ML20211Q541

Proposed Tech Specs Reflecting USQ Associated with Increases in Offsite Dose Consequences Reported in FSAR for SGTR & MSLB Accidents

ML20211Q541
Person / Time
Site:	Callaway
Issue date:	09/08/1999
From:	UNION ELECTRIC CO.
To:
Shared Package
ML20211Q539	List: ML20211Q541 ULNRC-04103, Application for Amend to License NPF-30,reflecting USQ Associated with Increases to Offsite Dose Consequences Reported in FSAR for SGTR & MSLB Accidents
References
NUDOCS 9909150090
Download: ML20211Q541 (38)

Text

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.o 0 ATTACHMENT THREE IMPROVED TECHNICAL SPECIFICATION BASES CHANGES i

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9909150090 990908 PDR ADOCK 05000483 P pm I

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RCS Specific Activity B 3.4.16 s B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.16 RCS Specific Activity BASES . . . . . .

-BACKGROUND The maximum dose to the whole body and the thyroid that an individual at the site boundary can receive for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during an accident is specified I in 10 CFR 100 (Ref.1). The limits on specific activity ensure that the i doses are held to a small fraction of the 10 CFR 100 limits during analyzed transients and accidents.

The RCS specific activity LCO limits the allowable concentration level of radionuclides in the reactor coolant. The LCO limits are established to minimize the offsite radioactivity dose consequences in the event of a steam generator tube rupture (SGTR) accident.

The LCO contains specific activity limits for both DOSE EQUIVALENT l-131 and gross specific activity. The allowable levels are intended to limit the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the site boundary to a small fraction of the 10 CFR 100 dose guideline limits. The limits in the LCO ate standardized, based on parametric evaluations of offsite radioactivity dose consequences for typical site locations.

a-

'The parametric evaluations showed the potential offsite dose levels for a SGTR accident were an appropriately small fraction of the 10 CFR 100 dose guideline limits. Each evaluation assumes a broad range of site l applicable atmospheric dispersion factors in a parametric evaluation. j i,r ped w 4 hen The LCO limits on th specific activity of the reactor coolantbsure inf -k APPLICABLE that SAFETY the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> d es at the site boundary will not exce ed a small ANALYSES fraction of the 10 CFR 00 dose guideline limits following g SGTR accident. The SGTR s ty analysis (Ref. 2) assumes thelhpecific activity of the reactor coolan he LCO limit andKin existing reactor coolant ***#

steam generator (SG) tube leakage rate of 1 gpm. The safety analysis N* j - assumes tnespecific activity of the secondary coolant is imit of l 0.1 pCi/gm DOSE EQUIVALENT l-131 from LCO 3.7.18,"Se ondary Specific Activity."

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. The analysis for the SGTR accident establishes the acceptance limits for RCS specific activity. Reference to this analysis is  ;

used to assess changes to the unit that could affect RCS specific activity, as they relate to the acceptance limits.

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CALLAWAY PLANT B 3.4.16-1 Revision 0

RCS Specific Activity B 3.4.16 I ## #8

,h BASES /'

w APPLICABLE The analysis is per ormed for two cases of reactor colant specific SAFETY activity. One cc&ssumes-operMcectMty et 4.0 Oilu m D O 5 E ANALYSES -EQU!'!^. LENT ! 131 ""th a concurrent large iodine spike that increases (continued) the rate of iodine release into the reactor coolant y a factor of about 500 immediately after the accident. The ccend 000. ssumes the initial reactor coolant iodine activityM S0.0 pCilgm DOSE EOU:V,^ LENT l-101 -

due to a pre-accident iodine pike caused by an RCS transient. In both cases, the noble gas activit in the reactor coolant assumes 1% failed fuel, which closely equals e LCO limit of 100/G pCi/gm for gross specific ac}t fr Ac4*- o P 4o ArgL en +4 n Cue I The analysis also assumes a loss of offsite power at the same time as the reactor trip after an SGTR event. The SGTR causes a reduction in g * # ##"M*^< reactor coolant inventory. The reduction initiates a reactor trip from a low

1. N Tecarts/ pressurizer pressure signal.

S~-Uerirra1la fg g j The loss of offsite power causes the steam dump valves to close to protect the condenser. The rise in pressure in the ruptured SG N8S "ence 3 discharges radioactively contaminated steam to the atmosphere through the SG atmospheric steam dump valves. The unaffected SGs remove l

core decay heat by venting steam to the atmosphere until the cooldown

(-- +he CM /S. S.3 $ra r-/fong The safety analysis shows the radi alogical consequences of an SGTR accident are withir -c em;" fracticfof the Reference 1 dose guideline limits. Operation with lodine specific activity levels greater than the LCO limit is permissible, if the activity levels do not exceed the limits shown in Figure 3.4.16-1, in the applicable specification, for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. l The safety analysis has concurrent and pre-accident iodine spiking 4evehr-

- up to 90.0 pC!!gm DOSE EOU; VALENT i-101. - carer .

The remainder of the above limit permissible iodine levels shown in Figure 3.4.16-1 are acceptable because of the low probability of a SGTR accident occurring during the established 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> time limit. The occurrence of an SGTR accident at these permissible levels could increase the site boundary dose levels, but still be within 10 CFR 100 dose guideline limits. 3 1

1 The limits on RCS specific activity are also used for establishing l standardization in radiation shielding and plant radiation protection practices. ,

RCS specific activity satisfies Criterion 2 of 10CFR50.36(c)(2)(ii). )

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CALLAWAY PLANT B 3.4.16-2 Revision 0

l Sccondary Specific Activity B 3.7.18

% B 3.7 PLANT SYSTEMS Sb] B 3.7.18 Secondary Specific Activity BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Reactor Coolant System (RCS). Under stesdy state conditions, the activity is primarily iodines with relatively short half lives and, thus, indicates current conditions. During transients,1-131 spikes have been observed as well as increased releases of some noble gases.

Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

A limit on secondary coolant specific activity during power operation minimizes releases to the environment because of normal operation, anticipated operational occurrences, and accidents.

This limit is lower than the activity value that might be expected from a 1 gpm tube leak (LCO 3.4.13, "RCS Operational LEAKAGE") of p-imary coolant at the limit of 1.0 pCi/gm (LCO 3.4.16, "RCS Specific Activity").

The steam line failure is assumed to result in the release of the noble gas and iodine activity contained in the steam generator inventory, the feedwater, and the reactor coolant LEAKAGE. Most of the iodine isotopes have short half lives, (i.e., < 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />).

Operating a unit at the allowable secondary coolant specific activity will assure that the potential 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exclusion area boundary (EAB) exposure is limited to a small fraction of the 10 CFR 100 (Ref.1) limits.

r preder +kn APPLICABLE The accident anafysTs of the main steam line break (MSLB), as SAFETY discussed in the FSAR, Chapter 15.1.5 (Ref. 2) assumes.the initial ANALYSES secondary coola,t specific activity to have a radioactive isotope concentration +#b.10 pCi/gm DOSE EQUIVALENT l-131. This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that the radiological consequences of an MSLB do not exceed a small fraction of the unit EAB limits (Ref.1) for whole body and thyroid dose rates.

With the loss of offsite power, the remaining steam generators are available for core decay heat dissipation by venting steam to the atmosphere through the MSSVs and steam generator atmospheric steam (continued)

'CALLAWAY PLANT B 3.7.18-1 Revision 0 i

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,. l ATTACHMENT FOUR PROPOSED FSAR CHANGES l

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CALLAWAT - SP The tables provided in Sections 11.1 and 11.1A document 4 ,

assumptions and input to complete the GALE Code Calculation t .. .

which provided an estimate of effluent releases required for i licensing purposes. Therefore, these tables may not reflect I current plant design and may differ somewhat from actual plant operation.

11.1.2 SHIELDING l

Reactor coolant and secondary coolant source terms used for shielding ~are based on 0.25-percent fuel defects. The source terms and the parameters used to calculate the source terms are i given in Table 11.1-4 and Appendix 11.1A, respectively. Table j 11.1-6 provides the isotopic composition of the contained sources for radioactive waste management systems and for large, g! r; potentially radioactive outside storage tanks.

11.1.3 ACCIDENT ANALYSIS SOURCE TERMS a LOCA and a fuel handling accident, the spec activity use ccident analysis releases is base n operating with 1-perc el defects. Table 1 -5 provides the isotopic composition of -

eactor co based on 1 percent fuel defects. Table 11. - vides the inventory of the contained sources for radioac

• e wa nagement systems and for large, potentially ra active outside orage tanks. -

Sources for the LOCA a* based on TID 14844.

{ Sources for the el handling accident are based on Regulato Guide 1.25.

Cha r 15.0 provides a complete discussion and a listing of source terms for each accident analyzed.

fyr.a derms use) in a ccislsorl- an*lfds are bare on **C"*fhona fretenY in Se c lfon /F 01 anol in Ne Individual an./ysas. Cec lfons hiscutrinb l{f. .

Rev. OL-7 11.1-2 5/94

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TABLE 15.0-7 SINGLE FAILURES ASSUMED IN ACCIDENT ANALYSES

[ Event Descriotion Worst Failure Assumed Feedwater temperature reduction One protection train l l l Excessive feedwater flow One protection train Excessive steam flow (1)

Inadvertent secondary depressurization .

One safety injection train Steam system piping failure One safety injection train .

Steam pressure regulator malfunction (2)

Loss of external load One protection train l Turbine trip One protection train Inadvertent closure of MSIV One protection train j Loss of condenser vacuum One protection train l l Loss of ac power One auxiliary feedwater pump Loss of normal feedwater One auxiliary feedwater pump Feedwater system pipe break One protection train Partial loss of forced reactor coolant flow One protection train l Complete loss of forced reactor coolant flow One protection train RCP locked rotor One protection train RCP shaft break One protection train l RCCA bank withdrawal from suberitical One protection train RCCA bank withdrawal at power One protection train Dropped RCCA, dropped RCCA bank (1)

Statically misaligned RCCA (3)

Single RCCA withdrawal One protection train Inactive RC pump startup One protection train

Flow controller malfunction (2)

Uncontrolled boron dilution Standby charging pump is operating (Modes 1 and 2), one source range i NIS channel (Modes 3-5)

Improper fuel loading (3)

RCCA ejection One protection train Inadvertent ECCS operation at power One protection train - - -

Increase in RCS inventory One pressurizer level channel BWR transients (2) g3 Inadvertent RCS depressurization One protection train W l

Failure of small lines carrying primary (3)

' coolant outside containment  :

SGTR One SG atmosphericaiwfgvalve BWR piping failures (2) J4,,m/ ump Spectrum of LOCA Small breaks One safety injection train Large breaks One safety injection train NOTES: (1) No protection action required (2) Not applicable to Callaway / ,

(3) No transient analysis involved b Rev. OL-10 11/98 1.

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CALLAWAY - SP 3 In case of loss of offsite power, the remaining steam er generators are available for dissipation of core decay heat by

?g'[ venting steam to the atmosphere via the atmospheric relief and safety valves. Venting continues until the reactor coolant temperature and pressure have decreased sufficiently so that the RHR system can be utilized to cool the reactor.

