488482 -1.60.0355 -1.667478E-02

• 23.800050 1.489351 -1.600834 -1.714033E-02 23.900050 1.490183 -1.601299 -1.758057E-02

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43. 199910 1.455789 -1.590800 9.284188E-03 43.299910 1.455388 -1.590647 9.567777E-03 43.399910 1.454996 -1.590496 9.841777E-03 43.499910 1.454613 -1.590344 1.010676E-02 Page 7

1of 43.S99910 1.454237 -1.590193 1.036406E-02 43.699910 1.453869 -1.590042 1.061368E-02 43.799900 1.453506 -1.589892 1.085660E-02 43.899900 1.453152 -1.589742 1.109205E-02 43.999900 1.452803 -1.589593 1.132236E-02 44.099900 1.452459 -1.589444 1.154736E-02 44.199900 1.452121 -1.589296 1.176662E-02 44.299900 1.451787 -1.S589149 1.198145E-02 44.399890 1.451459 -1.589001 1.219188E-02 44.499890 1.451133 -1.588854 1.239764E-02 44.599890 .1.450812 -1.588707 1.259945E-02 44.699890 1.450496 -1.588561 1.279690E-02 44.799890 1.450184 -1.588416 1.299107E-02 44.899890 1:449875 -1.588271 1.318214E-02 44.999890 1.449569 -1.588126 1.336916E-02 45.099880 1.449268 -1.587982 1.355402E-02 45.199880 1.448968 -1.587838 1.373451E-02 45.299880 1.448672 -1.587695 *1.391179E-02 45.399880 1.448380 -1.587551 1.408693E-02 45.499880 1.448089 -1.587409 1.426012E-02 45.599880. 1.447802 -1.587267 1.442968E-02 45.699870 1.447518 -1.587125 1.459639E-02 45.799870 1.447237 -1.586985 1.476063E-02 45.899870 1.446959 -1.586844 1.492265E-02, 45.999870 1.446684 -1.586704 1.508252E-02 46.099870 1.446410 -1.586564 1.524117E-02 46.199870 1:446137 -1.586425 1.539599E-02 46.299870 1.445869 -1.586286 1.554886E-02 46.399860 1.445603 -1.586148 1.570006E-02 46.499860 1:445339 -1.586010 1.584881E-02 46.599860 1.445078 -1.585873 1.599582E-02 46.699860 1.444818 -1.585736 1.614152E-02 46.799860 1:444560 -1.585599 1 628525E-02 46.899860 1.444304 -1.585464 1:642615E-02 46.999860 1.444052 -1.585329 1.656556E-02 47.099850 1.443801 -1.585194 1.670424E-02 47.199850 1.443553 -1.585060 1.684051E-02 47.299850 1.443304 -1.584926 1.697551E-02 47.399850 1.443060 -1.584793 1.710829E-02 47.499850 1.442817 -1.584661 1.724035E-02 47.599850 1.442576 -1.584529 1.736999E-02 47.699840 1.442336 -1.584397 1.749717E-02 47.799840 1.442100 -1.584267 1.762479E-02 47.899840 1.441864 -1.584136 1.175018E-02 47.999840 1.44163.0 -1.584006 1.787424E-02 48.099840 1.441399 -1.583877 1.799701E-02 48.199840 1.441168 -1.583748 1.811897E-02 48.299840 1:440938 . -1.583619 1.823897E-02 48.399830 1.440711 -1.583491 1.835712E-02 48.499830 1.440487 -1.583364 1.847461E-02 48.599830 1.440263 -1.583237 1.$59141E-02 48.699830 1.440040 -1.583110 1.870685E-02 48.799830 1:439819 -1.582984 1.882121E-02 48.899830 1.439599 -1.582859 1 493365E-02 48.999820 1.439381 -1.582735 1:904452E-02 49.099820 1:439167 -1.582610 1.915560E-02 49.199820 1.438951 -1.582486 1.926526E-02 49.299820 1.438739 -1.582.363 1.937333E-02 49.399820 1:438526 -1. 582240 1.948095E-02 49.499820 1.438316 -1.582119 1.958690E-02 49.599820 1.438107 -1.581997 1 969198E-02 49.699810 1.437900 -1.581875 1:019690E-02 49.799810 1.437693 -1.581755 1.989900E-02 Page.8

Estimate of Radiation Release for MIT Research Reactor During Design Basis Accident by Qing Li Submitted to the Department of Nuclear Engineering in partial fulfillment of the requirements for the degree of Master of Science in Nuclear Engineering at the MASSACHUSETTS INSTITUTE OF TECHNOLOGY May 1998

©Massachusetts Institute of Technology 1998. All rights reserved.

Author .... ................................................

Department of Nuclear Engineering May 8, 1998 Certified by.................................................

I John A. Bernard Director, MIT Nuclear Reactor Laboratory Thesis Supervisor Certified by.................................................

Jacquelyn C. Yanch Professor, Nuclear Engineering Department Thesis Supervisor Accepted by.................................................

.Lawrence M. Lidsky Chairman, Department Committee on Graduate Students

Estimate of Radiation Release for MIT Research Reactor During Design Basis Accident by Qing Li Submitted to the Department of Nuclear Engineering on May 8, 1998, in partial fulfillment of the requirements for the degree of Master of Science in Nuclear Engineering Abstract During a postulated design basis accident at the MIT Research Reactor (MITR),

radioactive fission products may be released from melted fuel plates into the con-tainment. To comply with regulations, the whole-body dose and thyroid dose at the boundary of the exclusion area as a result of this accident are determined.

The fractions of the fission products contained in the fuel that are released through the reactor coolant system (RCS) into the containment are determined based on current regulations, experimental tests, and results from TMI-2 accident.

After the fission products are released into the containment, a portion may be released to the outside through a containment crack or the stack. Also, the por-tion retained in the containment would contribute to the external gamma dose. The calculated dose due to atmospheric release depends on the source strength, the me-teorological conditions, and the dispersion model. For containment crack release and stack release, different dispersion models are used according to pertinent regulatory guides. The gamma dose through penetration or scattering depends on the struc-ture of the containment shielding and is determined analytically under appropriate approximations.

Because the MITR is considering upgrading its power level, results at power levels from 5 to 10 MW are determined. At 5, 6, 7, 8, 9, 10, MW, the whole body'doses at the back fence (8 meters away from the MITR) are 0.644, 0.764, 0.885, 1.00, 1.13, 1.25 rem respectively; the thyroid doses at the back fence are 0.112, 0.135, 0.157, 0.179, 0.202, 0.225 rem respectively; the whole body doses at the front fence (21 meters away from th e MITR) are 0.887, 1.06, 1.22, 1.39, 1.56, 1.72 rem respectively; and the thyroid doses at the front fence are 0.112, 0.134, 0.156, 0.179, 0.201, 0.224 rem respectively.

The results show that even under conservative assumptions, the released doses for power levels from 5 MW to 10 MW are well below the regulatory limit - 25 rem for whole body and 300 rem for thyroid.

2

Thesis Supervisor: John A. Bernard

Title:

Director, MIT Nuclear Reactor Laboratory Thesis Supervisor: Jacquelyn C. Yanch

Title:

Professor, Nuclear Engineering Department 3

Acknowledgments First, I would like to thank Dr. John A. Bernard for his overall guidance and assis-tance. I also want to thank Prof. Jacquelyn C. Yanch for her co-supervision of this work and Dr. Lin Wen Hu and Mr. Fred McWilliams for their great assistance.

4

Contents 1 Introduction 13 1.1 Description of MITR-II and Previous Work. .. .. .. .. ....... ..... 13 1.2 Regulatory Limit on Dose Release .. .. .. .. .... ............... 14 2 Development of the Containment Source Term 17 2.1 Fission Product Inventory .. .. .. .I... .... .... .... .... ........ 17 2.1.1 Fission Product Build-up in the Fuel. .. .. ..... .. .. ...... 17 2.1.2 Build-up of Ar 4 1 in the Containment .. .. .. .. ...... ...... 18 2.1.3 Fission Product Inventory in the Melted Fuel. .. ........ 20 2.2 Release. Fraction. .. .. .. .. .... .... .... .... .... .... ........ 20 2.2.1 Overview of Release Fraction .. .. .. .... .... .... ........ 20 2.2.2 Release Magnitude from the Fuel to the RCS .. .. .. .. ...... 22 2.2.3 RCS Retention ................. ......... 24

.2.2.4 Summary of Release Fraction. .. .. .. .... .... .... ...... 28 2.3 Natural Depletion in Containment .. .. .. .. .. .. .... .... ...... 29 3 Atmospheric Release 31 3.1 Introduction .. .. .. .. .... .... .... .... .... .... .... ........ 31 3.2 Release from Pressure Relief System .. .. .. .... .... .... ........ 31 3.2.1 Release Fraction Through the Stack Filter System .. .. .. .... 31 3.2.2 Release Rate .. .. .. .. .. .... .... .... .... ...... ...... 32 3.2.3 Atmospheric Dispersion Model .. .. .. .. .. .... .... ...... 32 3.2.4 Dispersion Coefficient. .. .. .. .. .... .... .... .... ...... 34 5

3.2.5. Meteorological Data .. .. .. .... .... .... .... .... ...... 34 3.2.6 Application of Dispersion Model. .. .. .. .. .... .... ...... 35 3.2.7 Total Activity Released .. .. .. .. .. ...... .... .... ...... 36 3.3 Release from Containment Leakage .. .. .. .... .... .... .... .... 42:

3.3.1 Leakage Rate .... .. .. .. .. .. ....... .. .. .... .... .... 42 3.3.2. Atmospheric Dispersion Model .. .. .. .. .. .... .... ...... 42

.33Application of Diffusion Models. .. .. .. .... .... .... .... 43 3.3.4 Total Activity Release. .. .. .. .... ...... .... .... ...... 44 3.4 Adj ustment of the. Release Term Outside the Containment. .. .. .... 44 3.5 External Gamma Dose from Plume .. .. .. .. ...... .... .... .... 44 3.6 Beta Dose. .. .. .. .... .... .... .... ...... .... .... .... .... 47 3.7 Thyroid Dose. .. .. .. .... .... .... ...... .... .... .... ...... 49 3.8 Summary .. .. .. .... .... ...... .. ...... .... .... .... ...... 50 4 Direct Gamma Dose, Scattered Gamma Dose, and Gamma Dose Through the Truck Lock 56 4.1 General .. .. .. .... .... ......... .. .. .. .... ...... .... .... 56 4.2 Gamma Source Term. .. .. .. .... .... ........ .. .. .... ...... 57-4.3 Direct Gamma Dose.. .. .. .... .... .... .... ...... .... ....... 58, 4.3.1 Steel Shell Penetration Gamma Dose .. .. .. .. .... ........ 59 4.3.2 Shadow Shield Penetration Gamma Dose. .. .. .. .. ........ 62 4.4 Scattered Gamma Dose .. .. .. .. ....... .. .. .. .... .... ...... 64 4.4.1. Air Scattering Gamma Dose .. .. .. .. .... .... .... ...... 66 4.4.2 Steel Shell Scattering Dose. .. .. .. .. .... .... .... ...... 67 4.5 Radiation Penetration Through the Truck Lock ... .. .. .. .. ...... 72 4.5.1. Concrete Scattered Dose. .. .. .. .... .... .... .... ...... 74 4.5.2, Steel Door Scattered Dose .. .. .. .. .... ...... .... ...... 75 4.5.3 Summary of Radiation Through the Truck Lock........ 76 5 Summary 78 6

A Tables 88 B Figures 99 7

List of Figures 3-1 Dependency of x/Q on SigY for stack release under condition of wind-speed = 11.9 KTS, class D stability, h. = 46 m. .. .. .. .. .. ...... 37 3-2 Dependency of x/Q on wind speed for stack release under condition of class D stability, h. = 46 m. The solid line curve is for a wind speed of 11.9 KTS (6.125 m/s), the dash - dot line curve is for a wind speed of 30 KTS (15.44 m/s) and the dash - dash curve is for a wind speed of 3 KTS (1.544 m/s). .. .. .. .. ... . ... . ... . ........ .. .. .... 38 3-3 Dependency of X/Q on stack height for stack release under condition of class D stability with a wind'speed of 11.9 KTS. The solid line curve is for. a stack height of 46 m, the dash - dot line curve is for a stack height of 46 m and the dash - dash curve is for a stack height of 46 m. 39 3-4 Dependency of X/Q'on SigZ for stack release under condition of wind-speed = 11.9 KTS, class D stability, h. = 46 m...........40 3-5 x/Q. Distribution as a function of plume distance for each atmospheric condition from stack release .. .. .. .. .. .. ... . ... . .... ........ 41 3-6 X/Q Distribution as a function of plume distance for each atmospheric condition from containment -leakage using "conservative" calculation.

A, B, C, D, E and F in the figure stand for the atmospheric stability classes .. .. .. .. .. .. ... . ... . ... . ... . ... . ... . .... ........ 45, 3-7 x/Q Distribution as a function of plume distance for each atmospheric condition from containment leakage using "exact" calculation. A, B, C, D, E and F in the figure stand for the atmospheric stabilities. . . . 46 8

3-8 Two' hour stack release showing beta-Dose, gamma dose and thyroid dose" versus distance. Dotted line is thyroid dose, dot-dash line is gamma dose, and solid line is beta dose. C, D and E is the respective atmospheric stability .. .. .. .. .. .... ....... .. .. .. .... ...... 51 3-9 Two hour containment leakage beta-dose, gamma dose and thyroid dose vs. distance using exact calculation .. .. .. .. .. .. ........... 5:2 3-10 Two hour containment leakage beta-dose' gamma dose and thyroid dose vs. distance using conservative calculation. Different lines in each plot are for different power levels .. .. .. .. .. .... .... ....... 3

.3-11 Two hour containment leakage whole-body dose(rem) and thyroid dose(rem) vs. reactor power for "exact" model .. .. .. .. .... .... .... ...... 54 5-1 Exclusion area doses as a function of reactor power. The solid lines are for whole-body doses, and the solid-dash lines are for thyroid doses.

The circle sign is for doses at 21 meters and the plus sign is for doses at 8 meters. Thyroid doses at 8 meters and at 21 meters are not distinguishable in the plot. .. .. .. .. .. .... .... .......... .. ... 81.

B-i Meander factors for correction of Pasquill-Gifford sigma y values by atmospheric stability class. D, E, F, and G are the stability classes. .100 B-2 Direct dose containment volume transformations. .. .. .. .. .. .... 101 9

List of Tables 1.1 Exclusion Area Distance. .. .. .. .... .... .... .... .... ........ 15 2.1 Source Term Contributions .. .. .. .. ....... .. .. .. .... ........ 19 2.2 Release From the Core in the TMI-2 Accident .. .. .. .. ... . ...... 23 2.3 Fuel Release Fractions From Severe Fuel Damage Tests. .. .. .. .... 24 2.4 Release Fraction From Core to RCS .. .. .. .. .. .... .... ........ 25 2.5 Summary of Experiment Retention Fractions (%of Source) .. .. .... 25 2.6 NUGREG-1150 Expert Elicitation Median RCS Retention factors 27' 2.7 RCS, Retention Factors. .. .. .. .. ......... .. .. .. ... . ........ 27' 2.8 Release Fraction From Core to RCS,.. .. .. .. .. .... ............. 28 3.1 Formulas for cry and o,, by Briggs (197 3). ...... .. ........ .. ..... 35.

3.2 Wind-Speed for Each Stability Category (KTS) Averaged Over'All Directions .. .. .. .. ... . ... . .... .... .... .... ...... ........ 36 3.3 Total containment leakage dose (rem) in two hours using "exact" at-mosphere dispersion model .. .. .. .. .. .... .... ...... .... ....

3.4 Total Containment Leakage Dose (rem) in two hours Using "Conser-vative" Atmospheric Dispersion Model .. .. .. .. .... ...... ...... 55 4.1 Average Containment Volume Source Strength .. .. .. .. .. ........ 58 4.2 Steel Dome Penetration Doses (rem) at 8 Meters .. .. .. .. .. ...... 62 4.3 Steel Dome Penetration Doses (rem) at 21 Meters...... .. .. .. .. 6.63 4.4 Shadow Shield Penetration Doses (rem) at 8 Meters .. .. .. .. ..... ...6 4.5 Shadow Shield Penetration Doses (rem) at 21 Meters. .. .. .. ...... 60 10

4.6 Air'Scattering Doses .(rem) From Upper Source at 8 Meters. .. ... 68

.4.7 Air Scattering Doses (rem) From Upper Source at 21 Meters . - 69 4.8 Air Scattering Doses (r em) From Lower Source at 8 Meters. .. ... 69 4.9 Air Scattering Doses (rem) From Lower Source at 21 Meters .... 69 4.10 Air Scattering Doses (rem) From All Sources at 8 Meters 701 4.11 Air Scattering Doses (rem) From All Sources at 21 Meters 71.

