ML20203J908
| ML20203J908 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 08/01/1986 |
| From: | Agosti F DETROIT EDISON CO. |
| To: | Adensam E Office of Nuclear Reactor Regulation |
| References | |
| VP-86-0115, VP-86-115, NUDOCS 8608060068 | |
| Download: ML20203J908 (18) | |
Text
{{#Wiki_filter:- t c' M2," p Nuclear Operat6ons kd Detroit re-a Edison 5" F" Wa:1. E August 1, 1986 VP-86-Oll5 i Office of Nuclear Reactor Regulation Ms. Elinor G. Adensam, Director Project Directorate No. 3 Division of BWR Licensing U. S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Ms. Adensam:
Reference:
1) Fermi 2 NRC Docket No. 50-341 NRC License No. NPF-43 2) Detroit Edison to HRC Letter, " Service Life of Main Steam Bypass Line", VP-86-0097, dated July 22, 1986
Subject:
Radiological Analysis of a Rupture of One Main Steam Bypass Line Detroit Edison has analyzed a rupture of one main steam bypass line for radiological consequences to plant personnel. The steam and radiological impacts are presented in Enclosure 1. As stated in Enclosure 1, the analysis performed is ultra conservative and is inconsequential when compared to the design basis accident of a steam line break outside containment for both offsite and inplant consequences. Furthermore the consequences of a bypass line break during the startup test program are even less significant due to the low source inventory. The test program discussed in Reference 2 will not result in any significant exposure to individuals in the vicinity of the bypass line and normal industry safety precautions will be practiced. 0Y $$0$0341 f PDR
Ms. Elinor G. Adensam August 1, 1986 VP-86-0115 Page 2 Should you have any questions regarding this matter, j please contact Mr. R. L. Woolley (313) 586-4211. Sincerely, Oc + 4 cc: Mr. M. D. Lynch Mr. W. G. Rogers USNRC Document Control Desk Washington, D.C. 20555 i -{ i
Enclosure 1, Page 1 Radiological Analysis of a-Rupture of A Main Steam Bypass Line 1 NRC-NRR has requested that Detroit Edison analyze a rupture of a main steam bypass line for radiological consequences to plant personnel. This analysis is set forth in the following paragraphs. j 1.0 Plant Conditions Assur.ed 1.1 Fermi 2 is operating with two main steam bypass lines passing 25% of full rated steam flow to condenser (Table 1). 4 1.2 Activity in steam is representation of equilibrium core l inventory at 100% power (Table 2). } 1.3 Bypass line valves are full open. 2.0 Assumptions ' 2.1 Rupture occurs in one of the two main steam bypass lines downstream of open bypass valve (Figure 1). j 2.2 Steam flow through rupture is terminated when MSIV's close on high temperature signal in the Turbine Building area and i the bypass line valve closes to maintain pressure. I 2.3 There are no radiological consequences offsite because the i Turbine Building ventilation system is isolated by a signal from the SPING noble gas radiation monitor channel at radiological Technical Specifications limits. i 2.4 All of the radioactivity in the steam is released to the j Turbine Building environment. j 3.0 Description of Situations Analyzed l 3.1 All of the radioactivity is released na a homogeneous mixture to the volume of the third floor of'the Turbine l Building. The Tagging Center / Operational Support Center is located on this floor outside the main Control Room. This l area is not shielded and is occupied as a Tagging Center twenty-four (24) hours a day (Figure 2). 3.2 All of the radioactivity is released as a homogeneous mixture to a relatively small accessible area located on the second floor of the Turbine Building in the vicinity of the bypass line rupture. This analysis provides the upper limit dose / dose rate values. In reality, the steam would not be contained in this small area for a long period of time. The second floor is only occupied when making normal plant l rounds or during inspections (Figure 3). I ( ---,-r, e -. -,-- --..,,-_,- ..-me,--_..--.--,----.~,...,----,,--,---m,-,.-,,v--wnw w-. m,w m,----,,,,,.-----w.