15.1.5.3.1.2 Assumptions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Tables 15.1-3 and 15A-1.

The assumptions used to determine the concentrations of radioactive isotopes within the secondary system for this

(+g) accident are as follows:

a. P' sec dary yst ini 1i Ine tiv y 'i l gg , ass d to e th dose quiv ent O. pCi gm f A k *~

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b. A primary-to-secondary leakage rate of 1 gpm is assumed to exist and is assumed to be in the affected steam generator.

c. The reactor coolant initial iodine activity is l determined by two methods, and both cases ar3 analyzed. These are: fc A/f s g r [$

7 6[5 ac rc ac vit is

()a re 1 Ess inidt al be ed ee ant odi Ival eo 1.0 Ci/ o I- 1w ha iod e sp e th inpeas t r e io ne leas int the r actoy coo nt y ac ro 500. ~-

(2m se 2 -- An ss d re tor colan iod .e a ivi wi a A l r #ds iv ent 60 1/gm f I- 31 a sult

'qf .. eac dent odine pike .

d. The initial reactor coolant concentrations f e j

,, gas correspond to 1-percent failed fuel.

!"$s UF f# e. Partition factors used to determine the secondary system activities are given in Table 15.1-3.

The following specific assumptions and parameters are used to calculate the activity release:

a. Offsite power is lost, resulting in reactor coolant pump coastdown.

Rev. OL-7 15.1-21 5/94

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, , l INSERT A I

The initial secondary side radio-iodine concentrations are assumed to be 10% of the initial Case 1 primary side concentrations.

i' INSERT B The Case 1 initial radio-iodine concentrations are the same as the Case 1 concentrations used for the Steam Generator Tube Rupture accident sequence.

Refer to Table 15.6-4.

INSERT C The Case 2 initial radio-iodine concentrations are the same as the Case 2 conceritrations used for the Steam Generator Tube Rupture accident sequence.

Refer to Table 15.6-4.

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CALLAWAY - SP

b. No condenser air removal system release and no normal  ;

operating steam generator blowdown is assumed to occur jeg j during the course of the accident. t c, Eight hours after the occurrence of the accident, the residual heat-removal system (RHRS) starts operation to cool down the plant.

d. After the accident, the primary-to-secondary leakage continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time the reactor coolant system is depressurized.

e. The affected steam generator (steam generator connected to the broken steamline) is allowed to blow down completely.

((}.a.

f. Steam release to the atmosphere and the associated activity release from the safety and relief valves and the broken steamline is terminated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident, when the RERS is activated to complete cooldown.

g. The amount of noble gas activity released is equal to j the amount present in the reactor coolant, which leaks l to the secondary during the accident. The amount of  !

iodine activity released is based on the activity present in the secondary system and the amount of ~

leaked reactor coolant which is entrained in the steam that is discharged to the environment via the safety and relief valves and the broken steardine. Partition factors used for the unaffected steam generators after the accident occurs are given in Table 15.1-3. An iodine partition factor of 1 is used !or the affected steam generator.

h. She activity released from the broken steamline and the safety and relief valves during the 8-hour duration of the accident is immediately vented to the atmosphere. . . .

15.1.S.3.1.3 Mathematical Models Used in the Analysis ,

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"M Mathematical models used in the analysis are described in the following sections:

a. The mathematical models used to analyze released during the course of the accidentthe areactivity' Amf,2 an dercribed in .'.p;:ndix 12.- 4Ae .e,rrump /imr //f./e/ , joy, ,

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum.

Rev. OL-1 l 15.1-22 6/87 j

4 CALLAWAY 'dP

c. The thyroid inhalation dose end total-body gamma 62'

'?W immersion doses to a receptor A*. the exclusion area boundary and outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A.

15.1.5.3.1.4 Identification of Leakage Pathwaya and Resultant Leakage Activity For evaluating the radiological consequences due to a postulated MSLB, the activity released from the affected steam generator (steam generator connected to the broken steamline) is released directly to the environment. The unaffected steas j{,i,

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generators are assumed to continually discharge steam and entrained activity via the safety and relief valves up to the time initiation of the RHRS can be accomplished.

Since the activity is released directly to the environment with no credit for plateout or retention, the results of the analysis are based on the most direct leakage pathway available. Therefore, the resultant radiological consequences represent the most conservative estimate of the potential integrated dose due to the postulated MSLB.

15.1.5.3.2 Identification of Uncertainties and Conservatisms i in the Analysis gfgg7- B l

a. React coo nt ac ivities re ba d on e Tech cal I S ific ion 1 t of 1 pCi/ I-13 dose quiva nt wit extrem y lar iodi spike alues pers . ting f the e ire du ation the ciden ,

r lting equiv ent co entra ens ma time '

reater an the eactor colant ctivi es ba d on 0.12 p cent fa' ed fu assoc' ted wi no (quera ng con tions -

b. A 1-gpm steam generator primary-to-secondary leakage is assumed, which is significantly greater than that anticipated during normal operation. Furthermore,-it was conservatively assumed that all leakage is to the

_ p., affected steam generator only.

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E. !? ' c. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that the assumed meteorological conditions would be present during the course of the accident for any extended period of time. Therefore, the radiological consequences evaluated, based on the me' eorological conditions assumed, are conservative.

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Rev. OL-1 15.1-23 6/87

CALLAWAY - SP

b. No condenser air removal system release and no normal operating steam generator blowdown is assumed to occur during the course of the accident. f;-

c. Eight hours after the occurrence of the accident, the residual heat-removal system (RHRS) starts operation to cool down the plant.

d. After the accident, the primary-to-secondary leakage continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, at which time the reactor coolant system is depressurized.