4.12 Single Steel, Scattering Doses (rem) 8 Meters vs. Source 73; 4.13 Single Steel Scattering Doses- (rem) 21 Meters vs. Source 73 4.14 Total Steel Scattering Doses (rem). .. .. .... ........ .. .. .... 73 4.15 Direct Dose at the Concrete Wall..... .. .. .. .. .. .. .. .. .... 75, 4.16 Concrete Albedo Dose (rem)...... .. .. .. .. .. .. .. .. .. .... 76 4.17 Steel Door Scattered Dose (rem)...... .. .. .. .. .. .. .. ...... 77' 5.1 Total Dose at 5 MW .. ... 1..... .... .... .... ....... .. .. .... 80 5.2 Total Dose at 6 MW .. .. .. .... .... .... .... .... ......... ... 82 5.3 Total Dose at 7 MW .. .. .. .... ....... .. .. .. .. .... ........ 82 5.4 Total Dose at 8 MW. .. .... .... .... ...... ....... .. .. ...... 82 5.5 Total Dose at 9 MW .. .. .. .... .... .... .... .... .... ........ 83 5.6 Total Dose at 10 MW... .. .. .. .. .. .... .... .... ...... ...... 83 A.1 Total Core Fission Product Inventory.... .. .. .. .. .. .. .. .... 89 A.2 Values of N'/N' 3 for Neutron-Capture Influenced Isotopes at OT 4x 10 13 . . . . . . . .. . . .. .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . 90 A.3 Parameters for Calculating Atmospheric Doses by Isotope .. .. .. .. 91 A.4 Gamma Emission Energies by Isotope.... .. .. .. .. .. .. .. .... 93 A.5 Attenuation and Absorption Coefficients. ........ .. .. .. .. .... 95 A.6 Shield Thicknesses in Mean Free Paths... .. .. .. .. .. .. .. .... 96 A.7 Point Isotopic Source Exposure Build-Up Factors for Iron (Steel) ... 96 A.8 Coefficients of the Taylor Exposure Build-up Factor Formula. .. 97 A.9 Values of the Functions G(1,p,0,b~) and G(1,p,0,b) . ...... ...... 97 A.10 Air Scattering Input Parameters. .. .. .. .. .... ..... .. .. .. .. 98 11

A. 11 Steel Scattering Input -Parameters. .. .. .. .. .... .... .... ...... 98 12

Chapter 1 Introduction 1.1 Description of MITR-II and Previous Work The MIT reactor is a tank-type research reactor that is cooled and moderated by light water and reflected by heavy water. It currently runs at a power of 5MW. It is fueled by This core design maximizes the neutron flux in the D 2 0 reflector region where numerous experimental beam ports are located. The core is-contained within a light-water filled aluminum tank which is in turn contained within the D2C0 reflector tank. The H 2 0 Coolant is directed so as to flow down along the tank walls and then upwards through the fuel elements. Heat from the primary system is transferred by heat exchangers to the secondary system which dissipates it to the atmosphere through the cooling towers.

The reactor is located at the center of a gas tight cylindrical steel building equipped with a controlled pressure relief system. Access to the containment is through either a personnel or'a truck air-lock. There is also a small personnel airlock which lead',

directly into the control room. All building penetrations are either sealed permanently or can be sealed rapidly by manual or automatic operation. The building is designed to withstand a maximum overpressure of 2 psi and normally operates at a slightly negative pressure.

The design basis. accident. is the maximum credible accident which could result in 13

the r-elease of radiation from the' reactor [1] [2]. For MITR-IJ, the design basis accident is postulated as a coolant flow blockage in the fuel element which contains the hottest channel. This'could occur, for example, as the result of some foreign material falling into the reactor during refueling. After the pumps are started, the material would be swept from the bottom of the tank up to the entrance of the fuel elements. Because of the size of the openings in the adapters at the end of each fuel element, no material passing through the adapter would be large enough to block more than five of the coolant channels. So, the maximum number of plates that could be overheated is four. It is conservative to assume that these four plates could melt completely and release their inventory of fissio n products to the coolant water. For a more detailed description of the MITR please refer to the MITR-II reactor systems manual.

The most recent previous work on this topic is a thesis by Mull [3]. In it, he calculated the dose from radiation release through building leakage and the truck lock and the dose due to direct and scattered gamma radiation during a design basis accident of the MITR-II at 5 MW. His values for release fractions were mainly based on WASH-1400 [4] and other information available then.

Although the purpose of this thesis is to calculate the same doses via the same release paths, great revisions are made in the release fractions based on current ex.-

perimental and analytical studies and the results from the TMI-2 accident. Other revisions and additions included are:

" Radiation release through the pressure relief system to the stack.

" Build-up of Ar 4 1 source term in .the building due to, the sealing of the contain--

ment in the accident.

" Radiation release at different reactor powers from 5MW up to 10MW.

1.2 Regulatory*Limit on Dose Release In CFR 100.11, the limit on dose release is stated as:

14

Table 1.1: Exclusion Area Distance Sector Direction Minimum Exclusion Area Distance X(m)

N 20.6 NNE 22.1 NE 18.7 ENE 18.7 E 17.1 ESE 10.3 SE 8.00 SSE 8.00 5 8.00 SSW 9.53 SW 13.0 WSW 24.0 W 24.0 WNW 24.8 NW 21.0 NNW 20.6

" Exclusion area of such size that an individual located at any point on its bound-ary for two hours imme diately following onset of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from Iodine.

• A low population zone of such size that an individual located at any point on.

its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release during the entire period of its passage would.

not receive a total radiation dose to the whole body in excess of 25 rem or a, total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

To comply with the above regulation, we first have to define the exclusion area for the MITR. The exclusion area'around the reactor was divided into 16 sectors of 45S degrees each, centered on each wind direction. The shortest distance between the!

reactor containment shell and the exclusion area boundary within each sector has been designated as the sector distance, X. These values are listed in Table 1.1.

15

The back fence is defined at'a distance of 8 meters and the front fence is defined at 21 meters ( the Albany St. fence).

16

Chapter 2 Development of the Containment Source Term, 2.1 Fission Product Inventory 2.1.1 Fission Product Build-up in the Fuel, Because current regulations on source term estimation require a simultaneous release assumption, the fission product inventory in the fuel at the time of the accident is assumed to be equal to the maximum value of equilibrium fission products during the two hour release period. This is a conservative assumption.

Based on the volatility, quantity produced, half-life and degree of biological effec-tiveness, the fission produc -t isotopes are selected from a suggested list in Thompson and Beckerley and from those used in the Reactor Safety Study, WASH-1400[4]. The resulting fission product isotopes are listed in Appendix A.1[3]-

The saturation activities of the fission product isotopes can be calculated by both an analytical and a computational method. For the analytical method, the saturation activity, Qi in Curies, due to the presence of N', is QI 3.7 x 1010 17

where N' is the saturation number of nuclei of isotope i, Q1 is the saturation activity due to the presence of N' (Ci) and Ai is the decay constant for isotope i (s-1).

One megawatt equals 3.2 xý 1016 fissions/s if'one assumes that 195 MeV of energy per fission is recoverable, so,

_Y 1P(3.2 x 1016)X 5p 5 Y Q = 8.65 x1 (2.2) where P is the reactor power (MW), and Yj is the fission product yield for isotope i (atoms/fission).

For the computational method, the saturation activity, Qi in Curies, due to the presence of N~i, is i 1.49 x 10 2 7 AiP(NU/N2 35 ) (2.3)1 where MN'/N 5 is the saturated number of fission product atoms produced per initial atom of U 2 35 , and OT is the thermal neutron flux (neutrons/cm 2 - s).- The N,'/N2'35 values found from .literature[5] at -x 4r 1013 are listed in Appendix A.2.

For both methods, the saturation activity is proportional to the reactor power.

For more details of the derivation of the equations, please see reference[3].

The resulting saturation' activity of each isotope at 5, 6, 7, 8, 9 and 10 MW are listed in Appendix A.1.

2.1.2 Build-up of Ar4 1 in the Containment Argon-41 is produced by irradiating air, nearly one percent(0.93%) of which consists of Ar4 0 , with thermal neutrons. Ar40 has a neutron cross section of 0.65 barns and.

can produce Ar 4 1 , which is a gamma and beta emitter, thro .ugh a neutron capture!

rIeaction. The half-life of the Ar 4 ' is 1.83 hour9.606481e-4 days <br />0.0231 hours <br />1.372354e-4 weeks <br />3.15815e-5 months <br />. Because the MIT reactor is designed.

for research, air inher ently gets into the areas of significant neutron flux (in and around the core, the flux is in the order of 1013 1014' neutrons/cm 2 . s) [6]. Air could be expected to get into experimental ports, instrument ports, irradiation facilities, the lead shutter region, and the fission converter area.

18

Table .2.1: Source Term Contributions Source Air*Flow rate Sample Ar 4 ' Conc. Source Term (ft 3 /min) (ACi/ml) (pCi Ar 4 ' /min)

Pipe tunnel 11.8+/-1.8 2.28+/-0.01(x 10-2) 7.63+/-1.2(x 10j)

Core purge 5.75+/-0.30 6.80+/-0.04( x 10-3) 1.11+/-0.06( x 103)

Pneumatic tubes 8*1.9+/-4.1 1.85+/-0.03( x 10-4) 0.430+/-0.02( X103)

Basement hot cell .739+/-37 3.57+/-0.12(x 10- 5 ) 0.747+/-0.05( X10 3 )

Reactor floor hot cell 450+/-23 1.19+/-0.30(x 10-6) 0.015+/-0.004(x 103)

Primary chemistry 834+/-42 1.37+/-0.04( x 10-6) 0.032+/-0.002( x 103)

Medical room 587+/-29 2.12+/-0.05(x 10-6) 0.035+/-0.002(x 103)

Main -Ventilation,

..... 2708+/-135 1.04+/-0.03( x 10-6) 0.080+/-0.005( X103)

Total Input 5417+/-151 2.98 +/-0.01 (X 10-2) 10.08+/-1.2(x 103)

During typical operating co nditions, the ventilation system exhausts air through.

the stack to prevent the build up of an Ar 4 1 source term. In reference [6], the output rate of Ar'"1 was measured thoroughly in all the possible source-term producing areas.

The results are: presented in, Table 2.1. The total release rate from all sources for MITR II at 5MW at normal operating conditions is S = 10.08E3 pCi/mmn [6]. The.

average containment concentration was 2.18E-8 tsCi/ml (measured in 1984).

During an accident, the containment is sealed and the ventilation system is se-cured. The Ar 4 1 already generated in those source. places may be released to the containment and result in a build-up of Ar4 1 in the containment.

The total volume of those source places is 5% of the containment. volume. The

.concentration of Ar4 1 in the containment after sealing would be 2.98 x 10-2 X -5%

1.49 X10-3 pCi/ml for power level at 5MW. Compared with the Ar4 1 containment; concentration of 2.18 Xjo-8 pCi/ml at operation condition, it is much higher.

Measurements showed that the source producing rate is proportional to the power.,

Thus, the Ar 4 1 concentration for power levels other than 5MW can. be determinedi based on the data at 5MW. The .Ar4 ' concentrations for power levels of 6MW, 7MW.,

8MW, 9MW and 10 MW 'would be 1.79, 2.09, 2.38, 2.68 and 1.98 X 10-3 pCi/ml respectively. Compared to other fission products released from the fuel, this concen-tration is much lower- (by an order of 5 to 7), thus the dose contribution of Ar 4 ' at; 19

the exclusion area is negligible.

2.1.3 Fission. Product Inventory in the Melted Fuel In the previous sections the saturated core inventory of fission product activities was determined. But not all of this fission product inventory can be released. Only a small portion of that contained in the four fuel plates that are assumed to melt cou.ld be released.. If the core contains and the ,

then the fraction of the total saturated core inventory which is contained in the four fuel plates, F~, could be dete rmined to be:

F, =0.0176 (2.4)

Therefore, a maximum of 1.76% of each Q' is available for release from the melted core.

2.2 Release Fraction 2.2.1 Overview. of Release Fraction The Reactor Safety Study (WASH-1400), was the first systematic attempt to provide realistic estimates of public risk from potential accidents in commercial nuclear power plants. Based on WASH-.1400, the total release fraction from the fuel to the contain-.

ment is the fraction from the fuel to the primary coolant system(RCS) Ff times the fraction from the RCS to the containment FP and the release is instantaneous. Based on the information available then, the release, fractions were chosen as shown below-in Mull's thesis:

0 Fraction of release from fuel to primary coolant system Ff, 100% of the noble gases(Kr, Xe) 100% of the halogens(I,' Br) 70% of the Tellurium 20

30% of the alkali metals, (C, Rb) 1% of the remaining fissioni products

• Fraction of the release from primary coolant system to containment Fp, 100% of the noble gases 10% of all other isotopes.

Following the publication of WASH-1400 and the accident at Three Mile Island Unit 2 (TMI-2), work was initiated to review the predictive method for calculating fission product release and transport. The results of this review are contained in NUREG.-

0772 [7]. That review resulted in several conclusions that represented significant departures from the WASH-1400 assumptions including the suggestion that cesium iodide (GsI) will be the predominant iodine chemical form under most postulated light water reactor (LWR) accident conditions.

Updated fission product source term methods were developed under the sponsor.-

ship of NRC and the nuc lear industry. As a result, the Source Term Code Package (STCP) was developed as an integration tool for source term evaluation. NUREG.-

.1150 [8] documents a Probabilistic Risk Assessment (PRA) study of five U.S. com-~

mercial nuclear power plants by using the STCP. A limited number of source termn calculations were done for selected plant accident scenarios. The second draft of the study was published in April, 1989 and presents an update, extension,.and improve--

men~t upon the 1975 risk study, WASH-1400[4]. Thus, NUREG-1150 reflects current.

NRC thinking regarding' the source term. But the results are not directly applica.-

ble to MITR, because the. results are very sensitive to the specification of the plant's, design and accident scenarios. Another important document was prepared by the De-.

partment of Energy (DOE) which sponsored the Advanced Reactor Severe Accident-Program in support of the Utility/Electric Power Research Institute(EPRI) advancedi light water reactor (ALWR) program [9]. In this document, a physically-based source term backed by experimental and analytical results is provided and is in agreement-with that from NUREG-1150 under similar conditions.

21

Based on the above, documents, the results from the TMI-2 accident, and available experimental results, the release fractions have been re-determined here for the MITR.

In both references [8] and [9], the release progression is divided into an in-vessel release phase, and an ex-vess~el phase [8, 9]. The "in-vessel release phase" of a severe accident refers to that period of time during which the reactor core is damaged and begins to melt, but is still retained within the R.CS [8]. The "ex-vessel release phase" refers to that period of time after vessel penetration, in which the molten material and most of the remaining radioactive materials would transfer to the containment.

However, for the MITR-II, the core temperature is much lower. Therefore, the core will be retained in the vessel during the whole period. All the releases are due to in-vessel release. Detailed deduction of the release fraction is given in the following sections.

2.2.2 Release Magnitude from the Fuel to the RCS As discussed previously, we assumed four plates in the core melt in the maximum severe case. The radioactive materials contained in these four plates will be released into the primary coolant system and lead to release to the containment. In this section, fission product release from the melted fuel to the RCS is estimated and justifications for these releases are provided.

Noble Gases, Iodine, and Cesium Analysis of fission product releases from the TMI-2 accident [10, 11, 12, 13] and from the severe fuel damage experiments [14, 15, 16, 17, 18, 19, 20, 21] indicate that the releases of iodine, and cesium are approximately equal and are closely related to the fraction of the fuel that becomes molten in the accident sequence. In the TMI-2 accident, about 45% of the core was molten and the releases of iodine, and cesium were in the neighborhood of 55%.

Measurements of residual fission products in previously molten fuel indicate that up to about 10% of the original cesium inventory and somewhat less of the iodine can 22

Table 2.2: Release From the Core in the TMI-2 Accident Isotope Fraction of Core Inventory Released I 0.55 CS 0.55 Te 0.06 Sr 0.001 Ru 0.005 Sb 0.016 Ce 0.0001 be retained by the formation .of chemical species that are stable at high temperatures and/or geomet ries having low surface-to-volume ratios (see References [11] and [22]).

On the basis of these results, releases of 90% of the iodine and cesium from molten fuel are proposed. No residual fission gases were found in molten fuel debris from TMI-2(

see reference [101), so a 100% release of noble gas from molten fuel is proposed.

Tellurium Considerable study has resulted in the understanding that tellurium is released from

,the fuel at about the same rate as noble gases, iodine, and cesium, but is largely retained by the surrounding metallic z'ircaloy cladding and is then released during oxidation of the cladding [23, 24]. Tellurium has a chemical affinity for metallic zircaloy and most other metals such as aluminum. The results of the tellurium release from TMI-2 accident and severe, fuel damage tests are listed in Table 2.2 and 2.3.

Oxidation of the cladding has the effect of increasing the concentration (and there-fore the chemical activity) of tellurium in the remaining metallic cladding, thereby increasing the partial pressure of tellurium. A value of 0.23 for in-vessel tellurium release from the fuel is assumed- for use [9].

Semi-Volatiles and Low Volatiles The release of strontium, bariu m, antimony, and ruthenium have been, found to be quite low as demonstrated -in Tables 2.2 and 2.3 and are bounded by a value of 1%.

23

Table 2.3: Fuel Release Fractions From Severe Fuel Damage Tests Element/Exp.Cond. SFD-ST SFD1-1 SFD1-3 SFD1-4 I 0.51 0.12 0.18 0.26 Cs 0.32 0.09 0.18 0.44-0.56 Te 0.40 0.01 0.01-0.09 0.03 Sr 0.00002 0.00024 0.0088 Ba 0.011 0.006 0.004 0.008 Sb 0.00019 0.0013 Ru 0.0003 0.0002 0.00003 0.00007 Ce 0.000002 0.00009 0.00008 0.00013 Actinides < 0.0001 < 0.00001 Zr Oxidized (%) 75 26 22 32 Fuel Melted () 15 16 18 18 Cerium, lanthanum, and actinides are oxides with very low volatilities which are dissolved in the fuel mat~rix and thus are released to a very small extent (<0.01%)

(see reference [201).

Conclusion The proposed releases from. fuel. to the RCS are listed in Table 2.4. All numbers are fractions of the original core fission product inventory. They are based on experience gained, in the. analysis of core melt progression experiments and the TMI-2 accident.

For the MIT reactor, the release fractions are expected to be lower. The core o MITR is made of a cermet fuel that is more efficient in retaining fission products.

