ML20030A232

Radiological Consequences of Main Steam Line Failure W/Large Steam Generator Tube Leaks

ML20030A232
Person / Time
Issue date:	12/28/1979
From:	Fontecilla H, Grimes B NRC COMMISSION (OCM), Office of Nuclear Reactor Regulation
To:
References
FOIA-81-313, REF-GTECI-A-03, REF-GTECI-A-04, REF-GTECI-SG, TASK-A-03, TASK-A-04, TASK-A-05, TASK-A-3, TASK-A-4, TASK-A-5, TASK-OR NUDOCS 8001030399
Download: ML20030A232 (14)

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{{#Wiki_filter:. f; .= 5-: Is 3 h .=- f.: t". !=4 i K-RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE FAILURE [ih WITH LARGE STEAM 3ENERATOR TUBE LEAKS ' ?!? 55$ 2 H. M. Fontecilla, B. K. Grimes E.. Environmental Evaluation Branch s. Division of Operating Reactors U. S. Nuclear Regulatory Commission E" t.: t N J Ill. Ii. = BACKGROUND Recent occurrences of steam generator tube degradation have resulted in a number of analytical and experimental efforts to provide assurance of adequate steam generator integrity during normal operation and under p postulated accident conditions. Wile the primary thrust of tnese efforts

ji is to assure that the initial structural and caterials properties are kh maintained and that existing cargins are such that no additional tube-f3-failures need be postulated under loadings imposed by independently

~= initiated transients or accidents, it is appropriate tp perform scoping calculations which postulate such additional tube failures to obtain 5 perspective on the magnitude of the' potential hazard and to cetermine h) the degree of assurance of tube integrity required. The two accident events which result in loadings significantly different from those seen d in normal operation are the loss of coolant accident (which results in 7.? external, or collapse, forces on the steam generator tubes) and the steam g line failure accident (which results in internal, or burst, forces on the t.=9 steam generator tubes). The potential safety significance of steam ((f I5 generator tube failures during a loss of coolant accident is that steam If entering the primary system through failed tubes might. cause a ba~ck-j pressure which would slow the entry of emergency core cooling water into - Ei l the core. The potential safety significance of steam generator tube fail-5? ures is that radioactivity normally retained within the primary coolant i system could be released to the environcent. These relcases include radio-activity in the pri=ary coolant prior to the accident and radioactivity b released from the fuel during the pressure and te=parature transient re- !!E

• sulting from the accident. This analysis considers the steam'line break P

l with postulated additional large primary to secondary system leaks result-h ing from steam generator tube failures during the depressurization of the Esi secondary system. 5.iN

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E.5 2:? bI STANDARD REVIEW PLAN ANALYSIS ML The double ended rupture of a main steam line in a p essurized water FM-reactoia (PWR) is one of the postulated design basis accidents considered s.3: ~ during the licensing process. The analysis of this postulated accident at E!s?. the operating license stage (OL) assumes that, as described in the Appendix 5" to Standard Review Plan 15.1.5, the failure occurs outside the contain=ent EE and upstream of the safety and relief valves (Figure 1). Ttris failure re-Ei suits in the nearly i=:ediate release of the entire contents of one steam iEEE generator to the environment; the blowdown of more than one steam gener-5Z... ator being prevented by the rapid closure of the steam line isolation valves [-] in response to several isolation signals and,' in some cases, reverse flow sM:- check valves. Other postulated failure locations would result in lower [:?j,;, radiological ceasequences because the re' lease of activity would bs rapidly !=.fi.1 terminated by isolation valves or be retained within the containment. The !5s; M:: analysis assumes that the reactor trips and that feedwater ficw to the .affected steam generator is stopped during the first few =inutes after the i-Os F@ steam line fails. The unaffected steam generator (s) are thereupon used to cool down the system to a safe shutdown level. The =ain sources of radio-active releases for this postulated accident are: ~ EM ~ i 1.' ) the activity contained in the failed steam generator at the ti=e ,s[.{ of the accident.(usually less than 5 Ci of Dose Equivalent I-131), E::2 7 = 7_ 2) the activity carried by the steam released through'the unaffecte'd l=E' ' steam generators' safety and relief valves (usually less than 3 Ci

== of Dose Equivalent I-131 for 2-hour releases), and ] . j;..