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e. The affected steam generator (steam generator .

connected to the broken steamline) is allowed to blow down completely.

f. Steam release to the atmosphere and the associated activity re] ease from the safety and relief valves and the broken rateamline is terminated 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident, when the RHRS is activated to complete cooldown.

g. The amount of noble gas activity released is equal to the amount present in the reactor coolant, which leaks to the secondary during the accident. The amount of iodine activity released is based on the activity present in the seconda.y system and the amount of ~

leaked reactor coolanc which is entrained in the steam that is discharged to the environment via the safety and relief valves and the broken steamline. Partition factors used for the unaffected steam generators after the accident occurs are given in Table 15.1-3. An iodine partition factor of 1 is used for the affected steam generator.

b. The activity released from the broken steamline and I the safety and relief valves during the 8-hour duration of the accident is immediately vented to the i atmosphere. _ _ .

15.1.5.3.1.3 Mathematical Models Used in the Analysis m Mathematical models used in the analysis are described in the Gh".

"?

following sections:

a. The mathematical models used to analyze the activity released during the course of the accident are Saf,2 an dccaribcd in .".pycndix 1:Av-lke urump lions lir/ed ajoye .

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurement programs described in Section 2.3 of the Site Addendum.

Rev. OL-1 15.1-22 6/87

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CALLAWAY - SP Q c. The thyroid inhalation dose and total-body gamma g" 1 immersion doses.to a receptor at the exclusion area boundary and outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A.

15.1.5.3.1.4 Identification of Leakage Pathways and Resultant Leakage Activity For evaluating the radiological consequences due to a ,

postulated MSLB, the activity released from the affected steam generator (steam generator connected to the broken steamline) is released directly to the environment. The unaffected steam

%- generators are assumed to continually discharge steam and entrained activity via the safety and relief valves up to the time initiation of the RHRS can be accomplished. j I

Since the activity is released directly to the environment with i no credit for plateout or retention, the results of the analysis are based on the most direct leakage pathway available. Therefore, the resultant radiological consequences represent the most conservative estimate of the potential integrated dose due to the postulated MSLB.

15.1.5.3.2 Identification of Uncertainties and Conservatisms in the Analysis ggggg D a.

[eact coo nt ac 'vities re ba d-on e Tech cal )

IS ific- ion 1 t of 1 pCi/ I-13 dose quiva nt wit extrem y lar iodi spike alues )

pers ting f the e ire du ation the ciden ,

r lting equiv ent co entra ons ma time reater an the eactor colant ctivi es ba d on 0 cent fa' ed fue assoc ted wi no h.12p era ing con tions. -

b. A 1-gpm steam generator primary-to-secondary leakage is assumed, which is significantly greater than that anticipated during normal operation. Furthermore;-it was conservatively assumed that all leakage is to the j affected steam generator only.

c. The meteorological conditions which may be present at the site during the course of the accident are uncertain. However, it is highly unlikely that the assumed meteorological conditions would be present during the course of the accident for any extended ,

period of time. Therefore, the radiological i consequences evaluated, based on the meteorological conditions assumed, are conservative.

Rev. OL-1 15.1-23 6/87

INSERT D Reactor coolant activities are based on an initial radio-iodine spectrum that would conservatively bound those found in either open or tight type fuel defects.

Tight fuel defects tend to produce limiting results for thyroid dose, while open fuel defects tend to produce limiting results for whole body dose. The assumed .

concentrations of longer-lived isotopes represent the values that would be I reached in the presence of tight fuel defects. The assumed concentrations of shorter-lived isotopes represent the values that would be reached in the presence of open fuel defects. Since the assumed iodine spectrum represents bounding values for different types of fuel defects, the initial radio-iodine inventory would exceed the Technical Specification limit of 1.0 pCi/gm.

Additionally, large spiking factors are assumed in the analycis.

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CALLAWAY - SP TABLE 15.1-3 r,

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e., . PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK I. Source Data:

a. Core power level, Mwt 3636

b. Steam generator tube leakage,.gpm 1

c. Reactor coolant initial iodine activity: SA/f6g7~ E l

1) Case 1 FDo e va tc 1.0 Ci/ j g '

I 31 th ass ed io ne ike at cre es er e of odin rel ase int the acto coo nt y acto of 50

2) Case 2 r%3 ss dp -acc' ent od e i ike ic as sul d t do equ alen of '/

I-1

d. Reactor coolant initial noble gas activity: M# @

1) Case 1 Based on 1-percent failed fuel as provided in Table 11.1-5

2) Case 2 Based on 1-percent failed fuel as provided in Table 11.1-5

e. Secondary system initial re : equif 1:nt of 0.- rC1/gm iodine activity cf ! 1 1-/0% sf Cu./ ,f u l

Ude oeNvodu

f. Iodine partition factors '/

l

1) Affected steam generator 1.0 l

2) Unaffected steam generator 0.01 1

Tpij g. Reactor coolant mass, lbs 5.30E+5

h. Steam generator mass

1) Affected steam generator, lbs 1.67E+5

2) Each unaffected steam generator, lbs 1.04E+5 II. Atmospheric Dispersion Factors See Table 15A-2 r

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Rev. CL-7 5/94

CALLAWAY - SP i

TABLE 15.1-3 l 4

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PARAMETERS USED IN EVALUATING THE RADIOLOGICAL  !

CONSEQUENCES OF A MAIN STEAM LINE BREAK  !