.. ~. ~ Enclosure 1, Page 2 J ( 3.3 The effects of the steam pressure and temperature on personnel in the immediate vicinity of the bypass line rupture. 4.0 Conservatisims I 4.1 Bypass valve was assumed to be full-open and passing full 1 design flow. Startup procedures indicate that the bypass valves are not normally open beyond 55% during startup. 4.2 A 10-second delay time was assumed for the Termocoup.le response time before' initiation of MSIV closure. The instruments are located in close proximity to the bypass valves and should respond to resulting high temperature almost immediately. 4.3 Maximum five (5) seconds was assumed for MSIV closure and { flow rate was not reduced during closure time. 4.4 Both equilibrium core and 100% power level was assumed for activity in steam. 4.5 Radioactive release is assumed to be contained within the second or third floor of Turbine Building. No credit was i taken for mixing in full Turbine Building volume. ) 4.6 No credit was taken for percentage of steam entering l condenser. l 1 o 4.7 No credit was taken for plateout of iodines. 5.0 Summary of Results- ) 5.1 Table 3 is summary of the gamma, beta, and thyroid dose rates and doses to an individual on the third and second floors of the Turbine Building. l 5.2 Tables 4 and 5 list the gamma and beta dose rates by radionuclide versus time for the third floor of the Turbine Building. t l 5.3 Tables 6 and 7 list the beta dose rates by radionuclide versus time for the second floor of the Turbine Building. j 5.4 Tables 8 and 9 show a comparison of the limits of 10CFR20, Appendix B, Table 1, Column 1 with the activity I concentration versus time by radionuclide for the third and i second floor of the Turbine Building. I i
Enclosure 1, Page 3 6.0 Conclusions 6.1 Effect of Steam Release from Rupture The rupture was postulated to occur immediately downstream of the bypass valve since this sould result in the maximum break flow rate. Steam flow through the' bypass line is choked at the bypass valve throat. Therefore, the bypass valve reduces the steam pressure from 965 psia to 557 psia U (478 F saturated steam). As the steam exits the rupture, it undergoes a constant enthalpy expansion to atmospheric 0 pressure. This results in 303 F superheated steam discharging through the rupture at a velocity dependent on the rupture area assumed. Anyone within the area shown on Figure 3 would be affected by the steam as well as the radiation. 6.2 Effect of Radiation Released in Steam from Rupture As shown on Tables 3 through 9, the major effect is from the iodines released from the rupture. This is to be expected since the ratio of iodines to noble gases is approximately 10 to 1 at the initiation of the rupture as shown in Table 1. The analysis presented herein is ultra conservative and would apply only if Fermi 2 had reached 100% power operation. It would not apply to the power levels of 5%, 8%, and 12% in Reference (2) which are part of the startup test procedures. The low power levels at which Fermi 2 has l been operated to date has resulted in very low source terms and the consequence of any rupture at levels below 100% will t be insignificant. 4 6.3 Potential Mitigatino Pactors Detroit Edison has in place Abnormal Operating Procedures { 20.000.02, Abnormal Releases of Radioactive Material and 20.000.17T, Turbine Bypass Line Failure. These procedures communicate to the Control Room what actions to take should i a failure of the bypass line occur. I; i- ) w-rnm,---wv- --,--r,.wwe mwm w-me,-r-, ~ - - - --,,->,---e-v ,,n,,,,-, ,,,-r-n---a ,. awe--e -,s---esc, n~n,---m---vn--..,mwm
Enclosure 1, Page 4 During the startup phase, the test program will provide measured pipe wall stress data as stated in Reference (2). The instrumentation involved is remotely read and does not necessitate any personnel being in the vicinity of the bypass line. Limited visual inspections will be made of the pipes, under pressure, to inspect for leaks. 4 i ~
Enclosure 1, Page 5 i TABLE 1 ESTIMATE OF STEAM RELEASED FROM THE RUPTURE OF ONE MAIN STEAM BYPASS LINE i Estimate 10,000 lb. of steam released through rupture prior to MSIV 4 isolation and bypass valve closure to maintain pressure. Assumptions i o T = 1 sec (to raise temperature in Turbine Building area to 0 200 F - MSIV trip setpoint) 10 sec (instrument response) 5 sec (MSIV stroke time) 16 sec (Total rupture flow time) 6 (0.125) x (14.8 x 10 lb/hr) x (16 sec) = 8222 lb. steam (3600 sec/hr) o Additional 1000 lbs. steam released from piping prior to bypass line valve closure to maintain pressure. o Assume 10,000 lb. steam released to be conservative. Total steam released = 9222 lb. i l l 4 _,.., _... -, _. _,.,.,,. _ _. ~ _.