3) the primary coolant activity assu=ed to leak through the failed steam generator tubes.

At the OL stage it is conservatively E-[:- assumed that the entire primary to secondary leak rate allowed s.== .., by the technical specifications (usually 1. spm) occurs in the sf5h affected steam generator. This operational leak is assu=ed to 55@ continue at a constant rate until the primary system is depres-Ess surized. The activity released through this leak usually amounts E=gy to less than 2 Ci of Dose Equivalent I-131 for a 2-hour,1 spm HE 1eak. E== Er The total activity released, therefore, i'sually amounts to less than EE:.' 10 Curies of Dose Equivalent I-131"and results in a 2-hour dose to the iEE C._ t

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.E.. #. . := ' .AG) ; IEf0 thyroid 'of only a few rem at the nearest exclusion area boundary.' g:5 =:r STEAM LINE BREAK WITH LARGE TUBE LEAKS ~;;;,;;;; 1 5i Recent occurrences of steam generator tube degradation, however, have }~$ raised questions regarding the effect of larger steam generator tube leaks t=s on the' calculation of offsit's doses were these to be postulated to occur, - {] folicwing a steam line failure accident. The NRC staff is currently ana-m lyzing both the manner, in which the normal operating tube le,aks may be ..T.{ EE expected to increase, if at all, and the corresponding response of the 521.:.- plant folicwing the pressure transients during the postulated steam line failure accident. This arlalysis has been performed prior to the completion $Ef of these ongoing studies with the purpose of obtaining a preliminary as-isf3 sessment of the radiological consequences of a postulated steam line failure M d-l j with la'rge t'ube leaks in the,affected steam generator. Future studies may .y also have to consider the possibility of leaks in the unaffected steam generator (s) and blowdown of core than one steam generator as a result of failure of the isolation valves to close. Additional fuel clad failures r~" are not currently postulated to result but this assumption may have to be $.5 re-examined for very large postulated primary,to secondary leak rates. EE@ Once the primary to secondary leak rate is increased above a few spm, the Es.9 EE activity associated with it becomes the main contributor to the offsite doses. This is because the pri=ary system operates with an iodine activity h 2 at least a factor of-10 higher than the secondary coolant. In addition, gcg the primary coolant' iodine activity will show a sharp increase for a few ~~~;ii hours following the transient because of the iodine spike' phenomenon. As a result, the contributions from the secondary system releases may be 15nored }.' = as negligible in the analysis of a postulated steam line failure with large l Z- ~ s steam generator tube leaks.

g E52 For the purpose of this preliminary analysis, the following sequence

{g of events is assumed to occur:

1) - Main steam line failure x=W x

2) Theentireconh.entof,theaffectedsteamGeneratorisdischarged

7 to the environment within a few seconds.

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~ = 8These' estimates are made under the assumption that the primary and d secondary coolant activities at the time of the acci' dent do not exceed I[~) the equilibrium licits specified in the Standard Technical Specifications ZZ for PWR's, i.e., 1.0 pCi/g and 0.1 pCi/g Dose Equivalent I-131, respec-tively. Other' assumptions indicated in the Appendix to Standard Review E53 ""9 Plan 15.1 5, such as an iodine spike factor of 300 and loss of offsite power are also taken into account in consequence calculations made during '],(( ". g f specific plant reviews. EE .=s=.d IE5 = =, $$5. .~ [5Et 8=8 5B

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3) Large steam generator leaks suddenly develop in the tubes of EE:b the affected steam generator.

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.Eyfj 5) Activation of ECCS (in order to maintain pressurizer level) 3i

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6) Rapid cooldown and depressurization of primary system

7) Feedwater flow to failed steam generator is stopped to slow m =d system cooldown.