I. Source Data:

a. Core power level, Mwt 3636

b. Steam generator tube leakage,.gpm 1

c. Reactor coolant initial iodine activity: ;C7df) . ~ ~ di

{

1) Case 1 (Do e va tc 1.0 Ci/ j yy I 31 th- ass ed ij ga <

o ne ike at cre es er e of odin rel ase int the acto coo nt y (

acto of 50

2) Case 2 -1 01 ss p -ace' ent od el ike ic as sul d t )

do equ alen of pC'/

I-1

d. Reactor coolant initial noble gas activity: -2WJer F

1) Case 1 Based on 1-percent failed fuel as provided in Table 11.1-5

2) Case 2 Based on 1-percent failed fuel j as provided in Table 11.1-5  !

e. Secondary system initial rcce equivalent of 0.1 rC1/um j iodine activity cf I 131- /0% sf Can / rru l j

Tide ocNv

f. Iodine partition factors

1) Affected steam generator 1.0

2) Unaffected steam generator 0.01 p g. Reactor coolant mass, lbs 5.30E+5

h. Steam generator mass 1

1) Affected steam generator, lbs 1.67E+5

2) Each unaffected steam generator, lbs 1.04E+5 II. Atmospheric Dispersion Factors See Table 15A-2

. . r; ,

Rev. OL-7 5/94

, o INSERT E The MSLB Case 1 initial radio-iodine inventory is the same as the Case 1 SGTR initial radio-iodine inventory. Refer to Table 15.6-4.

INSERT F 1

The MSLB Case 2 initial radio-iodine inventory is the same as the Case 2 SGTR I initial radio-iodine inventory. Refer to Table 15.6-4.

l

\

l l

l l

. . l CALLAWAY - SP i

TABLE 15.1-4 RADIOLOGICAL CONSEQUENCES OF A MAIN STEAM LINE BREAK Doses (rem)

CASE 1, -1.0 pCi/F I-121 acc//,,r/- fn//7Qs/

-eT_'4 =' er_t =/I rpike /,//n, fp7p, Exclusion area boundary

( 0-2. hr)

Thyroid -ES 53 Whole body 5.222-3 /, // E-/ ' i.4 Low-population zone outer boundary (duration)

Thyroid M 4.f4 Whole body I.102 P.4 E-2 CASE 2,-60 pCi/p I 12'-  ;

eTli':31C=t fre-A cci en)- fo/ina ry;ke ..

Exclusion area boundary (0-2 hr)

Thyroid 3.47 i Whole body '

. 2 2 /. f4 E-2  !

Low population zone outer boundary (duration) i Thyroid 1.42 Whole body 1.in2- D.13 E-3 l4 0)2,h E

t Rev. OL-2 6/88

E ',

• O 'C CALLAWAY - SP ,

l auxiliary building, and the environmental conditions at the time. Each of these uncertainties is treated by taking worst-case or extremely conservative assumptions. ,

The extent of coolant contamination assumed greatly exceeds the levels expected in practice. The rupture is postulated in a l

seismic Category I, ASME Section-III, Class 2 piping system.

It is assumed that the leak goes undetected for 30 minutes. It is expected that considerable holdup and filtration occurs in the auxiliary building, but no credit is assumed.

The purpose of all these conservatisms is to place an upper .

bound on doses.

15.6.2.1.3 Conclusions 15.6 2.1.3.1 Filter Loadings No filter is credited with the collection of radionuclides in '

this accident analysis. The buildup on these filters (auxiliary building and control building charcoal filters) that may be expected due to the adsorption of some of the iodine is  !

very small compared with the design capacity of these filters.

15.6.2.1.3.2 Dose to Receptor at the Exclusion Area Boundary and Low-Population Zone Outer Boundary The radiological consequences resulting from the occurrence of a postulated letdow:1 line rupture have been conservatively analyzed, -

. using assumptions end models described in previous sections.

The thyroid inhalation total-body immersion doses have been analyzed for the 0-2 hour dose at the exclusion area boundary and for the duration of the accident at the low-population zone outer boundary. The results are listed in Table 15.6-3. The resultant doses are well'within the guideline values of 10 CFR 100.

15.6.3' STEAM GENERATOR TUBE FAILURE 15.6.3.1 Identification of Causes and Accident Descrintion The letters listed under Reference 3 discuss the reanalysis of the SGTR accident The licensing basis SGTR accident discussed .44Bt in this section represents an update to the original SNUPPS hk$$l~~

generic analysis that assumes the failure of a steam generator atmospheric r-li-f valc; ';l;; r:f:rr:d to as th; atrerpheric-steam dump valvedp in the open position in order to maximize offsite doses. This update reflects current plant design, '

operation, and analysis parameters (e .g. , VANTAGE 5/ VANTAGE +

fuel, uprated power, 15% equivalent steam generator tube plugging), i The accident examined is the complete severance of a single steam generator tube. This event is considered an ANS t Condition IV event, a' limiting fault (see Section 15.0.1). The I

accident is assumed to take place at power with the reactor coolant contaminated with fission products corresponding to Rev. OL-9 15.6-6 5/97

. )

, c l

CALLAWAY - SP-

'p# continuous operation with a limited amount of defective fuel hsg$

j - rods. The accident leads to an increase in the contamination l of the_ secondary system due to the leakage of radioactive coolant from the RCS. In the event of a coincident loss of offsite power or failure of the steam dump system, discharge of activity to the atmosphere takes place via the steam generator safety and/or power-operated lief valves. I a h morp h e rIc N ea m dum f In view of the fact that the steam generator tube matetial is l Inconel-600 and is a highly ductile material, it is considered that the assumption of a complete severance is somowhat conservative.  ;

1 The more probable mode of tube failure would be one or more minor

--s leaks of undetermined origin. Activity in the steam and power con-version system is subject to continual surveillance, and an accumu-i!{$$ lation of minor leaks which exceed the limits established in the Technical Specifications is not permitted during plant operation. l In order to select the reference worst case, a spectrum of l

SGTR events was analyzed. The letters of Reference 3 provide a detailed description of the selection process.

l

' r- Arb Major concerns associated with n steam generator tube rupture

( SGTF.) are: (1) the potential fc r overfill of the faulted },"" ggf l steam generator with water enter ing the main steam line ["*7 resulting in water relief throuc h an atmospheric :licf valve

_(AP"P and (2) failure of an-ARVJ'in the open position leading to continued release of steam generator fluid and contained radioactivity. To examine these concerns, the SGTR Scoping Code (discussed in Appendix B of the SNUPPS report attached to SLNRC 86 see Reference 3), in conjunction with other analyses, was used to evaluate the sensitivity of SGTR events to a number of parameters. Parameters investigated were:

single active failures; availability of offsite power; location of tube rupture; operator action times; power level; and iodine spiking. The analysis presented in this section was chosen, based upon those investigations, as providing the worst case dose.