2.2.3 RCS Retention After the fission products are released from the fuel into the RCS, substantial quan-.

tities of fission products may be deposited in the RCS correspondingly reducing the source term to containment.

The NRC and the commercial nuclear industry have developed computer codes to predict the extent of deposition in the RCS for various accident sequences and have.

undertaken experiments to validate their calculational methods. Detailed analysis 24

Table 2.4: Release Fraction From Core to RCS

.Element Releases From Fuel to RCS Noble Gases (NG) 1.0 I 0.9 CS 0.9 Te 0.23 St 0.01 Ba 0.01 Ru 0.01 La 0.0001 Ce. 0.0001 Other 0.0001 Table 2.5: Summary-of Experiment Retention Fractions (%of Source)

Test Species Deposition Close to Fuel Total Piping Deposition LACE LA3A CsOH/MnO=.21 26 77 LA3B CsOH/MnO=.13 15 51 LA3C CsOH/MnO='.61 46 83 LA1 CsOH/MnO=.43 -99 Marviken - 74 SFD 1-4 Iodine- 10 95 Cesium 30 95 LOFT FP-2 Iodine 66 70 Cesium 60 71 using these codes and supporting experimental evidence from tests, indicate that iodine, cesium and the less volatile radionuclides will condense on or interact with other structural materi als released from the damaged core to generate aerosols[9].

Experimental Results on RCS Retention Experimental evidence of aerosol retention processes in the RCS is provided by the LACE[25, 26] and Marv~iken[27] aerosol transport tests as well as by the SFD 1-4 test (see reference [18]) and the LOFT FP-2 test [28]. Table 2.5 summarizes the measured deposition results [91.

25

Additional evidence of fission. product retention during severe accidents is provided by the TMI-2 accident evalluation. Water pathways that existed throughout the duration of the accident retained nearly 100% of the iodine, cesium and other aerosols generated during the accident..

Analytical Results on RCS Retention In support of NUREG-1150, the NRC's TRAP-MELT code (one of the modules of' the Source Term Code Package (STCP)) estimates the amount of RCS retention that can be expected for a variety of accident sequences in modern, operating PWRs and BWRs[29]. The predicted retention factors for aerosols in the RCS range from approximately 15% to 85%ý. The lowest values are associated with large, hot-leg pipe break accidents in PWRs and low-to-intermediate pressure sequences in BW~s in which core uncovery occurs early. Because the design of the MITR-II is different from most of the operating plants evaluated in NUREG-ilSO, the probability of large primary pipe breaks is very low. Hence, the low values of RCS retention associated with large break Loss of Coolant Accident (LOCA) are not applicable to the MITR-Ll.:

The version of TRAP-MELT used in the Source Term Package is recognized. to underpredict aerosol retention within the RCS because of unmodeled pheniomena[9].

An uncertainty analysis (MAAP) was conducted as part of NUREG-1150 in which the range of the RCS retention fraction was determined by polling source term experts.

Table 2.6 shows the resulting median values for RCS retention for different scenarios.

The cases considered by the expert panel are defined as [8]:

" PWR- 1: System setpoint pressure (2500 psia) release through a cycling Over-pressure Relief Valve (PORV).

" PWR-2: High pressure (600 to 2000 psia) release through a very small break or pump seal LOCA..

" PWR-3: Intermediate pressure (200 to 600 psia) release through'a break of approximately two. inches diameter.

26

Table 2.6: NUGREG-1150 Expert Elicitation Median RCS Retention Factors Cases Conditions Iodine Cesium Low Volatility Aerosol PWR 1 Setpoint Pressur'e- 91 96 97 PWR 2/3 High and Intermediate Pressure 59 71 76 PWR 4 Low. Pressure 48 60 66 BWR 1 High Pressure', Early Melt 91 97 .97 BWR 2 Low Pressure, Early Melt 59 70 74 BWR 3 High Pressure, Delayed Melt 72 75 92 Table 2.7: RCS Retention Factors Aerosol Chemical Species Retention factor Iodine 0.7 All Other 0.7

" PWR-4: Low pressure (below 200 psia); release through a large break.

" BWR-1: High pressure fast station blackout.

• BWR-2: Low pressure fast station blackout.

" BWR-3: High pressure ATWS sequences.

The-RCS retention values are higher than the TRAP-MELT predictions and thu~s appear to have corrected the underpredictions.

Conclusions for RCS Retention The RCS retention~for CsI is on the order of 70% for both PWRs and BWR~s according to the STCP and MAAP. calculations [9]. Experimental results from Marviken, SFD, LOFT, and LACE also support such a high retention. So, we assume a retention of 70% for iodine and 70% for all other aerosols. The RCS retentions are summarized in Table 2.7 Because the kinetics a d the mechanism of the interaction of the volatile fission products with the RCS asesn on soldstrutue are complicated and not well

7) 27.

Table 2.8: Release Fractiý rFrom Core to RCS Element, F, Ff FP FF F' NG 0.0.176 1.0 1.0 1.0 0.0176 1, Br 0.0176 0.9 0.30 0.27 0.0047 Cs 0.0176 0.9 0.30 0.27 0.0047 Te 0.0176 0.23 0.30 0.07 0.0012 Sr, Ba 0.0176 0.01 0.30 0.003 0.0001 Ru 0.0176 0.01 0.30 0.003 0.0001 LA 0.0176 0.0001 0.30 0.00003 0.000001 Ce 0.0176 0.0001 0.30 0.00003 0.000001 Other 0.0176 0.0001 0.30 0.00003 0.000001 known, and because of the stochastic, nature of the process, it is difficult to identify an accurate prediction of the. RCS retention rate. Our choice is a 70% retention of' iodine. Thus, a 30% release is conservative compared to the 10% release Mull used.

In the TMI-2 accident, nearly 100% retention of iodine was achieved. The primary-coolant system of MITR is at low temperature and atmospheric pressure. Therefore, leakage from RCS to the containment is also expected to be lower than that in nuclear power reactors. Thus, our assumption provides a big margin. 4Z 2.2.4 Summary of Release Fraction We use the notation of F, to represent the fraction of fission products contained in.

the melted fuel that is available for release, Ff to represent the release fraction from.

fuel to RCS, and FP to represent the release fraction from RCS to the containment.

Hence FP = 1 - RCS retention factor. The total fraction of fission product inventory-in melted fuel released into the containment is FF = Ff x FP. The total fraction of' the fission product in the whole core released into the containment is F' = F' x F,.

These values are summarized in Table 2.8.

28

2.3 Natural Depletion in Containment The chemical form of radionuclide releases to the containment would be: The-noble g~ases are gaseous form; iodine is 97% particulate, 2.85% elemental, and 0.15% organic; the remaining nuclides are particulate [9]. This is based on recent experimental data, including that from the SFD tests, LOFT, and STEP tests, TMI-2 post accident ex-amination, and the ACE tests-as well as an extensive review of the potential chemical.

reactions in the RCS and -containment.

Because the MITR has.. no containment spray or other engineered safety features to reduce the quantity of fission products in the containment atmosphere, depletion of' the radioactive isotopes. released to the containment can occur only through natural processes. These include agglomeration, sedimentation, hygroscopicity and diffusio-phoresis. The noble gases are not expected to undergo any of these depletion process thus have a 100% release fraction.

Agglomeration is the process by which the size distribution of airborne particu-late tends to shift. with time to larger sizes until an equilibrium condition is reached.

This. process affects the. other depletion processes. Sedimentation is deposition due-

,to gravitation. Hygroscopici~ty*is a removal process due to the affinity of the released product for water. As discussed before,,Cs and I will enter the containment in the chemical form of CsOH and CsI, both of which are hygroscopic. In an atmosphere near saturati on, these substances would be absorbed by water by a large ratio. Dif-fusiophoresis occurs when steam condenses on a surface, the aerosol particles will migrate with the water vapor moving to the surface and be deposited.

From simulation 'results[(9], the activity in the containment is varying with time.

It steadily increases as more. fission products are released from the melted core to the containment until it r eaches a maximum. Then it drops because of the natural depletion processes described above and leakage. The drop is fast, about a 99%

drop of the mass of suspended aerosols in 10000 seconds for BWRs, mainly due to the hygroscopocity e~ffects. Because current regulations require a simultaneous release assumption, we-assume the containment activity is at its maximum from the 29

beginning. Under this assumption, we should also include the natural depletion frorri the beginning. For a two hour period, we assume the depletion to be 70% for iodine and cesium and 10% for the others.

The fraction of fission products released to the containment which remain airborne in the containment atmosphere will, be designated as Fc.

• 100% of the noble gases

• 30% of the I, Cs o 90% of others 30

Chapter:3.

Atmospheric Release 3.1 Introduction There are two- ways for the isotopes in the containment to be released to the outside.

One is through a crack in the containment, which is called containment leakage. The other is through the pressure relief system -stack, which is called stack release. Both are discussed below'.

3.2 Release from Pressure Relief System 3.2.1 Release Fraction Through the Stack Filter System During abnormal condition s, the plenum monitors would trip the exhaust dampers thereby sealing the building automatically. The building could also be sealed manually from the control room. In such a situation, changes in atmospheric pressure and temperature may cause the internal building pressure to rise. If the building pressurE should approach its des ign set-point of 2.0 psig, a safe effective relief can be achieved by use of the pressure relief system which can filter the exhaust air and discharge it to the base of the ventilation exhaust stack above the manually operated exhaust control damper. It was shown in the safety analysis of the system that the pressure relief system can be safely operated during a design basis accident.

31

The inside diameter ofthe stack is 0.4318 mn at the exit point and the stack height is 46Gm.

The building pressure relief exhaust line contains two high-efficiency absolute par-,

ticulate air filters that are 99.9% efficient for particle sizes of 0.3 microns, and an activated charcoal filter that is 99% efficient for removal of elemental Iodine. The ac-tual system flow would be determined by the difference between the internal building and atmospheric pressure which is assumed to be 2.0 psig. Experimental data show that flow at 2 psig overpressure is 355 cubic feet per minute (cfm) for filter 1 and 330 cfm for filter 2[301. Thus, the average volumetric flow rate through the stack is 342.5 ft3 -min-'.

The fractions penetrating the filters of the pressure relief system are:

• 100% of noble gases and Br.

• 5% of Iodine.

e 50% of all other isotopes.

The total fraction of the initial, inventory that is released from the stack is:*

Fk,s FS .; - FP- F)lite,. (3.1) 3.2.2 Release Rate The release rate through the stack Als,L is 342.5 ft3 - min, or .0.1616 m3 s_ or 3.42 x 10-5 V - s-1 (V is the volume of the containment. V = 4.73 x 103 in3 )

3.2.3 Atmospheric Dispersion Model Atmospheric dispers ion of a pollutant- is primarily dependent on (1) meteorological conditions such as: ambient- temperature, wind speed, time of day, insulation and cloud cover, (atmoospheric. stability), and (2) pollutant stack emission parameters such as gas velocity and temi- perature. The stability of the atmosphere is determined by the atmospheric. thermal gradient, which is called the lapse rate. Neutral. stability exists 32

for a temperature gradienlt of -1 0 C/100 meters, or a temperature decrease of 1'C for every 100 meters of vert ical ascent. Unstable conditions with lapse rates greater than

-1'C/100 m add to the buoyancy of an emission, and stable conditions (lapse .rates less than -1 'C/100 m) tend to inhibit vertical motion of the pollutant gases (plume).

Dispersion from an elevated source(stack) is effected by the mixing and dilution of polluted gases with the atmosphere.

For a stack release', the maximum ground-level concentration in a sector may occur beyond the exclusion area boundary distance. Therefore, for stack releases, the atmospheric relative.concentration (x/Q) values are calculated at various distances.

The basic equation for atmospheric diffusion from an elevated release is (31]:

1/

1 X/Q

= T zexp[- " ] (3.2) where:

X: ground level concentration (C/in 3 )

Q: pollutant exit rate (Ci/s)

Uh: mean win~d-speed at the release height, in in/s. (In this calculation, the wind-speed at .10-meter level is used.)

he: effective stack height, in in; o~y: lateral plume dispersion coefficient, in m; or.: vertical plume dispersion coefficient, in m; he h,+ hp, ht -c (3.3) hýstack height, in m; hp, rise of the plume above the release point, in m; htmaximum terrain height ( above the stack base )between the release point and the point for which the calculation is made (>=0), in in; c : when vertical exit velocity is less than 1.5 times the horizontal wind-speed, cor-33

rection for down-wash, in- in; c 3(1.5 - Wo/Uh)D (3.4) 6Th F =2.45 WoD( ) (3.6)

W = 14F 5 /1 (3.7)

Wovertical exit velocity of the plume, in m/s; D inside diameter of the stack, in mn; Tstack temperature in K; T, air temperature in K; 3.2.4 Dispersion Coefficient Values of dispersion coefficients, which depend on the downwind distance and the.

atmospheric stability. category, can be determined from the Pasquill curves [32] (a set of diffusion coefficient curves versus plume travel distance). In most references, the dispersion coefficients are given as a set of curves over the range of 102 to 10 meters.

It is impossible to extrapolate accurately to the range of the MITR's exclusion area distance, 8 to 25 meters. One alternative is to use the interpolation formulas for oy and. o,, developed by Briggs which fit the Pasquill curves[4j, see Table 3.1.

3.2.5 Meteorological Data The meteorological data needed for x/Q calculation include wind-speed, wind direc-tion, and a measure of atmospheric stability. The meteorological data used in this thesis were recorded at the Boston Station, MA 240BS 93-95. The wind speed data, are expressed in the unit of knots (KTS) and one KTS equals 1853 meters/hour. The annual average wind-speed for each stability category in the Boston area is listed in.

Table 3.2. We can see that class D is the most frequent stability condition, accounting 34

Table 3A.1:Formulas for cry and a. by Briggs (1973)

Pasquill stability category 0() ()

A O.22x(1 + 0.0001x) 1 0.20x B O.16x(1 + O.OO0lX)- 11 2 0.12x C O-llx(1 + O.OO0lx)" I2 O.08x(1 + 0.0002x)-'/'

D O.08x(1 + O.OO0lx)" I2 O.06x(1 +/- 0.0015x)- 1 / 2 E O.06x(l +/- O.OO0lX)- 11 2 O.03x(1 + 0.0003x)'

F O.14x(1 + O-.OOlX)-1 2 O.016x(1 + 0.0003x)-'

for 73.9423% of the total events.

3.2.6 Application of Dispersion Model We introduced the dispersion model in the previous section. Now, we will discuss the resulting x/Q value and its dependence on the input parameters based on the model.

First, from equation 3-5, we can see that x/Q is proportional to the inverse of ay, And the mean wind-speed. This is also shown in Figures 3-1 and 3-2 for class D stability. The effective* stack height, h., is in the exponential term. Because the flow rate from the stack during an accident is low, we can assume that the effective height equals the stack height (see also Figure 3-3). The buoyant effect of the plume is negligible. The most significant parameter that affects the final X/Q value is 0', which is included in both the exponential term and the magnitude term, and thus affects both the shape of the x/Q distribution and its magnitude. This is shown clearly in Figure 3-4. Because a, depends on the atmospheric stability, the distribution of X/Q also depends on the atmospheric stability. The more unstable an atmospheric condition, the more a pollutant will be deposited in a shorter range with a higher concentration. In contrast, a more stable atmosphere would disperse the pollutant; over a wider range and thus result in a lower concentration. From the meteorological data, we can see that in'the Boston area, the C,D and E categories account for most of the atmosphere cases.

35

Table 3.2: Wind-Speed for, Each Stability Category (KTS) Averaged Over All Direc-tions A B C D E F N 0.0 5.4 7.7 10.3 7.2 4.8 NNE 0.0 6.1 8.2 11.0 6.3 4.5 NE 0.0 5.0 8.4 12.4 6.0 3.8 ENE 5.0 6.3 9.6 11.8 6.5 3.8 E 5.0 6.6 9.8 10.4 6.8 3.8 ESE 5.0 6.2 9.6 10.8 6.9 3.8

.SE 4.5 7.1 8.4 9.4 6.3 4.1 SSE 5.0 5.8. 7.3 9.0 6.3 4.4 5 1.0 5.0 8.5 10.6 6.6 4.8 SSW 4.5 .5.6 9.1 12.1 7.4 5.1 SW 5.*0. 6.6 9.9 12.0 7.9 5.1 WSW 0.0 6.5 9.7 12.0 8.1 5.3 W .5.0 6.7 9.7 13.2 8.4 5.0 WNW 3.0 6.7 9.0 13.4 8.4 5.0 NW 5.0 6.1 10.0 13.2 8.3 5.0

• NNW 4.0 6.5 9.0 .12.5 8.2 4.6 avg. 3.8 6.4 9.2 11.9 7.7 4.6 relative freq.(%) 0.00823 1.8254 8.3007 73.9423 12.0338 3.8154 The. distribution of the X/Q is plotted in figure 3-5 for the six stabilities and the probability of each distribution equals the relative frequency of each stability in Table 3.2.

3.2.7 Total Activity Released Over the two hour release period, the total activity released in Ci from the stack for each isotope is:

Qt' f FRSQS Le(\+\+Ai)tdt (3.8)

=3 Q~A~

F1 ~, 1 -720O(A\S+AG+Ai)(39 S )A + AG + Ai 39 36

x 10 -

5 De'pendency of X/Q on SigY For Stack Release 0

0..........I5 10' 102 10 01 10 PLUME TRAVEL DISTANCE (METERS)

Figure 3-1: Dependency of X/Q on SigY for stack release under condition of wind-.

speed = 11.9 KTS, class D stability, h. = 46 m.