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8) Within a few minutes primary system pressure decreases to about 400 psi and then slowly decreases to about 300 psi at a

.252 te=perature of about 270*F'(Figures 2 and 3).

Ig.f.j) ~ ~ ~ - 6:H:5. m.mm.t These events.may be expected to happen within 10 to 20 minutes after Ig== km~E the accident, depending en the plant characteristics and the size of the leaks. t=nGE Soce time later, possibly 30 minutes after the accident or longer, [7"" the pricary system may be expected to be cooled down to about 212*F or eEEE lower. After this ti=e only a small, fraction of the pri=ary cociant leak-ing to th3 secondary side will. flash to steam, the rest accumulates at.the *

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.??.5l.? bottom of the failed steam generator. From this time on the radioactive releases are greatly reduced because of the smaller amount of the leaking ej2[

ffi pri=ary coolant that flashes to the environment (only a few percent as co= pared to 100% at the beginning of the accident) and because an increasing

[isi portion of the iodine will re=ain entrapped in the water. gg.5E Therefore, the t:Essi contribution to the doses from these releases has been neglected. @EEE It may also be expected that at this time the operator will have ade - [EEI . quate knowledge of the situation (e.g., conditions in the steam generators, E.EME l primary coolant makeup flow, etc.), particularly in the case of very large ...,;2: j@E@ 1caks, t.o take appropriate actions. It may be expected that the operator will cocpletely depressurize the primary system by opening the pressurizer "~ME . = = ralief valves and by switching to the RHR system for cooling, thereby de-j[jE . pressurizing and cooling at a faster rate than if no action were taken.. Once the primary system is depressurized, the leak will stop. ' Egg = r ewe = During this ti=e the operator may also reopen the feedwater flow to the )(([ fciled steam generator with the purpose of maintaining the. leak under water if possible. El55i This will be accomplished more rapidly for leaks at the bottom E55 of the tubes than for those at the top. f_. ="._ 65_-I aamE E55.5-5:5.k M - z...

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We have performed a series of calculations with the purposes of bound-M ing the radiological consequences for this accident and identifying the gg:; coct important parameters. In all these calculations we have assu=ed that

.

;;r tha different parameters under consideration are independent from each Q

cther. In this manner, it is possible to assume extrece value3 for several Eq pcra:eters, although in reality these conditions may not occur si=ulta EEE. cc:usly. Therefore, the accident described here does not necessarily F=""- r0flect a likely event sequence, but rather a hypothetical accident de-E5if. signed to bound a range of consequences. ~ (('E' 2-We have assumed in the calculations that the primari to secondary IS:! lesk starts at the ti=e of the accident (t=0) and continues at z. constant 1:)ii rete for the length of time (at) necessary for the coeplete cooldown E.5 cnd depressurization of the system. The calculations were performed ar;:c m using the basic characteristics of an. average 1700 Mwth Westinghouse, E!::m.: PWR' with the following assu=ptions: }-{T r.:

1) Primary coolant activity prior to the accident s 1.pCi/g of j.}.}.'{

Dose Equivalent I-131. This is the caximum equilibrium level

=

allowed by the Standard Technical Specifications for PWR's. ~ This is considered conservative because rest plants operate at Q;.

, equilibrium levels at least a factor of 10 lower.

3.E.50 No additional fuel cladding failure: occur as a result of the [.l.3:5I - 2)' accident. This is based on the' expectation that the ECCS will QIj be activated in response to the loss of pressurizer level and 5s5h' that the capacity of the charging pumps is sufficient to cain-tain the core covered at all times during the accident. ........ = Emis

3) The equilibrium I-131 source term to the pri=ary coolant (release 5@.7 rate from fuel to coolant),.13 Ci/ min with 40 gpm cleanup rate