~

/>J The recovery sequence for a SGTR is discussed in Section 15.6.3.2.

%w'$

The operator is expected to determine that a SGTR has occurred i and to identify and isolate the affected steam generator on a restricted time scale to minimize contamination of the secondary system and ensure termination of radioactive release to the atmosphere from the affected generator. The recovery procedure l can be carried out on a time scale which ensures that break flow to the secondary system is terminated before water level in the affected steam generator rises into the main steam line. Sufficient indications,. controls, alarms, and procedures are provided to enable the operator to carry out these functions satisfactorily.

Consideration of the indications provided at the control board,

'- together with the magnitude of the break flow, leads to the con-clusion that the accident diagnostics and isolation procedure can Rev. OL-5 15.6-7 6/91

F- _,

• s l:

l CALLAWAY - SP ba : completed such that pressure equalization between the 1 n..

primary and secondary can eventually be achieved and break ish '

flow terminated within 67.3 minutes of accident initiation. {

l Operator actions in response to an SGTR are assumed to follow plant-specific emergency procedures, which are based on procedure E-3 (SGTR response) and related procedures of the generic emergency response guidelines (ERGS) for Westinghouse plants.

The timing of operator actions utilized in the tube rupture analysis presented in this section has been estimated using data from the following sources: (1) plant simulator exercises; (2) SGTR events at the Ginna, Prairie Island, and Myg North Anna plants; (3) draft standard ANS 58.8, Revision 2; and (4) Callaway operating experience in closing an atmospheric relier manual block valve. Heaviest weight has been placed on th simulator and experience data because it reflects what platt operators have done using plant-specific procedures. The etters of Reference 3 provide more detail on the timing of op atpr act,lons.

5+een, gum Assuming normal operation of,0the various plant control systems, the following sequence of events is initiated by the i tube rupture analyzed herein:

a. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the sec-i ondary side, there is a steam flow /feedwater flow mismatch before trip as feedwater flow to the affected I steam generator is reduced due to the additional break l flow which is now being supplied to that generator. l

b. Decrease.in pressurizer pressure (Figure 15.6-3a) due I to continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure. Resultant plant -

un \

n;,:-

'h l

l Rev. OL-5 15.6-7a 6/91

CALLAWAY - SP cooldown (Figures 15.6-3b and 15.6-3c) following o% reactor trip leads to a rapid reduction in pressurizer ts{h level (Figure 15.6-3n), and the safety injection signal, initiated by low pressurizer pressure, is assumed to occur coincident with reactor trip. The safety injection signal automatically terminates normal feedwater supply and initiates auxiliary feedwater addition.

I

c. The steam generator blowdown liquid monitor and/or the condenser air discharge radiation monitor will alarm, 4 indicating a sharp increase in radioactivity in the l secondary system, and will automatically terminate g steam generator blowdown if not already isolated by the 44j*g SGBSIS (AFAS) signal (see Figures 10.4-8 Sheet 1 and 7.3-1 Sheet 2).

con / enter

d. The reactor trip automatically tri s the turbine and, if offsite power is available, the steam dump valves open, permitting steam dump to the condenser. In the I event of a coincident loss of offsite power, as assumed in the transient presented in this section, the steam dump valves automatically close to protect the condenser. The steam generator pressure (Figure 15.6-3a) rapidly increases, resulting in steam A [,[_ discharge to the atmosphere through the steam generator atmosphericWrcli;f valves (Figures 15.6-3g and 15.6-3h). In Figures 15. 6 - 3 d and 15. 6 -3 e , the steam flow is presented as a function of time. The flow is constant initially until reactor trip, followed by turbine trip, which results in a large decrease in flow, but a rapid increase in steam pressure to the atmospheric / relief3 val e setpoint.

T4 tom umf9

e. Following reactor trip, the continued action of auxiliary feedwater supply and borated safety injection flow (supplied from the refueling water storage tank) provide a heat sink which absorbs the decay heat. j l

f. Safety injection flow results in increasing the l .gp pressurizer water volume (Figure 15.6-3n); the rate of I

d' -

which depends upon the amount of operating auxiliary equipment. gg l

g. The atmospheric reli;fVvalve --{??H}- f or the -fulled rw/On/

steam generator (SG) is assumed to fail open and steam release continues for 20 minutes until the.ARV- block l valve is manually closed (Figure 15.6-3g). Du ing this time, pressure falls in all SGs and RCS tempe ature drops in res nse to the steam release.

g

h. Once the .ARVVis isolated, controlled cooldown is initiated and continues until RCS temperature is reduced to 50*F less than the ruptured SG saturation temperature.

Rev. OL-7 15.6-8 5/94 1

1 L

a

. o CALLAWAY - SP

i. Primary depressurization is then performed until primary and secondary pressures equalize. This terminates break flow (Figure 15.6-3i). Safety t injection flow is terminated 3 minutes after RCS.

depressurization. However, in the interim, safety injection flow repressurizes the primary side.

j. Five minutes af ter safety injection termination, it is assumed that the operators minimize the primary-secondary pressure difference by opening a pressurizer PORV (Figures 15'.6-3a and 15.6-3m). Any primary side pressure rise after this second depressurization is moderate and a function of decay heat.