37

X 10-5 Dependency of X/Q on Windspeed For Stack Release 4.5

.4 3.5

`ýE2.5-a) 1.5 I 1wsp= 11k\9 KTS=6.125 rn/s 0.5- wsp= 0KTS 15.44 mr/s, 10 1021 10 10s 106 PLUME TRAVEL DISTANCE (METERS)

Figure. 3-2: Dependency of x/Q on wind speed for stack release under condition of class D stability, h. = 46 mn. The solid line curve is for a wind speed of 11.9 KTLS (6.125 m/s), the dash - dot line curve is for a wind speed of 30 KTS (15.44 m/s) and the dash - dash curve is for a wind speed of 3 KTS (1.544 m/s).

38

X 10- D ependency of X/Q on Stack Height For Stack Release 6 r--'

4-0 2-0-

10 102 103 104 0 10 PLUME TRAVEL DISTANCE (METERS)

Figure 3-3: Dependency of x/Q on stack height for stack release under condition of class D stability with a wind speed of 11.9 KTS. The solid line curve is for a stack height of 46 m, the dash - dot line curve is for a stack height of 46 m and the dash -

dash curve is for a stack height of 46 m.

39

x1 - Dependency of X/Q on Sigz For Stack Release 1

0 0.5 0 1 ... . . .. . 4 1..................I-101 .102 10o 04 105 10 6

.PLUME TRAVEL DISTANCE (METERS)

Figure 3-4: Dependency of x/Q on SigZ for stack release under condition of windspeed

=11.9 KTS, class D stability, he = 46 m.

40

x 105 Stack Release 0

10 11010 10 2 103 PLUME TRAVEL DISTANCE (METER) 1106 i01010 Figure 3-5: x/Q Distribution as a function of plume distance for each atmospheric condition from stack release.

41

3.3 Release from Containment Leakage 3.3.1 Leakage Rate

.The reactor building is designed to withstand internal pressure of 2.0 psig greater than atmospheric. If the building pressure should approach its design setpoint of 2.0 psig, a safe, effective relief can be achieved by use of the pressure relief system which can filter the exhaust air and discharge it to the ventilation exhaust stack.

The maximum permissible leakage rate is 1% of the building volume per day per psi of building overpressure. An integral air leakage test of the reactor building containment is performed annually wi th a maximum time of 18 months between tests to ensure above criteria.

When an accident happens, the containment building is assumed to reach its set-point pressure of 2.0 psig simultaneously. The leakage rate of the building is assumed at its maximum permissible value of 1%. With the above conservative assumptions, the leakage rate, L~ i L 0.02V/day = 2.3 x 10-7 V/s (3.10) where V is the volume of containment (4.73 x 103 in3 ).

3.3.2 Atmospheric Dispersion Model For neutral (D) or stable (E,F, or G) atmospheric stability conditions when the wind-speed at the 10-meter level is less than 6 meters per second, meandering of the hori-zontal plume may be considered. x/Q values may be determined by using following equationis[31]:

x/Q= 1 Ui0 (iraorz + A/2) (.1

_1 42

x/Q =-1 /(3.13)

U7 1o7rEY3Ca where x/Q is relative concentration, in s/rn3 ,

ir is 3.1415926, U10 is wind-speed at l0meters above plant grade, in m/s, oy. is lateral plume spread, in m, ouý is vertical plume spread, in m, EY, is lateral plume spread with meandering and building wake effects, in m. For distances of 800 meters or less, EY = Mo-y where M is determined from Appendix B-i; for distances greater than 800 meters, Ey, =(M - 1)ciysoom + ory, and A is the smallest vertical-plane cross-sectional area of the reactor. building, in inm The larger value from equation 3.11 and equation 3.12 should~ then be compared with the value from. equation 3.13 and the lower value should be selected as X/Q.

During all other meteorological conditions, plume meandering should not be con-sidered. The appropriate x/Q value is the higher value from equation 3.11 and 3.12.

These procedures for calculating x/Q are conservative. The reason that the higher value of equation 3.11 and 3.12 is chosen is' because the NRC specifies that the reduc-tion of x/Q due to the wake effect can be no more than a factor of three. We call the values derived from these procedures "conservative" values and those from equation 3.11 "exact" values. The resulting doses obtained by using these two methods will be compared.

3.3.3 Application of Diffusion Models In Figure 3-6 the resulting x/Q from the "conservative" calculation is shown for each stability class and in Figure, 3-7 the resulting x/Q from the "exact" calculation is shown for each stability class. The difference is obvious. For our case, where the plume distances are small (smaller than 100 meters), the wake effect from the containment building would be strong. Thus it would be justifiable to use the. "exact" equation instead of the "conservative" method.

43

Figure 3-7 shows that only~in class A stability would the x/Q value exceed that in class F stability. Class A has, frequency of occurrence of less, than 1%. Therefore, calculation of the dose for class F stability would give a conservative estimate of the dose with frequency greater than 99%.

3.3.4 Total Activity Release Over the two hour release period, the total activity released in Ci from the stack for each isotope is:

Q~~,I 70 ks~~ (As

- A+Ai)t (3.14)

Q FR'QA~ 3.5 tG SQ' A~ A ¶++A (315 3.4 Adjustment of the Release Term Outside the, Containment Reduction due to decay, ground deposition, and precipitation scavenging of the fission products after leaving the containment can be conservatively neglected.

3.5 External Gamma Dose from Plume It is assumed that the plume is infinitely large in calculating the external doses. This assumption simplifies the computations and gives conservative results. Consider a hemispherical uniform cloud with infinite radius located above ground level, contain-ing a radionuclide with a concentration of X G/m 3 , emitting gamma rays with an average ene.rgy. of E MeV. The exposure rate (R/s) to the center point is[33]:

-y O.262EX, (3.16) 44

"Conservative" 70 60-50-F 40-0 30-20-E 10 A,B3,C,D 002 100 10'1 PLUME DISTANCE (METER)

Figure 3-6: x/Q Distribution as a function of plume distance for each atmospheric:

condition from containment leakage using "conservative" calculation. A, B, C, D, E and F in the figure stand for the atmospheric stability classes.

45

X1-3 "Exact" 2.5 2A 3

1F 0.5-03 100 101 102 10 PLUME DISTANCE (METER)

Figure 3-7: x/Q Distribution as a function of plume distance for each atmospheric condition from containment leakage using "exact" calculation. A, B, C, D, E and F in the figure stand for the atmospheric stabilities.

46

Thus the-total exposure, in roentgen due to isotope i s

,y 0. 262Ei,QY' (X/Q) (3.17)

The E.' are obtained by evaluating the gamma energy spectrum of each isotope.

To obtain the dose equivalent, 'yjx must be multiplied by the f-factor, which converts the roentgen to dose in tissue, a nd by the quality factor, which converts rad to rem.

Both of the factors are approximately unity, so that

=0.262E,'Q'..(X/Q) rem (3.18)

Another inethod was developed by using computer-generated conversion factors

[41:

H iiXQ (319 where C'is photon dose conversion factor for immersion in contaminated air due to isoto~pe i, in rem per Gi-s/m.

For those isotopes whose C!' are not available, equation 3.17 is used to determine the gamma dose. All the parameters used in the calculation of the gamma dose are listed in Appendix A.3.

The estimated total gamma exposure distribution with distance due to all isotopes from containment leakage (from "exact" model) is illustrated in Fig 3-9 and from stack.

release is illustrated in Fig 3-8. In the stack release, only the results for 10 MW are plotted and the stabilities of classes C, D and E, the total frequency of which -are around 94%. For comparison, the "conservative" values from containment releases are illustrated in Fig. 3-10.

3.6 Beta Dose, The dose rate in air from an infinite uniform cloud of beta radiation is determined from[331:

47

pD=87.5 xO.262EBjX = 0.229E,3X rad/s (3.20) 100 where, 3':Beta dose rate (rad/s)

E13: Average beta energy per disintegration (MeV/dis)

X: Concentration of beta-emitting isotope (Ci/m 3 ).

The dose equivalent rate in tissue is then given by

'H=0. 229EgXx ff(d, Emo,,,) rem/s (3.21) where f is an experimentally determined function of d, the distance into the tissue, and E,,,, the maximum energy of the emitted #~rays. The dose rate is largest at the surface of the skin, where f = 1, and decreases rapidly with distance into the tissue.

To be conservative, the external dose due t o the 03 plume is computed with f = 1.

The total beta dose equivalent in rem in two hours is:

H= 0.229Bf X(t)dt (3.22)

The X (in Ci/m3 ) can be related to the previously determined x/Q value by the relationship XWt WO(OQQ~) (3.23) which when integrated yields:

fJX(t)dt =(x/Q) f Q(t)dt =(X/Q)QT (3.24)

The total beta dose equivalent(in rem) received due to isotope i is therefore:

=g 0. 232ýQ'(x/Q) (3.25)

The value of R'3 equals. one-third the value of the maximum beta energy for isotope 48

i and are listed in Appendi x A.3.

The estimated two hour total beta dose equivalent distribution with distance due to all -isotopes from containment leakage (using the "exact" model) is illustrated in Fig 3-9 and from stack release. is illustrated in Fig 3-8. In the stack release, only the results for 10 MW are plotted and the stabilities include classes C, D and E, the total frequency of which are around 94%. The "conservative" values from containment releases are illustrated in Fig. 3-10 for comparison.

3.7 Thyroid Dose The thyroid dose equivalent is calculated according to WASH-1400[4]:

Tiff BCýQ'(X/Q) (3.26) where TfP: Dose to thyroid from isotope i (rads),

Br: Breathing rate (M 3 /S)*

Ci : Thyroid inhalation conversion factor for isotope i (rem per Ci inhaled) in 0-2 days. The values are listed in.Appendix A.3.

Isotopes of interes t which are not included in WASH-1400 were checked against ICRP Report #2 [34) and found to have no contribution to the thyroid dose. The standard breathing rate for the calculation of internal dose is 3.47 x 10-4 M 3 /S [31].

The estimated total thyroid dose distribution with distance due to all isotopes from containment leakage( from the "exact" model) is illustrated in Fig 3-9 and fromn stack release is illustrated in Fig 3-8. In the stack release, the results for 10 MW are plotted for stability classes C, D and E, the total frequency of which is around 94%).

For comparison, the "cons ervative" values from containment release are illustrated in Fig. 3-10 49

Table 3.3: Total containment leakage dose (rem) in two hours using "exact" atmo-sphere dispersion model

-Power(MW) Beta Dose(rem) Gamma Dose(rem) Thyroid Dose(rem)

Dose at 8 m 5.0 0.0054 0.0084 0.1121 6.0 0.0065 0.0101 0.1346 7.0 0.0076 0.0118 0.1570 8.0 .0.0087 0.0135 0.1794 9.0 0.0098 0.0152 0.2018 10.0 0.0109 0.0168 0.2246 Dose at 21 m.

5.0 0.0054 0.0084 0.1116 6.0 0.0065 0.0101 0.1339 7.0 0.0076 0.0117 0.1562 8.0 0.0086 0.0134 0.1786 9.0 0.0097 0.0151 0.2009 10.0 0.0108 0.0168 0.2235 3.8 Summary The beta, gamma, and thyroid doses at the front and back fence for MITRi at power levels of 5MW up to 10 MW are listed in Tables 3.3 and 3.4. The former were obtained using the "exact" -atmospheric dispersion model and the latter by using the.

"conservative" atmospheric dispersion model. Even in the latter, the doses are well within the limitation of 25 rem for whole body dose and 300 rem for thyroid dose.

Because of the short distance of the exclusion area, the wake effect of the atmospheric dispersion should be dominant, and hence the "exact" values are more reasonable for MITR.

50

.STACK RELEASE

.,*~ I-0.0 C

0. 02-

-0.0 15-w D E

ci) 'V 0

o0. 01-0.0 05-I I V.

n

10 0 101 10210 31 10 10 PLUME TRAVEL DISTANCE (METER)

Figure 3-8: Two hour stack release showing beta-Dose, gamma dose and thyroid dose versus distance. Dotted line is thyroid dose, dot-dash line is gamma dose, and solid line is beta dose. C, D and E is the respective atmospheric stability.

51

X1 --3 Beta Gamma GammaThyroid 11i . I 0.018 0.24 0.22I 10 0.016 0.2 9

0.014I _7 .....

.. 0.18 w 8 0.012. 0.16~

U) 0 7 0

0.14k 0.01 6

0.121 0.008 5 0.1 4101. 2.0 0 1 1.2 0.0' 10 102 1010 10 10o 10 10 10 1 DISTANCE (METER) DISTANCE (METER) DISTANCE (METER)

Figure 3-9: Two hour containment leakage beta-dose, gamma dose and thyroid dose:

vs. distance using exact calculation.

52

Beta Dose a Dose Dose BetGamma TyodDs Thyroid Dose 5

4 80 6

ý3 5 60 CC) 02 3 40 2-1 20-0 00 101 102 101 102 101 DISTANCE (METER) DISTANCE (METEI:

10)

DISTANCE (METER)

Figure 3-10: Two hour containment leakage beta-dose, gamma dose and thyroid dose vs. distance using conservative calculation. Different lines in each plot are for different power levels.

53

Whole-body Dose At 8 MeterThriDoeA8Mtr Thyroid Dose At 8 Meter 0.02 0.25 0.015[ o0.2 Gammajj6se CI:

Cl) 0 0.011- 00.15 0

Bet se 0.005' 0.1 L.

I. 6 8' 10 4 6 '8 10 Reactor Power (MW) Reactor Power (MW)

Whole-body Dose At 21 Meter Thyroid Dose At 21 Meter o 0.015 0) cc, Gammna Dse cc w

Cl) 0 0.01 01 0

B ose 0.005I 0.1 1.

I. 6 8 10 4 6 8 10 Reactor Power (MW) Reactor Power (MW)

Figure 3-11: Two hour containment leakage whole-body dose(rem) and thyroid dose(rem) vs. reactor power for "exact'. model.

54

Table 3.4: Total Containment Leakage Dose (rem) in two hours Using "Conservative" Atmospheric Dispersion Model

-Power(MW) Beta Dose(rem) Gamma Dose(rem) Thyroid Dose(rem)

Dose at 8 m

.5.0 2.0559 3.1888 42.4761 6.0 2.4670 3.8265 50.9713 7.*0 2.8782 4.4643 59.4665 8.0 3.2894 5.1020 67.9617-9.0 3.7006 5.7398 76.4570.

10.0 4.1117. 6.3778 85.0591 Dose at 21 m 5.0 0.2997 0.4649 6.1923 6.0 0.3597 0.5578 7.4308 7.0 0.4196 0.6508 8.6693 8.0 0.4795 0.7438 9.9077.

9.0 0.5395 0.8368 11.1462 10.0 0.5994 0.9298 12.4002 55

Chapter 4-..

Direct Gamma Dose, Scattered Gamma Dose, and Gamma Dose Through the*Truck Lock 4.1 Genieral Those isotopes that do not leak from the containment will constitute a source of gamma radiation. The gamma dose at the exclusion boundary from the isotopes, retained within the containment building includes the penetration or direct gamma dose, the scattered gamma dose, and the gamma dose through the truck lock.

The containment building shield consists of two parts. One is the sides which are shi elded by concrete and steel. The other is the dome which is shielded only by steel. This would result in two sets of dose for both the direct gamma dose and the scattered dose.

The methods used in this- chapter are the same as those used by Mull(3]. A brief summary of the methods and the resul ts calculated from the methods are provided.

56

4.2 Gamma Source Term Those isotopes that are deposited in the containment and those that remain airborne in the containment would contribute to the direct and scattered gamma dose.

The initial quantity of fission product i airborne, in the containment is equal to F FRQ~. This will be reduced, over time due to leakage and decay. The quantity which deposits inside the c6ntainment is equal to FFk(l - orQ1-eFi Qi~ This would be reduced over time due to decay only.

The time-dependent containment inventory of fission product i for direct and scattered gamma dose is therefore:

Q~()=F Qi [e-(AL+AX)t + 1-1)i](41 where Qi is in Curies.

The total number of decay emissions from isotope i over the two, hour period is given by:

Q2 (t) (3.7 x 1010) f20Q'(t)dt (4.2) which, after integration, gives 1 e-(XL +AjYT200 1 1 e-A(4.3)

QcT(t)

QiFQ

=(3.7 x 1010 )Q[ AL+A + (= - 1)(43 The energy and abundance of each isotope's gamma decay spectrum are also shown in Table A.5. For convenience, photons have been grouped into discrete energies following a logarithmic scale, with individual photons being allocated to the closest; energy.

The total number of 'emissions of each energy is then equal to the product of the number of emissions, QT' for each isotope and the photon abundance for that; isotope at that energy, summed over all isotopes. The resulting total number of gamma emmisio~ns for each energy is divided by the containment volume and duration 57

Table 4.1: Average Containment Volume. Source Strength Average Volume Source Strength (Photons/cm 3 - s)

E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.3 3.06E+/-02 3.67E+/-02 4.29E+/-02 4. 90E+/-02 5.51E+02 6.1 2E+02 0.04 1.65E-02 1.98E-02 2.31 E-02 2.64E-02 2 .97E-02 3.30OE-02 0.05 1.92E+02 .2'.30E+02 2.68E+02 3.06E+/-02 3.45E+02 3.84E+02 0.06 1 .39E+/-02 1.67E+/-02 1.95E+02 2.23E+/-02 2.51 E+02 2. 79E+02 0.08 1.11E+i04 1.34E+04 1.56E+04 1.78E+04 2.01E+04 2.23E+04 0.10 5.67E+/-00 6.80E+00 7.94E+00 9.07E+/-00 1.02E+/-01 1.14E+01 0.15 6-93E+03 8.32E+03 9.71E+03 1.11 E+/-04 1.25E+04 1.39E+04 0.*20 6.39E+/-03 7.67E+03 8.94E+03 1.02E+/-04 1.15E+04 1.28E+04 0.30 7.47E+03 8.97E+03 1.05E+04 1.20E+/-04 1.34E+04 1.49E+04 0.40 1.65E+04 1.78E+04 1.91E+04 2.04E+/-04 2. 17E+04 2.30E+04 0.50 1.08E*+04 1.30E+/-04 1.52E+04 1.73E+04 1.95E+04 2.17E+04 0.60 2.33E+04 2. 79E+04 3. 26E+04 3. 72E+04 4.1 9E+/-04 4.66E+04 0.80 3.55E+/-04 4.26E+04 4.97E+04 5.68E+04 6.39E+04 7. 10E+/-04 1.00 4.31E+/-03 5..18E+03 6.04E+03 6.90E+/-03 7.77E+/-03 8.63E+03 1.50 6.81E+/-03 8..17E+03 9.53E+03 1.09E+/-04 1.23E+04 1.36E+04 2.00 1.0.4E+04 1.25E+04 1.45E+04 1.66E+/-04 1.87E+/-04 2.08E+04 3.00 1.07E+/-03 .1.28E+/-03 1.50E+03 1.71E+/-03 1.93E+/-03 2. 14E+/-03 4.00 3.35E+t01 4.02E+01 4.69E+01 5.36E+01 6.03E+01 6.69E+01 of release to obtain the time-averaged total containment volumetric: strength, SVT.