Enclosure 1, Page 6 TABLE 2 MAIN STEA!! BYPASS LINE RUPTURE ACTIVITY RELEASED TO TURBINE BUILDING (Curies)("} ID) Isotopes Activity I-131 1.594 E-1 I-132 1.471 I-133 1.091 I-134 2.942 I-135 1.594 Total Halogens 7.259 Kr-83m 7.173 E-3 Kr-85m 1.257 E-2 Kr-85 4.904 E-5 Kr-87 3.916 E-2 Kr-88 4.015 E-2 Kr-89 1.670 E-1 Xe-131m 4.006 E-5 Xe-133m 5.989 E-4 Xe-133 1.678 E-2 1 Xe-135m 4.911 E-2 Xe-135 4.529 E-2 Xe-137 2.207 E-1 Xe-138 1.670 E-1 Total !!oble Gases 7.656 E-1 (a) From FSAR Table 15B.6.4-3, Steam System Piping Break Outside i Containment I (b) Activity in 10,000 lb. steam i i
Enclosure 1, Page / ~ TABLE 3 IRIN SITR1 BYPASS LIIC RUISURE STJ1ARY OF GAIIR, BETA, ND 'IEYROID DOSE RATES RD DOSES 3rd floor Wrbine BuildinM_ t=0 min 15 min 30 min 1 hour Ga raa dose rate, mreWhr 26.2 22.5 20.1 16.6 Beta dose rate, mrefnr 73.7 56.4 49.7 40.4 L yroid dose rate, Re Whr 25.51 24.75 24.10 22.88 Thyroid dose, Rem 0 6.29 12.39 24.2 N-16 ga:xaa dose rate = 2.45 ReWhr maxinum at 16 seconds after rupture when release terminates. AT 90 seconds after rupture occurs, N-16 is effectively decayed to insignificant quantity. 2nd floor Turbine BuildinM_ t=0 min 5 min 10 min Gamma dose rate, mreW hr 206 191 185 Beta dose rate, mreVhr 1507 1330 1220 Thyroid dose rate, Redhr 522.44 510.91 512.56 Thyroid dose, Rem 0 43.2 86.2 N-16 ganra dose rate = 50.30 Reghr maxinum at 16 seconds after rupture when release terminates. At 90 secords after the rupture occurs, N-16 is effectively decayed to insignificant quantity. (a) Volume of 3rd floor of Wrbine Building = 1,738,080 cuft. (b) Volume of 2nd floor of Wrbine Buidling = 85,054 cuft. (Refer to Figure 3) I
Enclosure 1, Page 8 TABLE 4 GYIM DOSE IMTE VS TIME POST BYPASS LINE RUP'IUPE LOCATION: 3PD FIDOR 'IURGIIE BLDG l Gamna Dose Rate (mrefar) l Time Post Bypass Rupture Radionuclide 1 0 !!in l 15 !!in i 30 flin l 1 Hour l l l l I-131 1 0.23 1 0.23 1 0.23 l 0.22 I-132 l 3.84 l 3.56 l 3.31 l 2.85 I-133 l 2.18 l 2.16 l 2.15 1 2.11 I-134 l 11.20 l 9.17 l 7.51 1 5.03 I-135 l 6.69 l 6.52 l 6.35 l 6.03 l l l l Kr-83m l ul l ul l <G l ul Kr-85m l 9.0E-3 l 8.5E-3 l 8.2E-3 l 7.6E-3 Kr-85 l <d I <G l <G l ul Kr-87 l 0.20 l 0.18 l 0.15 l 0.12 Kr-88 l'0.21 l 0.19 1 0.18 l 0.16 Kr-89 l 0.81 l 3.2E-2 l 1.2E-3 l G1 Xe-131m i 1 l 1 l 1 l ul Xe-133ra l <d i <d l <d l ul Xe-133 l 4.9E-3 l 4.9E-3 l 4.9E-3 l 4.9E-3 Xe-135a l 9.lE-2 l 4.6E-2 l 2.4E-2 l 6.3E-3 Xe-135 l 4.0E-2 l 4.0E-2 l 4.0E-2 1 4.0E-2 Xe-137 l 0.15 l 1.0E-2 l <d l <d Xe-138 l 0.57 l 0.31 l 0.17 l 5.3E-2 l l l l 'Ibtal l 26.22 l 22.5 1 20.13 l 16.63