.15.6.3.2 Analysis of Effects and Consecuences k Method of Analysis Mass and energy balance calculations are performed using RETRAN (Section 15.0.11.8) to determine primary-to-secondary mass release and to determine the amount of steam vented from each of the steam generators from the occurrence of the tube rupture until after the second primary-secondary pressure equalization.

RETRAN provides time-dependent values of RCS mass, break flow, flashed fraction, steam generator liquid mass, and steam generator atmospheric rclicf valve flow for the calculation of radiological consequences. onservatively high values of break -

flow rate and flashed fract on are assumed for the first hour of the transient to maximiz radiological consequences.

Supplementary mass and ene balance calculations, with conservative assumptions, re performed for the period from  !

pressure equalization unt 1 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident, beyond 1 the time of RHR initiati n.

Neam dump 15 57 rge Rev. OL-5 15.6-8a 6/91

r- e a s CALLAWAY - SP 1

i In estimating the mass transfer from the RCS through the broken tube, the following assumptions are made:

l l

'%d$f a. Reactor trip and safety injection occur coincidentally )

as a result of low pressurizer pressure.

Overtemperature AT trip is not considered. This allows more break flow. Loss of offsite power occurs at i reactor trip. .

1 i

b. The tube rupture is a double-ended guillotine break of a single hot lec tube at the tube sheet of the steam generator. This break location maximizes the flashed fraction of the RCS break flow.

N,f8)h( c. As listed on Table 15.0-4, the low pressurizer pressure

?S safety analysis limit (SAL) for reactor trip is 1845 psig. This reactor trip SAL is lower than the actual (

setpoint of 1885 psig, which thereby delays the trip and results in increased break flow. Safety injection is assumed concurrent with reactor trip which decreases the time for initiation of safety injection, again resulting in increased break flow. Safety injection occurs 15 seconds after the SI signal. The actual SI setpoint is 1849 psig with a lower SAL in Table 15.0-4. This l minimum expected delay results in an early rise in RCS pressure due to SI and results in increased break flow.

d. Break flow is characterized by resistance-limited flow.

An additional 5% uncertainty is added to the flow.

< e. The assumption of a loss of offsite power at reactor trip prevents steam dump to the condenser and steam is discharged to the atmosphere via the ?.RY:.- With the condenser unavailable for retention of any leaked radioactivity, offsite doses are maximiz

. Asbr.

f. Pressurizer heaters and spray are not modelled.

I

g. Prior to reactor trip, the normal feedwater matches the  !

t steam flow in the intact steam generators. For the rupl ure/ l

_ faultcd steam generator, the total feed flow (including C$$$$ the break flow) matches the steam flow. The feedwater

1#D isolation signal occurs 2.3 seconds after reactor trip I

and the feedwater isolation valves stroke closed within 2.0 seconds. These are the minimum expected delay and stroke time, respectively, which tend to decrease heat l removal from the RCS resulting in higher RCS temperatures and pressures. This results in maximum flashed fraction and break flow.

h. The initial steam generator liquid level is 45% of the narrow range span. This is the minimum expected I

Rev. OL-8 15.6-9 11/95

CALLAWAY - SP level, minimizing the amount of secondary inventory available for decay heat removal. This increases the flashed fraction (the amount of leaked reactor coolant 40 that is vaporized on the secondary side). Auxiliary ;4 feedwater (AFW) flow is maintained to achieve a narrow l range level of at least 45% in all steam generators. l AFW is initiated 60 seconds after reactor trip and {

attains a flow rate of 250 gpm to all steam I generators. This maximum expected delay for AFW I initiation maximizes break flow and maintains high RCS l temperatures. This minimum expected AFW flow to the ruglurs/ fault;dsteamgeneratorresultsindecreasedRCSheat removal, maintains high RCS temperatures, and thereby maximizes the flashed fraction of leaked reactor Coolant. / :D p.u uw ASb &W

i. The _r"lted steam generators 's -ARV Vis set at 1184.7 psia. This is 4% higher than the nominal setpoint which delays the release of pressure from the-faultad-ruhrs/steamgenerator, resulting in increased valve Afb discharge flow and integrated break flow. The daVVon the-fruited steam generator fails open for 20 minutes, beginning o initial demand, shortly after reactor trip. Thi is the single failure that maximizes offsite d es.

rup lweel

j. The initial steam generator pressure is 939 psia, the minimum expected pressure associated with 15% tube -

plugging. This increases the leaked reactor coolant.

k. The decay heat multiplier is 1.2 (based on 3636 MWt).

This maximizes the heat to be transferred which increases break flow.

1. The narrow range level in all steam generators must be greater than 10% and the ruptured steam generator pressure must be greater than 615 psig prior to initiating RCS cooldown.

m. RCS depressurization is assumed to begin 3 minutes after completion of cooldown. When the ruptured steam generator pressure is higher than the RCS pressure, g@$fD the pressurizer PORVs are closed. ' ~

n. Safety injection is terminated 3 minutes after completion of RCS depressurization.

Other initial conditions, given in Table 15.0-2, are chosen to maximize RCS temperatures, decay heat, flashed fraction of RCS leakage, and break flow, thereby maximizing radioactivity transfer to the secondary and, consequently, offsite doses.

The above assumptions, suitably conservative for the licensing basis tube rupture, are made to maximize offsite doses.