Values of SVT for each energy E are listed in Table 4.1.

4.3 Direct Gamma Dose In order to calculate the direct gamma dose at a given point on the ground outside the containment, the containment is divided into two parts (see Appendix B-2). Part one is all locations from which gamma rays will reach the target point through the steel dome. The corresponding 'volume,is designated as V1 . Part two is all locations from which gamma rays will have to penetrate the concrete shielding to reach the!

target point. This part's volume is designated as V2 . The values of V, and V2 are determined for the back fence (8 m) and the front fence (21 m) as[31:

V1 (8) =0.O1V 58

V1 (21) = 0.05V V2 (8) =0.99V V2 (21) 0 .95V 4.3.1 Steel Shell P enetration Gamma Dose For simplicity, we make two approximations here. First, the radioactive isotopes are distributed uniformly in the containment. Second, the volume V, is approximated to be a sphere. Then, the spherical volume source of constant strength Sv (photons/cm 3-s) can be approximated by a disk of the same radius (RI) having a surface source strength SA 4-RjSv (4.4) 3 located at a self-absorption. distance z [351. If it is assumed that the containment atmosphere is primarily air,, then self-absorption will be small and it is conservative to assume z = 0.

The unscattered flux at a point lying behind a parallel.'shielding slab from -this.

disk source is (in photons/cn 2-s) [361:

1 b

1 s~ e-tdt (4.5)

By introducing the E,, functions defined by the integral:

E.(x)= X 1 tn dt (4.6) the flux can be expressed as:

-BSA 1

7tl[E 1 (bi) - EI(bisec~i)](47 ()

2 where

=~ photon flux (pho'tons/cm'-s)

B =buildup factor 59

• SA, = surface source strength for volume V, and energy E (photons/cm 2 -s)

-l ISTTST (number. of mhean free paths in the steel shield)

A.ST = linear attenuation coefficient for steel (cm-')

TST = steel thickness (C~m)

For the derivation of above equations, please see reference [3] [35]. Substituting the SA, the flux becomes 2-BR, Sy,[El (b1 ) - El (b, secO,)] (4.8) 3 Buildup and attenuation, in the air will be neglected. Both effects are small and tend to cancel each other.. Values of AS~ and subsequent values of b, are shown in Appendices A.5 and A.6.

The dose at P is determined using the conversion factor CD Dose =CDqS7 (4.9) where CD (1 rem/rad)(E MeV/potn)1.6 X .10-6 ergs/MeV)(,!acm2 /1g)(7200 s) 100 ergs/g - rad (4.10) which reduces to CD = 1.15 x 10-4 Ea (4.11) where pa, is true energy absorption coefficient in air (cm 2 /g). Substituting CD and 0., into Eq. 4.9, the dose (in rem) becomes 5

Dose =7.67 x 1- EPaBRISv1 [El (b,) - El(b, secO,)] (4.12)

For computational purposes it. is convenient to express the buildup factor as a mathe-matical function. One of the most useful forms is the sum of exponentials[37], namely:

60

B.- Ae'"pT + (1 - A)e6 Q2 pT (.3 (4.13) in which A, a,, and a 2 are functions of energy. Values of A, a,, and a 2 are listed in Appendix A.8. Substituting-the expression for B into the equation for E1 (b) and integrating, the result is' El (b1) AE1 (b',) + (1 - A)E1 (b") (4.14) where For 3 << 1 and b > 0, below relation would hold 1371:

El (b) - El [b(1 + 6)] 6e5 - b (4.15)

Let (1 + 3) =sec9, the final result would be[31:

D = 7.67 x lO-5 EiiaRiSyi[A(secOi - 1)e-b + (1 - A)(sec91 - I)e-b] (4.16)

Assuming the fission products are uniformly distributed in the containment, the volume relations lead to the source strength relations:

Sv1 (8)= 0.01 SVT Sv 1 (21) = 0.05 SvT The scattering geometry parameters are:

8 mn: 01= 0.179 radians; R, = 2.25 X 10 2 CM, 21 m : 01 = 0.169 radians; R, = 3.90 x101 cm.

The resulting doses are listed in Tables 4.2 and 4.3. For E < 0.5 MeV, where the Taylor coefficients are not available, appropriate tabulated point buildup factor data are used (Appendix A.6). The dose can be determined approximately by:

Dose =7.67 x 1O'EpaRiSviB(sec9j - 1)e-bl (4.17) 61

Table 4.2: Steel Dome Penetration Doses (rem) at 8 Meters

-E (MeV) 5 MW -6MW 7MW 8MW 9MW 10MW 0.10 8.41E-09 1.01E-08 1.18E-08 1-35E-08 1.51E-08 .1.69E-08 0.15 3.81E-05 4.58E-05 5.34E-05 6.1OE-05 6.87E-05 7.64E-05 0.20 6.92E-05 8.30E-05 9.69E-05 1.1E-04 1.25E-04 1.38E-04 0.30 1.32E-04 1.59E-04 1.85E-04 2.12E-04 2.38E-04 2.64E-04 0.40 4.19E-04 4.51E-04 4.84E-04 5.17E-04 5.50E-04 5.83E-04 0.50 4.02E-04 4.82E-04 5.63E-04 6-43E-04 7.23E-04 8.04E-04 0.60 1.04E-03 1.25E-03 1.46E-03 1.67E-03 1.88E-03 2.09E-03 0.80 2.10E-03 2.52E-03 2.94E-03 3.36E-03 3.78E-03 4.19E-03 1.00 3.09E-04 3.71E-04 4.33E-04 4.95E-04 5.56E-04 6.18E-04 1.50 .6.69E-04 8.02E-04 9.36E-04 1.07E-03 1.20E-03 1.34E-03 2.00 1.24E-03 1.49E-03 1.74E-03 1-99E-03 2.24E-03 2.48E-03 3.00 1-68E-04 2.02E-04 2.35E-04 2-69E-04 3.03E-04 3.36E-04 4.00 6.40E-06 7.69E-06 8.97E-06 1.02E-05 1.15E-05 1.28E-05 Total 6.60E-03 7.87E-03 9.13E-03 1.04E-02 1.17E-02 1.29E-02 Doses for E < 0.10 have not been determined because buildup factor data for steel.

in this energy range is not available and the increasing attenuation at lower energies makes the dose at these energies negligible.

4.3.2 Shadow Shield Penetration Gamma Dose The dose due to isotopes in V 2 can can be obtained by approximating the source as a right circular cylinder volume source with a radius of R 2 and a height of h2 shielded by a slab shield of thickness of b2 .

For this situation the flux at point P is given by [37]:

BR2SV2*G (k,,g R2,b 2 ) (4.18) 27r where:

pT2'- (must be >--; 1.25.)

As linear attenuation coefficient in the source medium (cm-1) b2 = tic + ISTTST = total shadow shield thickness in mean free paths 62

Table 4.3: Steel Dome Penetration Doses (rem) at 21 Meters E (MeV) 5 MW. 6MW 7MW 8MW 9MW 10MW 0.10 6.49E-08 7.78E-08 9.08E-08 1.04E-07 1. 17E-07 1.30E-07 0.15 2.94E-04 .3.53E-04 4.12E-04 4.71E-04, 5.30E-04 5.89E-04 0.20 5.34E-04 6.40E-04 7.47E-04 8.54E-04 9.61E-04 1.07E-03 0.30 1.02E-03 1.22E-03 1.43E-03 1.63F,03 1.84E-03 2.04E-03 0.40 3.23E-03' 3.48E-03 3.74E-03 3.99E-03 4.24E-03 4.50E-03 0.50 3.10E-03 8.72E-03 4.34E-03 4.96E-03 5-58E-03 6.20E-03 0.60 8.05E-03. 9.67E-03 1.13F,02 1.29E-02 1.45E-02 1.61E-02 0.80 1.62E-02 1.94E,02 2.27E-02 2.59E-02 2.91E-02 3.24E-02 1.00 2.38E-03 ý2..86E03 3.34E-03 3.81E-03 4.29E-03 4.77E-03ý 1.50 5.16E-03 6.19E,03 7.22E-03 8.25E-03 9.28E-03 1.03E-02 2.00 9.58E-03 1.15E-02 1.34E-02 1.53E-02 1.72E-02 1.92E-02 3.00 1.30E-03. 1.56E-03 1.82E-03 2.08E-03 2.33E-03 2.59E.-03 4.00 4-94E-05 5.93E-05 6.92E-05 7.90E-05 8.89E-.05 9.87E-05 Total 5.09E.-02 6.07E-02 7.05E-02 8.02E-02 9.OOE-02 9.98E-02 G = attenuation function Then the dose is given. by multiplying the flux by a conversion factor CD:'

Dose 1.15 < 10 4 1EpaBR2 SV2 G~,p s2 2 4..9 21rGkp iRb)(.9 There is no tabulated buildup factor for a laminated shield. To find the. exact buildup factor for a laminated shield, complicated numerical methods have to be used to solve the Boltzmann transport equation, with appropriate boundary conditions.

However, it is found that the buildup factor is largely determined by the total number of mean free paths and is characteristic of the material in the outmost region if that is at least two or three mean free path in thickness. If the outmost single region is not thick, the buildup factor for the materials constituting the outmost two or three mean free paths can be chosen. From Appendix A.6, one can see that below an energy of 0.1MeV the buildup factor of steel should be used and above an energy of 0.1MeV the buildup factor of concrete should be used.

Incorporating the buildup factor in the Taylor form into the G function, the dose 63

becomes [3]:

Doe 1.15 x 10- 4 EjiR2 Sv [AG(k, p, pR 2, b) +(1 -A)G(k, p, p 8 R 2 , b")] (4.20) where

= 1+c 1 )b2 .

2b (I + a 2 )b2 Values of the Taylor coefficients for concrete are listed in Ap-pendix A.8. The respective volumes for target points at 8 meters and 21 meters are V2 (8) =4.68 x i M.

3 V2(21) = 4.49 x 10 3 M.

For convenience, k is set equal to one. This eliminates one set of interpolations in the G function tables And is not too far from the actual containment h/R ratio.

Given that k = 1, and therefore R2 =h 2 , the radii can be solved for using[ 3]:

V2 = 7rfqh 2 to yield R2 (8) = 11.4 mn R2 (21) = 11.3 mn Because s is the total distance from the center Of V 2 to P and the thickness of the shadow shield is 0.61 m (2 ft) the variable p can be determined to be[3]:

p(8) =(11.4 + 0.61 +8)/(11.4) = 1.75 p(21) = (11.3 + 0.61 + 2.1)/11.3 = 2.90 Because self-absorpininelce 8 R=0.Vus of b'2 and b'" are listed in Appendix A.9 along with the corresponding G function values.

The resulting doses are listed in Table 4.4 and 4.5 4.4 Scattered Gamma Dose The gamma rays going upwards would be possibly scattered back to the ground by the steel dome or by the. air*. This scattered radiation is also called skyshine.

64

Table 4.4 Shadow Shield Penetration Doses (rem) at 8 Meters

-E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.10 6. 04E- 15 7.-24E-15 8.45E- 15 9.66E-15 1.09E-14 1.21 E-14 0.15 6.8 6E-09 8.23E-09 9.60E-09 1. IOE-08 1.23E-08 1.37E-08 0.20 1.33E-*07 1.60E-07 1.87E-07 2.13E-07 2.40E-07 2.'6 7E-07 0.30 2.87E-06 1.45E-06 4.02E-06 4.59E-06 5.17E-06 5.74E-06 0.40 3.48E 05 3.76E-05 4.03E-05 4.31E-05 4.58E-05 4.85E-05 0.50 7.70E-04 9.24E-04 1.08E-03 1.23E-03 1 .39E-03 1.54E-03 0.60 3. 18E-04 3.81E-04 4.45 E-04 5.08E-04 5.72E-04 6.35E-04 0.80 2. 02 E-03 .2 .42E-03 2.83 E-03 3.23E-03 3.64E-03 4.04E-03 1.00 8. 36E-04 1 OOE-03 1.17E-03 1.34E-03 1.51E-03 1.67E-03 1.50 6.59E-03 7.91E-03 9.23E-03 1-05E-02 1.19E-02 1.32E-02 2.00 2.59E-02 .3.11E-02 3.63E-02 4. 14E-02 4.66E-02 5.18E-02 3.00 1.11E-02. 1.33E-02 1.55E-02 1.77E-02 2.00E-02 2.22E-02 4.00 5.16E 6.1.9E-04 7.22E-04 8.26E-04 9.29E-04 1.03E-03 Total 4.81E-02 5 .77E-02 6.73E-02 7.69E-02 8.65E-02 9.61E-02

.Table 4.5: Shadow Shield Penetration Doses (rem) at 21 Meters.

-E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.10 3.37E-15 4.04E-15 4.72EB-15 5.39E-15 6-06E-15 6.76E- 15 0.15 5.OOE-09 6.OOE-09 7.OOE-09 8.OOE-09 9.OOE-09 1.60E-08 0.20 9.66E-08 1. 16E-07 1.35E-07 1 .54E-07 1.74E-07 1.93E-07 0.30 2.OOE-06 2.40E-06 2.79E-06 3.19E-06 3.59E-06 3.99E-06 0.40 2.39E-05 2.58E-05 2. 77E-05 2-96E-05 3.15SE-05 3.33E-05 0.50 5.32E-04 6-38E-04 7.45E-04 8.51E-04 9.57E-04 1.06E-03 0.60 1.83E-04 2.19E-04 2.56E-04 2.92E-04 3.29E-04 3.66E-04 0.80 1-34E-03 1.61E-03 1.88E-03 2.15E-03 2.42E-03 2.68E-03 1.00 3.14E-04 3,77E-04 4.39E-04 5.02E-04 5.65E-04 6. 28E-04 1.50 3-10E-03 3.72.E-03 4.34E-03 4.96E-03 5.58E-03. 6.19E-03 2.00 1.25E.-0.2 1.50E-02 1.75E-02 2.OOE-02 2. 25E-02 2.50E-02 3.*00 .4.77E-03 5.73E-03 6.68E-03 7.64E-03 8.59E-03 9.54E-03 4.00 2.78E-04 3.34E-04 3. 89E-04 4.45E-04 5.01E-04 5.56E-04 Total 2.31E-02 .2.77E-02 3.23E-02 3.69E-02. 4. 15E-02 4.6 1E-02 65

Because forward scattering is favored for high energy photons, the effect of sources located at different positions within the containment will be different. Thus the containment volume will be divided into two regions. One is the dome portion (V~,,)

above the shadow shield, where the photons only need to be scattered through small angles. The other is the portion below the shadow shield (VI), where the photons need to be scattered through large angles. Volume V, will be further subdivided into three portions with different heights. For each volume portion, the source is assumed to be a point source with the total activity of that part of volume located at the center of the volume.

The relationships between the volumes are:

V,, 0. 3 V, V,0.7 V.

4.4.1 Air Scattering Gamma Dose The air scattering two hour dose (in rem) from sources for each energy group in Vu is[3]:

Dose = 1.5 -V O4u0tab 1

-0O&b dO=(O = + 0) (4.21) 4x 40d Similarly, the air scattering two hour dose from a source in V, is Dose = 5 -f4i~iel -. ~ ~ d du 8 (9 = + qS) (4.22) 127rx 0 d where:

N: electron density in air at STP (3.6 x 1020 electron/cm 3 )

E: incident photon energy, in MeV, P.a: approximate photon absorption coefficient of air for photon energy E, in cm 2 /g, 66

bl: steel thickness in: number of mean free path, 00: initial value of 0, In radians, 2k0: initial value of V), in radians, d~s: Klein-Nishina differential scattering energy cross section, in cm 2 /steradian,,

given by:

dit =r 2 ( - (4.23) dQ 2 E E(l E sn9 where r2: classical radius of the electron = 2.818x10-13 CM, E': scattered phot .on energy, in MeV. The quantities E and E' have the relation-ship:

E ++/- ~(1 - COS) (4.24)

The above equation was evaluated for each energy group using the numerical program package Maple TM. The resulting air scattering doses from the upper source at 8 meters for each power level are listed in Table 4.6, and those at 21 meters are listed in Table 4.7. .The resulting air scattering doses from the lower source at 8 meters for each power level are listed in Table 4.8, and those at 21 meters are listed in Table 4.9. The resulting air scattering doses from all sources at 8 meters for each power level are listed in Table 4.10, and those at 21 meters are listed in Table 4.11.

We can see that although the upper port ion has a smaller volume and therefore a smaller total radiation source strength, they contribute more to the total air scattered dose.