=@E for the case analyzed, is assumed to instantaneously increase by a specified factor at the ti=e of the accident and remain constant g=.=5 5F# ]' for two hours (Figure 4). This factor is referred to as the iodine fd spiking factor (SF). Primary coolant cleanup is assumed to stop I.==E at the time of the accident. Calculations have been perfor=ed for $-3... several values of the spiking factor. A spiking factor of 500 is !:M:T recoc.:: ended in the Standard Review Plan 15.1.5 for licensing cal- !3gg culations as a conservatively high spike, although higher spikes {1=, have been observed. A spiking factor of 100 may be considered to L==- represent a typical spike. It should be noted that these are release rate factors, not factors of increase on the primary _==f coolant iodine concentration. Ib E IM. ' Calculations for other power levels and system configurations have not 5 ?fe b;cn perfor:ed at. this time. [Q$ y;.:. =n...;;. .. =. --5J., - ,h:5 p s:.

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4) The primary coolant has been assumed to leak at a constant rate from the time of the accident (t:0) to some time later (At) when p

the leak stops because of the different events described above . ""s i (Figure 5). The leak thrdus: the tube failures would be a function w of the primary coolant pressure. As the system depressurizes the E55:1 [' ; leak woUld be greatly reduced from the initial value. For example, ?., a leak rate of about 1500 gpm at the outset of the eccident is 3 estimated to be reduced to about 750 gpm in aboat 100 seconds, the

i = =

EM time necessary for the pressure to be reduced from 2300 psia to 500 psia. r.::.:..:. For comparison, we estimate the leak through a' double ended steam ~ [....= generator, tube to be about 300 spe after the first minute of the E.2~ 15 transient, while the largest flow observed through an actual tube E""i failure has been 120 gpm (Foint Beach Unit 1 in February, 1975). 5 Tre..

5) The primary coolant inventory is maintained constant during the accident, i.e,, the charging pumps cake up for all losses.

u. This has th,e net effect of diluting the primary coolant activity i!# concentration as a function of time. =l:....

6). All the activity in the leaking primary coolant is released.to

[llj.:; IEEE the environment. T. :c.. f-...2 CALCULATICN RESULTS The results of our calculations are shown in Figures 6, 7, and 8 as a functior' of leak rate, At, and SF, respectively. For a conservative spike

g

fcctor or 500 the releases are expected not to exceed a few hundred curies cf I-131 for leaks of up to 1000 spm for 40 minutes. This is a conservative L-scquence of events because leak rates of such a high cagnitude should be j{nEcc rapidly recognized by the operator, allowing him to take procpt action to ,..u, 2" TF cool the primary system. In' addition, the higher the lesk rate, the faster th9 system will cool down. On the other hand, lea!? rates of about 150 spa ETim cay be assumed to continue for as long as 90 minutes with comparable re-I." .? [ leases. " Er .== We hav's also considered the less probable event in whien the stean line and nteam generator tube failur'es o'ccur at a ti=e when the plant is, @.fh E555 operating with an iodine coolant activity higher than 1 pCi/gm resulting from previous power level changes.' Assuming an initial iodine coolant [.L._.; cetivity of 10 pCi/g of I-131, we estimate that leak rates of 200 to 300 gpm would not result in r : leases in excess of 100 Ci of I-131. The

m. st:

rosults of these calculations are presented in Figure 9 MM EE E E..E ~ 5 55

9. ?.s-coperation under these conditions is limited to 10% of the yearly' operating time by the Standard Technical Specifications for PWR's.

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. tJa-Me: Based on these results, we conclude that for plants on the order of 1700 Mwth even if one or several steam generator tubes were to fail subse -

====. quent to a postulated ste;m line failure, the releases would be limited t,o IANA no core than a few hundred Ci of I-131 as long as the primary system can be ts= depressurized and the leaks stopped within a reasonnble length of ti=e. In Rff" addition, the maximum releases from this event would only be a few thous,and ~ ~ ~ Ci of I-131 everi for very large leak rates, assuming no adcitional fuel

• = s cladding da= age as a result of the rapid pri=ary system depressurization.

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