s e-
,e
Enclosure 1, Page 9 TABLE 5 BETA DOSE RATE VS TIZE POST BYPASS LI!E RUP'IURE IlX%TIOth 3TD FTOOR 'IURBIIIC BLDG l Beta Dose Rate (arem/hr) l Time Post Bypass Rupture 4 Radionuclide 1 0 Itin i 15 tlin 1 30 flin i 1 Hour l l l l I-131 l 0.49 l 0.49 l 0.49 1 0.48 I-132 l 12.0 l 11.2 l 10.4 l 8.9 I-133 l 7.8 l 7.7 1 7.6 l 7.5
- -1 34 l 30.0 l 24.5 l 20.1 l 13.5 I-135 l 9.9 l 9.6 l
9.4 l 8.9 l 1 l l Kr-83m* l-l- l-1- Kr-85m l 4.9E-2 l 4.6E-2 l 4.5E-2 l 4.lE-2 Kr-85 l <G l <<1 1 <C l <<1 Kr-87 1 0.87 l 0.76 l 0.66 l 0.50 Kr-88 l'0.24 1 0.23 1 0.22 1 0.20 Kr-89 l 3.8 1 0.15 l 5.4E-3 l <G xe-131m l-I- 1-l- xe-133ra l-l- l-l- xe-133 l 2.7E-2 l 2.7E-2 l 2.7E-2 l 2.7E-2 xe-135m l-I- l-l- xe-135 1 0.23 l 0.23 1 0.23 1 0,22 xe-137 ! 6.6 1 0.5 1 0.03 l <G xe-138 l 1.73 l 0.96 l 0.52 l 0.16 l l 1 l Total l 73.7 l 56.4 l 49.7 l 40.0
- Ibis ;tsotope has no betas, i
Enclosure 1, Page 10 TABLE 6 GAlta DOSE RATE VS TIME POST BYPASS LIIC RUP7JPE LT.^. TION: 2:D "IOOR 'I'JPOII:E PlOC l nrem/hr fluclide 1 0 Min 1 5 Min i 10 Min l l 1 I-131 1 1.85 l 1.85 l 1.85 I-132 l 30.31 1 30.31 l 28.82 I-133 l 17.52 l 17.52 l 17.45 I-134 l 87.50 l 81.81 1 76.83 I-135 l 52.81 l 52.33 l 51.93 i Kr-83m l <<1 l <<1 l <<1 Kr-85m l 7.24E-2 l 7.14E-2 l 7.05E-2 Kr-85 l < <1 l <<1 l < <1 Kr-87 l 1.56 l 1.49 l 1.42 Kr-88 l 1.64 l 1.61 l 1.57 Kr-89 l 6.34 l 2.15 l 0.726 Xe-131m l <<1 l <<1 l <<1 Xe-133m l < <1 l < <1 l < <1 Xe-133 l 3.94E-2 l 3.94E-2 l 3.94E-2 Xe-135m l 7.26E-1 l 5.8E-1 l 4.66E-1 Xe-135 l 3.22E-1 l 3.20E-1 1 3.17E-1 Xe-137 l 1.21 1 0.499 l 0.21 Xe-138 l 4.44 l 3.65 l 2.99 l l l l 206.34 l 191 l 184.7
Enclosure 1, Page 11 TABLE 7 B3TA DOSE RATE VS TII1C POST BYPASS LIIE RUP1URE UX:ATIGII: 2ID FIDOR TJRBIITE BLIX3 l Beta Dose Rate araa/hr l Time Post Bv;ms Rupture Radionuclide 1 0 !!in 1 5 liin i 10 :lin l l l I-131 l 10.0 l 10.0 l 10.0 I-132 l 245.2 l 245.2 l 233.8 I-133 l 159.4 l 159.4 l 158.3 I-134 l 612.9 l 574.2 l 539.2 I-135 l 202.3 l 200.0 l 198.5 Kr-83m* l-l- l-Kr-85m i 1.0 l 0.97 l 0.96 Kr-85 l <<1 l <<1 l <<1 Kr-87 l 17.8 l 17.0 l 16.2 Kr-88 l 4.9 l 4.9 l 4.8 Kr-89 l 77.6 l 26.5 l 8.9 Xe-131m l-l- l-Xe-133n I-l- l-Xe-133 1 0.6 l 0.6 l 0.6 Xe-135m l-l- l-Xe-135 1 4.7 1 4.7 1 4.6 Xe-137 l 134.8 l 55.2 l 22.7 Xe-138 l 35.3 l 29.0 l 23.7 l l l Total l 1.507E3 l 1.33E3 l 1.22E3
- 1his isotope has no betas.