4.4.2 Steel Shell Scattering Dose The dose due to a single scattering of a photon with the steel wall can be approximated as [31:

67

Table 4.6: Air Scattering Doses (rem) From Upper Source at 8 Meters E (MeV) 5 MW. 6MW 7MW 8MW 9MW 10MW 0.03 2.43E-28 2.92E-28 3.40E-28 3.89E-28 4-37E-28 4.86E-28 0.04 5.1 9E- 19 6. 23E- 19 7.27E-19 8.31E-19 9.35E.-19 1.04E-18 0.05 8.39E-10 1.01E-09 1.18E-09 1.34E-09 1.51E-09 1.68E-09 0.06 6.98E-08 8.38E-08 9-77E-08 1.12E-07 1.26E-07 1.40E-07 0.08 3.21E-04 3.85E-04 4-50E-04 5.14E-04 5.78E-04 6.42E-04 0.10 6.43E-07 7.72E-07 9.01E-.07 1.03E-06 1.16E-06 1.29E-06 0.15 2.85E-03 3.42E-03 3-99E-03 4.56E-03 5. 13E.-03. 5.71E-03 0.20 4.09 E-03 4.91E-03 5.73E-03 6.55E-03 7.36E-03 8. 19E-03 0.30 6.59E-03 7.91E-03 9.23E-03 1.05E-02 1.19E-02 1.32E-02 0.40 1.'64E-02 1.77E-02 1.89E-02 2.02E-02 2.15E-02 2.28E-02 0.50 1.13E-02 1.36E-02 1-58E-02 2.03E-02 2. 26E-02 0.60 2.45E-02 *2.94E-02 3.42E-02 3.91E-02 4.40E-02 4.89E-02 0.80 3. 74E-02 4.49 E-02 5-23E-02 5.98E-02 6.73E-02 7.47E-02 1.00 4.25E-03 5.10E-03 5.95E-03 6.80E-03 7.66E-03 8.51E-03 1.50 6.02E-03 7.22 E-03 8.42E-03 9.63E-03 1.08E-02 1.20E-02 2.00 8-14E-03 9.77E-03 1. 14E-02 1-.30E-02 1.46E-.02 1.63E-02 7.54E-04 .9-04E-04 1.06E-03 1.21E-03 1.36E-03 1.51E-03 4.00 1-61E-05 1.93E-05 2.25E-05 2.57E-05 2.89E-05 3.21 E-05 Total 1.23E-01 .1.45E-01 1-68E-01 1.90E-01 2.13E-01 2.35E-01 68

Table 4.7: Air Scattering Doses (rem) From Upper Source at 21 Meters

-E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.03 2. 26E-28 .2.72E-28 3.17E-28 3.62E-.28 4.07E-28 4. 53E-28 0.04 4. 86E-19 5..84E- 19 6.81 E-19 7. 78E- 19 8.75E-19 9.73E-19 0.05 7.92E-10 9.51 E-10 .1.11 E-09 1.27E-09 1.43E-09 1.59E-09 0.06 6.664E-08 7.97E-08 9.30E-08 1.06E-07 1.20E-07 1 .33E-07 0.08 3.1OE-04 3.72E-04 4. 34E-04 4.96E-04 5.58E-04 6.20E-04 0.10 6. 28E-07 7.53 E-07 8.79E-07 1 OOE-06 1.13E-06 1.26E-06 0.15 2.88E-03 3 .46E-03 4-03E-03 4.61E-03 5.19E-03 5.77E-03 0.20 4.26E-03 5.11 E-03 5.97E-03 6.82E-03 7.67E-03 8. 53E-03 0.30 7.27E-03 8.72E-03 1.02E-02 1. 16E-02 1.31E-02 1.45E-02 0.,40 1.90E-02 2.05E-02 2.19E-02 2.34 E-02 2.49E-02 2.64E-02 0.50 1.36E-02 1.63E-02 1.91E-02 2.18E-02 2.45E-02 2.72E.-02 0.60 3.08E-02 3.69E-02 4.31 E 4.92E-02 5-54E-02 6. 15E-02 0.80 4.95E-02 5.94E-02 6.93 E-02 7.92E-02 8.90E-02 9.89E-02 1.00 6.19E-03 7.43 E-03 8.67E-03 9.91E-03 1.11 E-02 1.24E-02 1.50 1.01E-02 1.22E-02 1.42E-02 1.62E-02 1.82E-02 2.03 E-02 2.00 1.54E-02 1.85E-02 2.15E-0.2 2.46E-02 2.77E-02 3.08 E-02 3.00 1.50E-03 1.80E-03 2.10E-03 2.39E-03 2.69E-03 2.99E-03 4.00 4.25E-05 5.10E-05 5 .95E-05 6.80 E-05 7.66E-05 8.50E-05 Total 1.61E-01 .1.91 E-0 1 2. 20E-01 2 .50E-01 2.80 E-0 1 3.10QE-01 Table 4.8: Air Scattering Doses (rem) From Lower Source at 8 Meters Source Point 5 MW 6MW 7MW 8MW 9MW 10MW S, 5.14E-02 6-08E-02 7.02E-02 7.96E-02 8.90E-02 9-84F,02 S2 2.91E-02 3.44E-02 3.97E-02 4.50E-02 5.04E-02 5-57E-02 S3 1.78E-02 2.10E-02 2.43E-02 2.75E-02 3. 08E-02 3.40E-02 Total 9.83E-02 1.16E-01 1.34E-01 1.52E-01 1.70E-01 1.88E-01 ITable 4.9: Air Scattering Doses (rem) From Lower So urce at 21 Meters Source Point 5 MW 6MW 7MW 8MW 9MW 10MW S, 5.69E-02 6.74E-02 7.78E-02 8.83E-02 9. 87E-02 1.09E-01 S2 3.09E-02 *.3.65E-02 4.22E-02 4.78E-02 5.35 E-02 5.91 E-02 S2 1.86E-02, 2.20E-02 2.55E-02 2.89E-02 3. 23E-02 3.57E-.02 Total 1.06E-01 1.62E-01 1.46E-01 1.65E-01 1.85E-01 2.04E-01 69

Table 4.10: Air Scattering Doses (rem) From All Sources at 8 Meters E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.03 4.92E-28 5.91E-28 6.89E-28 7.88E-28 8.86E-28 9.85E-28 0.04 1.05E-18 1.26E-18 1.47E-18 1.68E-18 1.89E-18 2.1OE-18 0.05 1.69E-09 2.03E-09 2.36E-09 2.70E-09 3.04E-09 3.39F,09

ý0.06 1.40E-07 1.68E-07 1.96E-07 2.24E-07 2.52E-07 2.80E-07 0.08 6.40E-04 7.68E-04 8.95E-04 1.02E-03 1.15E-03 1.28E-03 0.10 1.27E-06 1.53E-06 1.78E-06 2.03E-06 2.29E-06 2.55E-06 0.15 5.56E-03 6.68E-03, 7.79E-03 8.90E-03 1.OOE-02 1.11E-02 0.20 7.97E-03 9.44E-03 1.10E-02 1.26E-02 1.42E-02 1.57E-02 0.30 1.25&-02 1.50E-02 1.75E-02 1.99E-02 2.24E-02 2.49E-02 0.40 3.04E-02 3.28E-02 3.52E-02 3.76E-02 4.OOE-02 4.24E-02 0.50 2.06E-02 2.47E-02 2.88E-02 3.30E-02 3.71E-02 4.12E-02 0.60 4.42E702- 5.30E-02 6.18E-02 7.07E-02 7.95E-02 8.84E-02 0.80 6.59E-02 7.91E-.02 9.23E-02 1.05E-01 1.19E-01 1.32E-01 1.00 7.52E-03 9.03E-03 1.05E-02 1.20E-02 1.35E-02 1.51E-02 1.50 1.04E-02 1.25E-02 1.46E-02 1.67E-02 1.88E-02 2.09E-02 2.00 1.40E-02 1.68E-02 1.96E-02 2.24E-02 2.52E-02 2.80E-02 3.00 1.22E-03 1.47E-03 1.71E-03 1.96E-03 2.20E-03 2.45E-03 4.00 2.80E.-05 3.36E-05 3.92E-05 4.48E-05 5.04E-05 5.60E-05 Total 2.21E-01 2.61E-01 3.02E-01 3.42E-01 3.83E-01 4.23E-01 70

Table 4.11: Air Scattering Doses (rem) From All Sources at 21 Meters E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.03 4.56E-28 5.47E-28 6.38E-28 7.29E-28 8.20E-28 9.11E-28 0.04 9.75E-19 1.17E-18 1.36E-18 1.56E-18 1.75E-18 1.95E-18 0.05 1.58E-09 1.90E-09 2.21E-09 2.53E-09 2.85E-09 3.17E-09 0.06 1.32E-07 .1.58E-07 1.85E-07 2.11E.-07 2.37E-07 2.64E-07 0.08 6.10E-04 7.32E-04 8.54E-04 9.76E-04 1-10E-03 1.22E-03 0.10 1.22E-06 1.47E-06 1.71E-06 1.96E-06 2.20E-06 2.45E-06 0.15 5.51E-03 6.61E-03 7.71E-03 8.81E-03 9-91E-03 1.10E-02 0.20 7.98E-03 9.57E-03 1. 12E-02 1.28E-02 1.44E-02 1,.60E-02 0.30 1.32E-02 1.58E-02 1.85E-02 2.11 E-02 2.38E-02 2.64E-02 0.40 3.35E-02 3.62E-02 3.88E-02 4.15E-02 4.41E-02 4'.67E-02 0.50 2.34-E-02 2.81E-02 3.28E-02 3.75E-02 4.22E-02 4.69E-02 0.60 5.20E-02 6.24E-02 7.28E-02 8.32E-02 9.36E-02 1.04E-01 0.80 *8.09E-02 9.71.E-02 1.13E-01 1.30E-01 1.46E-01 1.62E-01 1.00 9.92E-03 1. 19E-02 1.39E-02 1.59E-02 1.79E-02 1.99E-02 1.50 1.54E-02 1.85E-02 2.16E-02 2.47E-02 2.77E-02 3.08E-02 2.00 2.25E-02 2.70E-02 3.15E-02 3.59E-02 4.04E-02 4.49E-02 3.00 '2.11E-03 2.53E-03 2.95E-03 3.37E-03 3.79E-03 4.21E-03 4.00 .5.87E-05 7.04E-05 8.22E.-05 9.39E-05 1.06E-04 1. 17E-04 Total 2.67E-01 3.17E-01 3.66E-01 4.15E-01 4.65E-01 5.14E-01 71

Dose =1. 15 x1O 4 SNSTVSTEIt~e b1 f0 2 dV) 0 dOo ( +0 (.5 47rl2(2 1)(412-0) 0 d41-j(= P+4) 4.5 where:

NST: the electron. density in steel in STP, 2.19x l0 2'electron/ cm',

VST: the volume of the steel in the dome:

VsT (8) = -8.19 x 105 c'm3 ,

VST(21) = 2.91 X 106 cm 3 ,

The dose due to double scattering of a photon with the the steel wall can be

'approximated as:

Doe-1.15 x lO-4 SNS2+VSTV'STEfA~eb1 0~2 042 da, (0 dor(, (.6 Dose.= 2 2 x~2 -* 'P1(1 ]W'(426 do (0))

where V'ST is the volume of steel between the two scattering points, and 0' is the second scattering angle., The effect of the double scattering has been estimated by evaluating equation' 4.26 for'the three energies (E =0.4, 0.8 and 2.0 MeV) which contribute the most to'the total double steel scattering dose. The results indicate that the total steel scattering dose should be increased by a factor of 1.20 at 8 meters and by a factor of 1.02 at 21 meters[31.

The single steel scattering doses (rem) at 8 meters and at 21 meters versus source are listed in Tables 4.12 and 4.13. The total steel scattering doses after including the double steel scattering effect. are listed in Table 4.14.

-4.5 Radiation Penetration Through the Truck Lock The truck lock is a, rectangular steel tube 8 meters long closed at 72

Table 4.12: Single Steel Scattering Doses (rem) 8 Meters vs. Source Source Point 5 MW 6MW 7MW 8MW 9MW 10MW Upper 1.95E.-0.1 2.32E-01 2.68E-01 3.05E-01 3.42 E-0 1 3.78E-01 Point 1 5 .30E-02 6.29E-02 7.2 7E-02 8. 26E-02 9.25E-02 1.02E-01 Point 2 2.98E-02 3.53E-02 4.08E-02 4.64E-02 5.1 9E-02 5.74E-02 Point 3 1.73E-02 2.,05E-02 2.37E-02 2.68E-02 3.O00E-02 3.32E-02 Total 2.95E-01 3. 50E-0 1 4.06E-01 4.61 E-01 5.16E-01 5.71 E-0 1 Table 4.13: Single Steel Scattering Doses (rem) 21 Meters vs. Source Source Point 5 MW 6MW 7MW 8MW 9MW 10MW Upper 3.83E-01 4.55E-01 5.28E-01 6.OOE-01 6.73E-01 7.45E-01 Point 1 8.39E-02 9.97E-02 1.15E-01 1-31E-01 1.47E-01 1.62E-01 Point 2 3 '49E-.02 4.14E-02 4.79E-02 5.43E-02 6-08E-02 6.73E-02 Point 3 2.02E-02 2.39E-02 2.76E-02 3.14E-02 3.51E-02 3.88E-02 Total 5.22E-01 6.20E-01 7.19E-01 8.17E-01 9.16E-01 1-01E+00 Table 4.14: Total Steel Scattering Doses (rem)

Target 15 MW 6MW 7MW 8MW 9MW 10MW 8m 3.54E-01 4.21E-01 4.87E.-01 5-53E-01 6.19E-01 6.86E-01 21m 5.32E-01 6.33E-01 7. 33E-01 8.34t-01 9.34E-01 1.03E-00 73

The radiation reaching the lock will be treated as a point source located at the center of the inner surface of the inner door. The source strength for the truck lock penetration is the total source strength in the containment times a geometry factor.

The resulting source strength is ST =7.12x10 3 1S, photon/s, where S, is the total source strength of the containment.

4.5.1 Concrete Scattered Dose Unattenuated Dose at, the Concrete Wall The dose on the concrete wall before penetration is determined as:

=o 1.15 x 10-4 BSTEy,u e-EIT (4.27) where B: point buildup factor for steel, EptT: number of mean free paths through the two doors, x: distance to the wall, in cm.

The values of the corresponding doses on the concrete wall are listed in Table 4.15..

Concrete Albedo, Dose The concrete albedo dose is that due to the back scattering of photons from the surface of the'truck lock side walls. It is found that the northern boundary would, receive the maximum dose. This dose can be arrived at[31:

Dose = 8.23 x 1-2 Do Cd(O E)x126+ (4.28) 1 + cosOosecO, where C1, C 2 are energy and material dependent constants, DO: incident dose, in -rem, and 0,: reflection angle, in radians, The resulting concrete albedo doses at the northern boundary are listed in Table 74

Table 4.15: Direct Dose at the Concrete Wall E (MeV) 5 MW -6MW 7MW 8MW 9MW 10MW 0.10 8.84E-09. 1.06E-08 1.24E-08 1.41E-08 1.59E-08 1.77E-08 0.15 3.57E-04 4.28.E-04 4.99E-04 5.71E-04 6.42E-04 7.14E-04 0.20 1.08E-03 1.30E-03 1.51E-03 1.73E-03 1.95E-03 2.16E-03 0.30 .3.31E-03 3.98E-03 4.64E-03 5.30E-03 5.96E-03 6.63E-03 0.40 1.26E-02 1.36E-02 1.46E-02 1.56E-02 1.66E-02 1.76E-02 0.50 .1.07E-02 1.29E-02 1.50E-02 1.72E-02 1.93E-02 2.15E-02 0.60 2.93E.-02 3.52F,02 4.11E-02 4.69E-02 5.28E-02 5.87E-02 0.80 6.27E-02 7.52E-02 8.78E-02 1.OOE-01 1.13E-01 1.25E-01 1.00 9.65E-03 1.16E-02 1.35E-02 1.54E-02 1.74E-.02 1.93E-02 1.50 2.28E-02 2.73E-02 3.19E-02 3.65E-02 4.10E-02 4.56E-02 2.00 4.43E-02, 5.31E-02 6.20E-02 7.08E-02 7.97E-02 8.85E-02 3.00 5.89E-03 7.07E-03 8.25E-03 9.43E-03 1.06E-02 1.18E-02 4.00 2.23E-04 2.68E-04 3.12E-04 3.57E-04 4.02E-04 4.46E-04 Total 2.03E-01 2.42E-01 2.81E-01 3.20E-01 3.59E-01 3.98E-01 4.16.

4.5.2 Scattered Dose The second way that gamma radiation can reach the exclusion area th-rough the truck lock is by scattering on the . The same equation used in the previous chapter can be used here:

Dose 1.1 x 10- 4 STNSTVSTE/If~e-EIT 4 02 f4p 2 dod '0 (4.29) where ST =7.12x10 3 'S. photons/s, and NST = 2.19x10 24 electrons/cm 3 . The location along the boundary that receives the maximum dose is defined by the pa-rameters:

r, = 9.14 x 102 CM, r2= 2.82 x10 3 CM,

=0.140 radians 02= 0.209 radians V)I= 0.436 radians 75

.. .Table 4.16: Concrete Albedo Dose (rem)

E (MeV) 5 MW 6MW 7MW 8MW 9MW 10MW 0.10 2.84E- 11 -3.41E-11 3.97E-11 4.54E-11 5.11E-11 5.69E-11 0.15 8.95E-07 1.07E-06 1.25E-06 1.43E-06 1.61E-06 1.79E-06 0.20 2.27&-06 2.72E-06 3.18E-06 3.63E-06 4.09E-06 4.54E-06 0.30 :5.45E-06 6.55E-06 7.64E-06 8.73E-06 9.82E-06 1.09E-05 0.40 1.71E-05 1.85E-05 1.98E-05 2.12E-05 2.25E-05 2.39E-05 0.50 .1.28E-05 1.54E-05 1.79E-05 2.05E-05 2.31E-05 2.56E.-05 0.60 3.14E-05 .3.77E-05 4.39E-05 5.02E-05 5.65E-05 6.28E-05 0.80 5.57E-05 6.69E-05 7.80E-05 8.92E-05 i.OOE-04 1.11E-04 1.00 7.47E-06 8.96E-06 1.05E-05 1.19E-05 1.34E-05 1.49E-05 1.50 1.39E-05 1.67E-05 1.94E-05 2.22E.-05 2.50E-05 2.78E-05 2.00 2.26E-05 2.71E-05 3.16E-05 3.61E-05 4.06E-05 4.52E-05 3.00 2.33E-06 2.79E-06 3.26E-06 3.73E-06 4.19E-06 4.66E-06 4.00 7.53E-08 9.04E-08 1.05E-07 1.20E-07 1.36E-07 1.50E-07 Total 1.72E-04 2.04E-04 2.37E-04 2.69E-04 3.01E-04 3.34E-04 V)i = 0.768 radians.

The volume -of the door was determined to be . An attenuation of was used, and the buildup and attenuation due to the air were neglected. The resulting maximum dose are listed in Table 4.17.

4.5.3 Sum'mary of Radiation Through the Truck Lock Comparison of the truck lock penetration doses with those from direct and scattered gamma doses show that it is much smaller (on an order of 3) and thus can be neglected.

76

Table 4.17: Scattered Dose (rem)

-E (MeV) .5 MW 6MW 7MW 8MW 9MW 10MW 0.10 3.73E- 11 4,48E-1 1 5.23E-11 5-98E-1 1 6.72E-11 7.49E-11 0.15 1.39E-06 1.66E-06 1.94E-06 2.22F,06 2.49E-06 2.77E-06 0.20 3.99E-06 4. 79E-06 5.59E-06 6-39E-06 7. 18E-06 7.99 E-06 0.30 1.24E-05 1.49E-05 1.73E-05 1.98E-05 2.23E-05 2.48E-05 0.40 4.56E-05 4.92E-05 5.28E-05 5.64E-05 6.OOE-05 6.36 E-05 0.50 3-83E-05 4-60E-05 5.37E-05 6. 14E-05 6.90E-05 7. 67E-05 0.60 1.07E-04 1.28E-04 1.50E-04 1.71E-04 1.92E.-04 2. 14E-04 0.80 2.12E-04 2.54E-04 2.96E-04 3.39E-04 3.81E-04. 4.23 E-04 1.00 3.10OE-05. 3.72E-05 4.34E-05 4.96E-05 5.57E-05. 6. 19E-05 1.50 6 *08E-05 7. 30E-05 8.51E-05 9.73 E-05 1.09E-04 1.22E-04 2.00 1.03E-04 .1.24E-04 1.44E-04 1.65E-04 1.85E-04 2.06E-04 3.00 1.10E-05 1.33E-05 1.55E-05 1.77E-05 1.99E-05 2.21E-05 4.00 3.25E-07 3.90E-07 4.55E-07 5.20E-07 5.85E-07 6.50E-07 Total 6.27E-04 7.46E-04 8.66E-04 9.86E-04 1-11E-03 1.23E-03 77

Chapter 5 Summary For the MIT research reactor, the design basis accident is the maximum credible accident. For a design basis accident, the worst case is that the four plates in the center of the fuel element with the hot channel melt completely. During such an accident the fission products contained in these four plates may be released into the RCS. The fission product activity in the fuel is assumed to be the maximum equilibrium value and was calculated deterministically. Argon-41 build up in the containment after isolation was also determined. Though the concentration in the containment would be five orders of magnitude higher than that in operating conditions, its activity was five to seven orders of magnitude lower than the activities of fission products from the fuel. The contribution of Ar-41 to the exclusion area dose was therefore neglected.

The fraction of the fission products that may be released from the melted fuel to the RCS are estimated as:

• 100% Noble Gases,

• 90% Cs and I,

• 23% Te,

• 1% Sr, Ba, and Ru,

• 0.01% La, Ce and others.

The release fractions from the RCS to the containment were estimated as:

78

• 30% for all elements and the release fraction that remained airborne in the two hour period in the con-tainment were estimated as:

• 100% Noble Gases,

• 30% Cs and 1,

• 90% others.

All these values were estimated based on current NUREG documents, experimental test results, and the results of the TMI-2 accident. Allowance was also made for reasonable margins. But because of the stochastic nature of the release process and the limitation of understanding all the physical and chemical processes involved, there are some uncertainties associated with these estimates.

After the fission products are released into the containment, part of them may be released from the containment to the outside environment through a building crack or through the stack. These would lead to pollution of the atmosphere and contribute to the whole body and thyroid dose. The stack is more efficient in mixing the pollutant plume with fresh air. Thus, a stack release would result in a smaller dose at each given point, but the distribution would be over a wider range, from 100 meter to 100 kilometer. For the release from a building crack, two sets of calculational methods were tried. Each gave different results. One is called "exact", that calculated by following the exact procedure provided in NRC regulatory guide 1.145. But actually this method is not appropriate for the MITR because the MITR exclusion area dis-tances are much smaller than those used in regulatory guide 1.145. By taking account of the short distances and thus a strong wake effect of the plume, we used a "exact" model and got a smaller dose compared to the "conservative" model.

'For those isotopes that were not released from the containment, the resulting direct and scat tered gamma doses were determined. The methods used were the same as those used by Mull [3]. The dose from the truck lock was negligible smaller compared to the direct and scattered gamma dose.

79

Table 5.1: Total Dose at 5 MW Component of the Dose Dose at 8m (Rem) Dose at 21m (Rem)

Whole-body :

Containment Leakage 1.38E-02 1.38E-02 Steel Dome Penetration 6.60E-03 5.09E-02 Shadow Shield Penetration 4.81E-02 23E0 Air Scattering 2.21E-01 2.67E-01 Steel Scattering 3.54E-01 5.32E-01 Total 0.644 0.887 Thyroid:

Containment Leakage 1.12E-01 1.1213-01 The whole body dose which includes gamma and beta dose and the thyroid doses from all sources at the front and back fences are listed below. In the whole body, dose, the scattering gamma doses contribute the highest portions, one or two order."

of magnitude greater than those from other sources. The results are listed in Tables 5.1 through 5.6. The exclusion area doses as, a function of reactor power are plotted in Fig. 5-1. The whole-body dose at 21 meters is greater than that at 8 meters. The thyroid doses at both distances are almost equal.

The regulation gives. a limitation of 300 rem for thyroid dose and 25 rem for whole-body dose. Our results show that the doses released in a postulated design basis accident of the MIT Research Reactor at a power level of 5 MW up to 10 MW are well below the limitation.

80

1.81 w

cc r) w 0

0 0

5 5.5 6 6.5 7 7.5 8 8.5 9 9.5 10 REACTOR POWER (MW)

Figure 5-1: Exclusion area doses as a function of reactor power. The solid lines are

• for whole-body' doses, and the solid-dash lines are for thyroid doses. The circle sign is for doses at 21 meters and the plus sign is for doses at 8 meters. Thyroid doses at 8 meters and at 21 meters are not. distinguishable in the plot.

81

Table 5.2: Total Dose at 6 MW Component of the Dose Dose at 8m (Rem) Dose at 21m (Remi)_

Whole-body :

Containment, Leakage 1.66E-02 1.66E-02 Steel Dome. Penetration 7.87E-03 6.07E-02 Shadow Shield Penetration 5.77E-02 2.77E-02 Air Scattering 2.61E-01 3.17E-01 Steel Scattering 4.21E-01 6.33E-01 Total 0.764 1.06 Thyroid:

Containment Leakage 1.3513-01 1.34E-01 Table 5.3: Total Dose at 7 MW Component of the Dose Dose at 8m (Rem) Dose at 21m (Rem)

Whole-body :

Containment Leakage 1.9413-02 1.93E-02 Steel Dome Penetration 9.1313-03 7.05E-02 Shadow Shield Penetration 6.73E-02 3.23E-02 Air Scattering 3.02E-01 3.66E-01 Steel Scattering 4.87E-01 7.33E-01 Total 0.885 1.22 Thyroid:

Containment Leakage 1.57E-01 1.56E-01

.Table 5.4: Total Dose at 8 MW Component of the Dose Dose at 8m (Rem) Dose at 21m. (Rem)

Whole-body:

Containment Leakage 2.22E-02 2.20E-02 Steel Dome Penetration 1.04E-02 8.02E-02 Shadow Shield Penetration 7.69E-02 3.69E-02 Air Scattering 3.4213-01 4.15E-01 Steel Scattering 5.53E-01 8.34E-01 Total 1.00 1.39 Thyroid:

Containment Leakage 1.79E-01 1.7913-01 82

Table 5.5: Total Dose at 9 MW

-Component of the Dose Dose at 8m (Rem) Dose at 21m (Rem)

Whole-body:.

Containment Leakage 2.50E-02 2.4813-02 Steel Dome Penetration 1.17E-02 9.OOE-02 Shadow Shield Penetration 8.65E-02 4.15E-02 Air Scattering .3.83E-01 4.65E-01 Steel Scattering 6.1 9E-01 9.34E-0 1 Total 1.13 1.56 Thyroid:

Containment Leakage 2.02E-01 2.01E-01 Table 5.6: Total Dose at 10 MW Component of the Dose Dose at 8m (Rem) Dose at 21m (Rem)

Whole-body :

Containment Leakage 2.77E-02 2.76E-02 Steel Dome Penetration 1.29E-02 9.98E-02 Shadow Shield Penetration 9.61E-02 4.61E-02 Air Scattering 4.23E-01 5.14E-01 Steel Scattering 6.86E-01 1.03E-00 Total 1.25 1.72 Thyroid:

Containment Leakage 2.25E-01 2.24E-01 83

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87

Appendix A Tables 88

Table A.1: Total Core Fission Product Inventory Isotope Half-life Ai(sec- 1 ) Yj (% Q (x 105 Ci) 5MW 6MW 7MW 8MW 9MW 10MW Kr 85m 4.36h 4.41E-5 1.5 0.6490 0.7788 0.9086 1.0384 1.1682 1.3000 87 78m 1.48E-4 2.7 1.1700 1.4040 1.6380 1.8720 2.1060 2.3400 88 2.77h 6.95E-5 3.7 1.6000 1.9200 2.2400 2.5600 2.8800 3.2000 Xe 131m 12.1Od 6.68E-7 0.03 0.0130 0.0156 0.0182 0.0208 0.0234 006 133m 2.3d 3.49E-6 0.16 0.0692 0.0830 0.0969 0.1107 0.1246 0.1380 133 5.27d 1.52E-6 6.5 2.8100 3.3720 3.9340 4.4960 5.0580 5.6200 135m 15.6m 7.40E-4 1.8 0.7780 0.9336 1.0892 1.2448 1.4004 1.5600 135 9.13h 2.11E-5 6.2 0.4130 0.4956 0.5782 0.6608 0.7434 0.8260 138 17m 6.79E-4 5.5 2.3800 2.8560 3.3320 3.8080 4.2840 4.7600 I 131 8.05d 9.96E-7 2.9 1.2500 1.5000 1.7500 2.0000 2.2500 2.5100 132 2.4h 8.02E-5 4.4 1.9000 2.2800 2.6600. 3.0400 3.4200 3.8100 133 20.8h 9.25E-6 6.5 2.8100 3.3720 3.9340 4.4960 5.0ý580 5.6200 134 52.5m 2.20E-5 7.6 3.2900 3.9480 4.6060 5.2640 5.9220 6.5700 135 6.68h 2.89E-5 5.9 2.5500 3.0600 3.5700 4.0800 4.5900 5.1000 Br 83 2.4h 8.02E-5 0.48 0.2080 0.2496 0.2912 0.3328 0.3,744 0.4150 84 30m 3.85E-4 1.1 0.4760 0.5712 0.6664 0.7616 0.8,568 0.9510 Cs 134 2.Oy 1.10E-8 0.0* 2.8600 3.4320 4.0040 4.5760 5.1480 5.7200 136 13d 6.17E-7 0.006* 0.4140 0.4968 0.5796 0.6624 0.7452 0.8280 137 26.6y 8.27E-10 5.9 2.3100 2.7720 3.2340 3.6960 4.1580 4.6200 Rb 86 19.5d 4.11E-7 2.8E-5* 0.6120 0.7344 0.8568 0.9792 1.1016 1.2200 Te 127m 90d 8.82E-8 0.056 0.0242 0.0290 0.0339 0.0387 0.01436 0.0484 127 9.3h 2.07E-5 0.25 0.1080 0.1296 0.1512 0.1728 0.1944 0.2160 129m 33d 24E7 0.34 0.1470 0.1764 0.2058 0.2352 0.2:646 0.2940 129 72m 1.60E-4 1.0 0.4320 0.5184 0.6048 0.6912 0.7776 0.8650 131m 30h 6.42E-5 0.44 0.1900 0.2280 0.2660 0.3040 0.3.420 0.3810 131 24.8m 4.66E-4 2.9 1.2500 1.5000 1.7500 2.0000 2.2:500 2.5100 132 77h 2.50E-6 4.4 1.9000 2.2800 2.6600 3.0400 3.4200 3.8100 133m 63m 1.83E-4 4.6 1.9900 2.3880 2.7860 3.1840 3.5820 3.9800 134 44m 2.63E-4 6.7 2.9000 3.4800. -4.0600 4.6400 5.2200 5.8000 89

Table A. 1! Total Core Fission Product Inventory Isotope Half-life :Ai(sec-') Yi M% Q, (x 10' Ci) 5MW 6MW 7MW 8MW 9M/W 10MW Sr 91 97h 1.99e-5 5.9 2.5500 3.0600 3.5700 4.0800 4.5900 5.1000 Ba 140 12.8d '6.27E-7 6.3 2.7200 3.2640 3.8080 4.3520 4.8960 5.4500 Ru 103 41d 1.96E-7 2.9 1.2500 1.5000 1.7500 2.0000 2.2500 2.5100 105 4.5h 4.28E-5 0.9 0.3890 0.4668 0.5446 0.6224 0.7002 0.7790 106 L.OY 2.20E-8 0.38 0.1640 0.1968 0.2296 0.2624 0.2952 0.3290 Rh 103 36.5h 5.-27E-6 0.9 0.3890 0.4668 0.5446 0.6224 0.7002 0.7790 Tc 99m 6.04h 3.19E-5 0.6 0.2590 0.3108 0.3626 0.4144 0.4662 0.5190 Mo 99 67h 2.881E-6 6.1 2.6400 3.1680 3.6960 4.2240 4.7520 5.2800 Sb 127 93h 2.07E-6 0.25 0.1080 0.1296 0.1512 0.1728 0.1944 0.2160 129 4.6h 4.32E-5 1.0 4.3200 5.1840 6.0480 6.9120 7.7760 8.6500 Nd 147 11.3d 7.10E-7 2.6 1.1200 1.3440 1.5680 1.7920 2.0160 2.2500 La 140 40.2h 4.79E-6 6.3 2.7200 3.2640 3.8080 4.3520 4.8960 5.4500 Ce 141 32d 2.51E-7 6.0 2.5900 3.1080 3.6260 4.1440 4.6,620 5.1900 143 32h 6.01E-6 6.2 2.6800 3.2160 3.7520 4.2880 4.8240 5.3600 144 290d 2.76E-8 6.1 2.6400 3.1680 3.6960 4.2240 4.7520 5.2800 Zr 95 63d 1.27E-7 6.4. 2.7700 3.3240 3.8780 4.4320 4.9860 5..5400 97 17h i-.13E-5 6.2 2.6800 3.2160 3.7520 4.2880 4.8240 5.3600 Nb 95 35d 2.29E-7 6.4, 2.7700 3.3240 3.87 80 4.4320 4.9860 5.5400 Table A.2: Values of N,'/NO3 for Neutron-Capture Influenced Isotopes at OTr 4 x 10131 Isotope Ns'/N20 35 Xe 135 1.05 x10 5-Cs 134 1.4 x 10-'

Cs 136 3.6 X 10-4

• Cs 137 1.5 x 10 0 Rb 86 8.0 X 10-4 90

Table A.3: Parameters for Calculating Atmospheric Doses by Isotope Isotope E'(Mev/di~s). P,' (Mev/d is) C' (rem/ '~) CT (rem/Ci inhaled)

Kr 85m 2.7E,01 3.64E-02 2.OE-O1 87 1.3E00 1.81E-01 9.7E-01 88 9.3E-01 4. 67E-0 1 2.OEOO Xe 131m- 16A.E-O 1 133m 2.33E-01 133 1.15E-01. 9-06E-03 3.9E-01 135m 5.27E-01 135 3.1E-O1 5.67E-02 9.1E-O1 138 8.OE-O01 9.4E-Ol I 131 1.9E-01. 8.72E-02 1.3E05 132 7.70E-O1 5.1 1E-01 6.6E03 133 4.23E-01 1.54E-01 1.2E05 134 8.10E-01 5.33E-01 1.1E03 135 4.7E-01 4.19E-01 4.3E04 Br 83 1.18E-02. 5.30E-01 84 1.56E00 *1.52E00 Cs 134 2.21E-01 3-50E-01 5.8E02 136 1.14E-O01 4-78E-01 6.9E02 137 1-71E-01 1.22E-01 M.EN2 Rb 86 5-93E-01 2.07E-02 5.0E02 Te 127m 2.43E-01 1.1 OE-03 1.6E-01 127 2.33E-01 9.36E-04 2.9E00 129m 5-33E-01 7-83E-03 4.3E01 129 4.83E-01 1 .47E-02 8.1E-O1 131m 3.OE-O1 3.14E-01 4.5E03 131 7.13E-01 3.4E-O1 132 7.3E,02 4.75E-02 4.8E04 133m 8.OE-O1 *6.5E-01 91