y._ ,m--. g 7 m--
Enclosure 1, Page 12 TABLE 8 ACTIVITY CO; CENTRATION (CI/CC) VS TIZE FOR RADIO!KELIDES IN THE 'IU! SINE BUILDI:K; 3m I'IOOn V00,0:0 Ci/cc at tirae t 10CFR20 Appendix B nadionuclide 1 0 !!in l 15 !!in 1 30 !!in 1 60 Ilin l Table 1, Col 1 l l l l I-131 1 3.24E-12l 3.24E-121 3.24E-12l 3.18E-121 3E-13 I-132 1 2.99E-11l 2.77E-lli 2.57E-lll 2.22E-111 9E-13 I-133 l 2.31E-lll 2.29E-11l 2.27E-lll 2.23E-11l 2E-13 I-134 1 6.0E-ll l 4.91E-lll 4.02E-lll 2.70E-lll 3E-12 I-135 l 3.24E-11l 3.16E-lll 3.08E-lll 2.92E-11l 4E-13 Kr-83m l 1.45E-13l 1.33E-131 1.21E-13l 1.01E-13l - Kr-85a l 2.55E-13l 2.45E-13l 2.36E-131 2.18E-131 6E-12 Kr-85 l 9.95E-16l 9.95E-16l 9.95E-161 9.95E-16l lE-ll Kr-87 1 7.95E-13l '6.92E-13l 6.05E-13l 4.59E-13l lE-12 Kr-88 l 8.16E-131 7.68E-131 7.19E-13 l 6.38E-131 lE-13 Kr-89 l 3.39E-121 1.32E-13l 5.llE-15l 7.73E-18l - Xe -131m l 8.16E-16l 8.16E-16l 8.16E-16l 8.16E-16l 2E-ll Xe-133ra l 1.22E-14 1.22E-14l 1.22E-14l 1.22E-14l lE-ll Xe-133 l 3.41E-131 3.41E-13l 3.41E-131 3.41E-13l lE-ll Xe-135m l 1.0E-12 5 5.14E-131 2.64E-131 6.97E-14l - Xe-135 l 9.19E-13l 9.19E-13l 9.19E-131 8.54E-13l 4E-12 Xe-137 l 4.48E-121 3.13E-13l 2.17E-141 1.05E-161 - Xe-138 l 3.39E-12l 1.88E-121 1.03E-12l 3.16E-13 l -
Enclosure 1, Page 13 TABLE 9 ACTIVITY CO!CE!TfPATIOIT (CI/CC) VS TIE FOR RADIOIECLIDES IN 'IIC Tv33IIiB BUILDIIU 21D E7J3OR VG,i.IIE Ci/cc at time t 10CFP20 Appendix B Radionuclide I t=0 Min I t=5 !!in I t=10 t4inl Table 1, Col 1 l l l I-131 l 6.62E-lll 6.62E-lll 6.62E-11l 3E-13 I-132 l 6.11E-10l 6.11E-10l 5.81E-10l 9E-13 I-133 l 4.72E-10l 4.72E-10l 4.70E-10l 2E-13 I-134 l 1.23E-9 l 1.15E-9 l 1.08E-9 l 3E-12 I-135 l 6.62E-101 6.56E-10l 6.51E-10l 4E-13 Kr-83m l 2.96E-12l 2.87E-12l 2.79E-12l - Kr-85m l 5.21E-12l 5.14E-121 5.07E-12l 6E-12 Kr-'85 l 2.03E-14l 2.03E-141 2.03E-141 lE-ll Kr-87 l 1.62E-lll 1.55E-lli 1.48E-lll lE-12 Kr-88 l 1.67E-lll 1.64E-lll 1.60E-lll 1E-13 Kr-89 l 6.93E-11l 2.35E-lll 7.94E-12l - Xe-131m l 1.67E-14l 1.67E-14l 1.67E-14l 2E-ll Xe-133m l 2.49E-13 2.49E-13l 2.49E-13l lE-ll Xe-133 1 6.97E-12l 6.97E-13l 6.97E-121 lE-ll Xe-135m l 2.04E-ll5 1.63E-lll 1.31E-lll - Xe-135 l 1.88E-lli 1.87E-lli 1.85E-lll 4E-12 Xe-137 l 9.15E-lll 3.77E-lll 1.55E-lli - Xe-138 l 6.93E-111 5.69E-lll 4.66E-lll -
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