Table A.3: Parameters for Calculating Atmospheric Doses by Isotope Isotope El (Mev/dis) E(Mev/di s) C,' (R e m/ Ci (Rem per Ci inhaled)

Sr 91 8.9E-0I 1.69E-01 1.3E02 Ba 140 3.4E-01 4.44E-02 2.2E02 Ru 103 7.OE-02 1.11 E-0 1 5.2E01 105 3 .88E-01 1.79E-01 1.4E01 106 1 .3E-02. 4.31E-02 4.8E01 R~h 103 1.89E -01 1.82E-02 6.4E00 Tc 99m 3.06E-02 .4.6E01 Mo 99 4.1E-01' 3.64E-02 9.4E01 Sb 127 5.OE-O1 1.51E-01 1.0E02 129 6. 23E-01 2 .68E-01 3. 7E01 Nd 147 2.7E-01 3.14E-02 1.2E01 La 140 4.53E-01 5.67E-01 1.5E02

'Ce 141 1.94E-01, 1 .38E-02 6.OEOO 143 4.63E-01 6-81E-02 1.8E01 144 1 .03E-01 4.31E-03 5.1EOO Zr 95 1.32E-01 1-62E-01 7.9E01 97 6.37E-01 4.22E-02 7.7E01 Nb 95 5. 33E-01 1.66E-01 8.1E01 92

Table A.4: Gamnma Emission Energies by Isotope Isotope Photon Energy (Mev) and Distribution(%

Kr 85m 0..15.(78), 0.3 (14) 87 .0.4 (50), 0.8(8), 3.0(14) 88 0.03 (2), 0.15 (7), 0.2(35), 0.4(5), 0.8 (23),

1.5 (14), 2.0 (53)

Xe 131m 0.15 (2) 133m 0..2 (10) 133 0.08 (37) 135m 0.5 (81) 135 0.3 (91), 0.6 (3) 138 0.15 (10), 0.3 (30), 0.4 (12), 0.5 (3), 2.0 (37)

I 131 0.08 (3), 0.3 (5), 0.4 (82), 0.6 (7), 0.8 (2) 132 0.5 (20), 0.6 (99), 0.8 (85), 1.0 (22), 1.5 (8), 2.0 (2) 133 0.5 (86) 134 0.15 (3), 0.4 (8), 0.5 (8), 0.6 (18), 0.8 (160),

1.-0 (11), 1.5 (9),7 2.0 (5) 135 0.4 (6), 0.8 (8), 1.0 (38), 1.5 (46), 2.0 (10)

Br 83 0.5 (1.4) 84 0.6 (1), 0.8 (48), 1.0 (8), 2.0 (25), 4.0 (7)

Cs 134 0.5(1)-7 0.6 (121), 0.8 (95), 1.0 (3), 1.5 (3) 136 0.06 (11), 0.08 (6), 0.15 (36), 0.3 (71), 0.8 (100),

10(82), 1.5 (20) 137 0.6 (85)

Rb 86 1,0 (9)

Te 127m 0.06 (1) 127 0.4 (1) 129m 0.6 (3)

'129 0.0.3 (17), 0.5 (7), 1.0 (1) 131m '0.08 (2), 0.10 (5), 0.2 (16), 0.3 (9), 0.8 (91),

1.0 (24), 1.5 (3), 2.0 (3) 131 0.15 (68), 0.5 (21),.0.6 (4), 1.0 (13) 132 0..05 (14), 0.2 (88) 133m 0.3 (11), 0.4 (1), 0.6 (23), 0.8 (8), 1.0 (89) 134 .0.08 (21), 0.2 (48), 0.3 (21), 0.4 (19), 0.5 (35),

0.8'(45).

93

Table A.4: Gamma Emission Energies by Isotope Isotope Photon Energy (Mev) and Distribution()

Sr 91 0.6 (15), 0.8 (27), 1.0 (33), 1.5 (5)

Ba 140 0.03 (11), 0.15 (6), 0.3 (6), 0.4 (5), 0.5 (34)

Ru 103 0.5 (88.), 0.6 (6) 105 0.3'(17), 0.4 (6), 0.5 (20), 0.6 (16), 0.8 (48) 106 -

Rh. 103 03(24)

Tc 99m 0.15 (90)

Mo 99 0.04 (2), 0.2 (7), 0.4 (1), 0.8 (16)

Sb 127 0.06 (1), 0.3 (3), 0.4 (9), 0.5 (29), 0.6 (45),

0.8 (17) 129 0.4 (5) , 0.5 (21), 0.6 (12), 0.8 (58), 1.0 (46)

Nd 147 0.10 (28), 0.3 (3), 0.4 (4), 0.5 (13)

La 140 0.3 (20), 0.5 (40), 0.8 (19), 1.0 (10), 1.5 (96), 3.0 (3)

Ce 141 0. 15 (48) 143 0.06 (11), 0.3 (46), 0.5 (3), 0.6 (7), 0.8 (10), 1.0 (1) 144 0.08 (2), 0.15 (11)

Zr 95 0.8 (98) 97 0.6 (92)

Nb 95 0.6 (100) 94

Table A.5: Attenuation and Absorption Coefficients Gamma Energy E(Mev) p, (cm-') P1 ST (CM'1) Pta (CM 2 /g) E,.i (MeV) gham7/g 0.03 2.*63 59.9 0.148 0.027 0.148 0.04 1.'31 26.3 0.0668 0.035 0.0668 0.05 0.848 14.0 0.0406 0.042 0.0668 0.06 0.642 8.60 0.0305 0.049 0.0406 0.08 0.470 4.19 0.0243 0.061 0.0305 0.10 0.402 2.61 0.0234 0.072ý 0.0243 0.15 0.329 1.41 0.0250 0.095 0.050 0.20 0.294 1.07 0.0268 0.112 0.0268 0.30 0.251 0810.0287 0.138 0.0287 0.40 0.225 0.707 0.0295--- --- 071.l56- -- Cl-.0295-----

0.50 0.205 0.636 0.0297 0.169 0.0297

.0.,60 0.190 0.584 0.0296 0.179 0.0297 0.80 0.166 0.510 0.0289 0.194 0.0297 1.0 0.150 0.460 0.0280 0.204 0.0297 1.5 0.123 0.373 0.0256 0.218 0.0297 2.0 0.105 0.32 6 0.0237 0.227 0.0297 3.0 0.0858 0.276 0.0211 0.235 0.0297 4.0 0.0750 0.254 0.0195 0.240 0.0297 95

Table A.6:' Shield Thicknesses in Mean Free Paths Gamma Energy E(Mev) Concrete (b) Steel (b1 ) Total (b2 )

0.0.3 160.4 56.9 217.3 0.04 79.9 25.0 104.9 0.05 51.7 13.3 65.0 0..06. 39.2 8.17 47.4 0.*08. 28.7 3.98 32.7 0.10 24.5 2.48 27.0 0.15 20.1 1.34 21.4 0.20 17.9 1.02 18.9 0.30 15.3 0.780 16.1 0.40 13.7 0.672 14.4 0.50 12.5 0.604 13.1 0.60 11.6 0.555 12.2 0.80. 10.1 0.485 10.6 1.0 9.15 0.437 9..57 1.5 7.50 0.354 7.85 2.0 6.41 0.310 6.72

.3.0 5.23 0.262 5.49 4.0 4.85 0.241 4.82 Table A.7:: Point Isotopic Source Exposure Build-Up Factors for Iron (Steel)

GamnmaEnergy E b (/pT)

(14ev) 1 2 3

.110 1.5 2.2 3.1 0.1-5 1.75 2.65 4.2 0,.20 2.0 3.1 5.3 130 2.05 3.15 5.8

.0.40 2.1 3.3 6.0

.0;.50 1.98 3.09 5.98 a60 1.96 3.02 5.90 1L80 1.91 2.95 5.62 1.87 2.89 5.39 1.0 1.82 2.66 4.76 1.76 2.43 4.13 3.0 1.55 2.15 3.51 4.0 1.45 1.94 3.03 96

Table A.8: Coefficients of the Taylor Exposure Build-up Factor Formula Gamma Energy E Concrete Steel (Mev) A a1l a2 A al C1 2 0.10 139.5813 -0.04127 -0.02927 - - -

0.15 97.7 220 -0.08301 -0.06400. - - -

0.20 87.8408 -0.10004 -0.07912 - - -

0.30 80.5000' -0.10500 -0.08400 - - -

0.40 46.6038 -0.10489 -0.07132 - - -

0.50 67.3716 -0.09198 -0.07061 31.379 -0.06842 -0.03742 0.60 70.0000 -0.08 400 -0.06500 30.095 -0.06694 -0.03486 0.80 65.7882 -0.07061 -0.05247 27.526 -0.06390 -0.02975 1.0 77.7911 -0.05818 -0.04420 24.957 -0.06086 -0.02463 1.5 15.1893 -0.06012 0.00252 21.290 -0.05357 -0.01495 2.0 17..1222 -0.04488 0.00448 17.622 -0.04627 -0.00526 3.0 13.7579 -0.02849 0.02761 13.218 -0.04431 -0.00087 4.0 14.2241 -0.02223 0.02316 9.624 -0.04698 0.00175 Table A.9: Values- of the Functions G(l,p,0,b') and G(l,p,0,b")

Gamma Energy E b'2 b/2 G(b')2 G(bI)2 G(b')2 G(b")2 (Mev) p=1.'75 p= 1 .7 5 p==2.90 p= 2 .90 0.10 25.9 26.2 5.4X10- 13 3.85 x1-3 4.5x 1013 3.6xl10' 0.15 19.6 20.0 3.9X10- 10 2.62x10 1 0o 3.05x 1010 2.07X 10-10 0.20 17.0 17.4 6.Ox 10- 3.9 x 10-9 4.6 x 10-9 3.0x10-9 0.30 14.4 14.7 9.4 x10-8 6.8 x 10-8 6.9 x 10-8 5.0 X10-'

0.40 12.9 '13.4 4.4x10-7 2.6 x 10-7 3.2 x 1- 1.9 X10-0.50 11.9 12.2 1.25x10-6 9.2 x 10-6 8.8 x 10-6 6.4X10 6" 0.60 11.2 11.4 2.7x 10- 6 2.2 x 10-6 1.8x 10" 1.5X10-.

0.80 9.85 -10.0' 1. 15 x10-5 9.84 x 10-6 7.4 x 10-6 6.23 x 10-6 1.0 9.03 9.17 2.8 x10-1 2.4 x 10-5 1.7xl10 5 1.55 x 10'-

1.5 7.38. 7.87 1.7xj10 4 9.6 x 10-4 9.2 X10-5 5.6 X10--

2.0 6.42 6.75 4.5 x10-4 3.2 x 10-4 2.45 x 10-4 1.8 X10)-4 3

3.0 5.33 .5.64 1.55x1V 1.05x 10-3 7.8 x 10-4 5.6 X10-4 4.0 4.74 4.93 2.95 x10-3 2.45 x 10-3 1.45 x 10-3 1.15x 10-3 97

Table A.10: Air Scattering Input Parameters Source Point ?ko(Radians) qOO(Radians) Wo(Radians) X(X1O3 cm) h'(m) R'(1m)_

8m:

Upper 0.349 0.314 - 2.27 - -

Point 1 0.544 0.486 0.395 2.08 1.60 10.8 Point 2 0.668 0.636 0.245 1.98 4.80 11.7 Point 3 0.730 0.794 0.086 1.93 8.00 .13.3 21mn:

Upper 0.138 0.070 - 3.43 - -.

Point 1 0.399 0.184 0.250 3.30 1.60 10.8 Point 2 0.572 0.285 0.194 3.24 4.80 11.7 Point 3 0.694 0.384 0.050 3.20 8.00 113.3 Table A.11: Steel Scattering Input Parameters

.Source Point V)i (Radians) V)2 (Radians) OI 1 (Radians) q$2 (Radians) r, (X103 cra) . 2X 10'C 8m:

Upper 0.349 0.679 0.314 0.390 1.13 1.61 Point 1 .0.544 0.950 0.486 0.563 1.19 1.70 Point 2 0.668 1.035, 0.636 0.713 1.34 1.71 Point 3 0.730 1.066 0.794 0.872 1.51 1.74 21m:

Upper 0.138 1.941 0.070 0.132 0.70 3.60 Point 1 0.399 1.821 0.184 0.248 0.97 3.60 Point 2 0.572 1.720 0.285 0.349 1.23 3.60 Point 3 0.694 1.621 0.384 0.448 1.46 3.60 98

Appendix. B Figures 99

10 1 0

w 100 100 101 WINDSPEED (M/S)

Figure B-i: Meander factors for correction of Pasquill-Gifford sigma y values by atmospheric stability class. D, E, F, and G are the stability classes.

100

(5i h2 R2 x--

Figure'B-2: Direct dose containment volume transformations 101

3o t/ý, -vcccL-LCJ2- Fr Reactor Power (MW) 5 5 6 6 7 7 8 8 9 9 10 10 Site (Meter) 8 8 21 8 18 1 41 8 21 Whole Body (Rem):

Containment Leakage Beta 4.29E-03 4.272- 03 5.15E-93 5.12E-03 6.0111-03 5.98E-03 6.87E-03 6.83E-03 7.72E-03 7.69E-03 8.56E-03 8.54E-03 Containment Leakage Gamma 5.6012-03 5.572- *03 6.72E-03 6.69E-03 7.84E-03 7.80E-03 8.96E-03 8.92E-03 1'.01E-02 11.002-02 1.122-02 1.112E-02*

Containment Leakage Total 9.89E-03 9.842- 03 1.19E-02 1.182E-02 1.39E-02 1.382-02 1.58E-02 1.582-02 1.782-02 1.77E-02 1.982-02 1.962-02 Steel Dome Penetration " 2.32E-03 1.792- 02 2.73E-03 2.11 E-02 3.15E-03 2.43E-02 3.56E-0W 2.75&2-O 3.972-03 3.072-02 4.39E-03 3.392-02 Shadow Shield Penetration 3.58E-02 1.73E--02 4.30E-02 2.0712-02 5.012E-02 2.422-02 5.732-02 2.77E-02 6.441E-62 3.112E-02 7.162-02 3.46E-02 Air Scattering 4.20E-02 5.51 E~ -02 4.84E-02 6.38E-02 5.48E-02 7.252-02 6.1122-02 8.132-02 6.77E-02 9.002-02 7.412E-02 9.87E-02 Steel Scattering 1.16E-01 1.812.-.01 1.34E-01 2.102-01 1.542-01 2.412-01 1.722-01 2.70E-01 1.91E2-01 3.001E-01 2.092-01 3.30E-01 Total 2.06E-01 2.812.-0O12.40E-01 3.28E-01 2.76E-01 3.76E-01 3.09E-01 4.23E-01 3.45E-01 4.69E-01 3.79E-01 5.17E-01 Thyroid (Rem):

Containment Leakage 1.13E-03 1..132- 03 1.362-03 1.35E-03 1.58E-03 1.58E-03 1.8112-03, 1.802-03 2.042-03 2.03E-03 2.27E-03 2.25E-03 Reactor Power (MW) 5 5 6 6 7 7. 8 8 9 9 .10 10 Site (Meter) 8 21 8 21 8 21 8 21 8 21 8 21 Whole Body (Rem):

Containment Leakage Beta 4.39E-03 4.37E- 03 5.27E-03 5.25E-03 6.152E-03 6.122-03 7.032-03 7.002-03 7.912E-03 7.87E-03 8.792-03 8.752-03 Containment Leakage Gamma 5.86E-03 5.83E. -03 7.03E-03 6.992-03. 8.20E-03 8.16E-03 9.37E-03 9.32E-03 1.052-02 1.052-02 1.172-02 1.17E-02 Containment Leakage Total 1.03E-02 1.02E- .02 1.2312-02. 1.222-02 1.44E-02 1.432-02 1.642-02 1.63E-02 1.842-02 1.842-02 2.051E-02 2.05E-02 Steel Dome Penetration 2.0912-02 1.79E. .02 3.20E-03 2.47E-02 3.69E-03 2.852-02 4.182-03 3.23E-02 4.67E-03 3.612E-02 5.172-03 3.982-02 Shadow Shield Penetration 3.69E-02 1.792- .02 4.43E-02 2.14E-02 5.17E-02 2.50E-02 5.912E-02 2.86E-02 6.64E-02 3.21 E-02 7.382-02 3.57E-02 Air Scattering 4.912E-02 6.45E--02 5.70E-02 7.51 E-02 6.48E-02 8.572-02 7.27E-02 9.63E-02 8.05E-02 1.072-01 8.8412-02 1.172-01 Steel.Scattering 1.372-01 2.112-01 1.O02-01 2.48E-01 .1.822-01 2.85E-01 2.052-01 3.202-61' 2.28E-01 3.57E-01 2.512E-01 3.93E-01 Total 2.542-01 3.22E-01 2.76E-611 3.81E2-01 3.17E-01 4.38E-01 3.58E-01 4.94E2-01 3.98E-01 5.512E-01 4.392-01 6.06E-01 Thyroid (Rem):

Containment Leakage 1.12E-02 1.122-02* 1.35E-02 1.34E-02 1.572-02. 1.562-02 11.80E-02 1.792-02 2.02E-02 2.0112-02 2.252-02 2.242-02