Rev. OL-5 15.6-9a 6/91

CALLAWAY - SP i

_ VMflure *^ Inhek~

[ "' Prior to reactor trip, steam is dumped to the condenser from

'Qd((f - both the feuited ;..d uvufaulted steam generators. After the condenser _is lost, following assumed loss of offsite power at reactor trip, steam from all steam generators is released to the atmosphere.

ruphre) e Th** Y"*f fn-},d Following isolation of the'/frulund steam generator, (it is assumed that atmospheric ccliefVfrom the *^"fruitcdvsteam generators is used to reduce the RCS temperature to 50 F below thejfruitei steam generator saturation temperature. From 2 to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, steam is. assumed to be relieved from the : r fruited-/n[ne1L ste un generators to reduce the RCS temperature and pressure to RH 3 conditions. The faultcd steam generator is depressurized "qq))

20 to the RHRS cut-in pressure u ing the emergency recovery pr cedures. After 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, rther plant cooldown is carried on with the RHRS. The O to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 2 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> steam re eases from the intact ste m generators required to remove d ay heat, metal heat, rea tor coolant pump heat, and stored f uid Vufh ure) I r u 4y w -e )

hkG3 dji!W l

4 a

Rev. OL-5 15.6-9b 6/91

E , .

l CALLAWAY.- SP cnergy in the RCS and' steam generators are determined based on gi, these assumptions. 4 Kov Recovery Sequence The recovery sequence to be followed consists of the following major operator actions:

ruflMre V

a-.

Identification of the {:ulted steam. generator}-

raj+ar-ed *

b. Isolation of the fcultcd-steam generator including j closure of the manual block valve i A \

c. Assuring subcooling of the RCS fluid to approximately <

50 F below saturation at the f uli Q stea generator l

j pressurej ruf fure

d. Controlled depressurization of the RCS to a value equal to the f aulty+stea generator pressurej rur w j i

e. Subsequent termination of safety injection flowj and f

f. Further cooldown and depressurization of the RCS to conditions suitable for RHR initiation, Rasults _..

In Table 15.6-1, the sequence of events is presented. These ovents include postulated operator action times. Loss of offsite power is assumed to occur at reactor trip lWre

(~ ref The previously discussed assumptions lead tc an estimate of 450,000 pounds for the total amount of react or coolant trans-l

' forred to the secondary side of the fruit:Wsteam generator as J a result of a tube rupture accident. The steam releases to the condenser and atmosphere from both the 4-"1"A ="A //r/med steam generators are given in Table 15.6-4. mf ulted ruhlrt/ 8 The following is a list of figures of pertinent time dependent .o parameters:

Number Title I 15.6-3a Pressurizer and Steam Generator (Faulted and

. Intact Generators) Pressure Transients for Steam Generator Tube Rupture-Event 15.6-3b Reactor Coolant System Temperature (Faulted Loop)

Transient for Steam Generator Tube Rupture Event 15.6-3c Reactor Coolant System Temperature (Intact Loops)

Transient for Steam Generator Tube Rupture Event 15.6-3d Steam Flow Rate (Intact Generators) Transient for -

Steam Generator Tube Rupture Event Rev. OL-5 15.6-10 6/91

• ',s CALLhWAY - SP 15.6.3.3' Radiological Consequences i 15.6'.3.3.1 Method of Analysis 15.6.3.3.1.1 Physical podel V"fj '"j .

r-~ S11am umf The evaluation of the radiological cons ences due to a pos-tulated. steam genet ator tube rupture (SC R) with a stuck open l otmospheric relief @ valve on the-feult:dPsteam generator '

ecsumes a complete severance of a single steam generator tube while the reactor is operating at full rated power and a coincident 11oss of'offsite power. Occurrence of the accident loads to an. increase.in contamination of the secondary system due to reactor coolant leakage through the tube break. A [' ; l reactor trip occurs automatically, as a result of low pressurizer pressure. The reactor trip will automatically trip the turbine.

j- ton /eusr The resulting sharp irferease in radioactivity in the secondary system will:be detect 6d'by radiation monitors (refer to section 11.5), which will aut>matically terminate steam generator blowdown. The assum coincident loss of offsite power will ccuse closure of the team dump valves to protect the f ggQ,m dum/ o condenser. The steam generator pressure will then in ease rcpidly, resulting in steam discharge as well as acti ity re- .,s lease through the steam generator atmospheric-relief valves. 'l Venting from the affected steam generator, i.e., the steam '

generator which experiences the tube rupture, will continue until the manual block valve is closed, isolating the stuck open atmospheric relief valve on the faulted steam generator.

At this time, the affec.ed steam generator i effectively isolated. The remaini g unaffected steam g erators remove l core decay heat by ve ing steam through th atmospheric l

-reli^f,va4ves f until t e contrqlled cooldow is termipated. I Atam gmyo S4**m dumf Vafk/

The analysis of the radiological consequences of an SGTR considers the most severe release of secondary activity, as --

vall as reactor activity leaked from the tube break. The cy j

inventory of iodine and noble gas fission product activity $pg)s l available for release to the environment depends on the Tr3 l

1 i Rev. OL-5 L 15.6-11 6/91

a CALLAWAY - SP

,-[?5 primary-to-secondary break flow and coolant leakage rates, the U.(h'y percentage of defective fuel in the core, flashed fraction of reactor coolant, and the mass of steam discharged to the environment. Conservative assumptions were made for all these parameters.

15.6.3.3.1.2 Assumptions and Conditions The major assumptions and parameters assumed in the analysis are itemized in Tables 15.6-4 and 15A-1 and are summarized below. l The assumptions used to determine the concentrations of isotopes jffh)' in the reactor coolant and secondary systems prior to the l accident are as follows:

a. The assumed reactor coolant iodine activity is deter-mined for the following two cases:

{.7 NTEM G Case 1 - 6n 'niti 1 re tor olan iodi act ity e al the dose quiva ent o 1.0 if ,

f I- 31 wi h an odine spike that 'ncre ses l the scap rate rom e fue into the I co lant a f ctor 500 mmedi tely aft e acc' dent. This incre ed es ape ate s j ssume for he du ation f the acci ent 1 Case 2 - n itial eact r coo ant io ne a tiv y e al to e do e equ alent f 60 0 p /

o I-131 due a pr -accid t io ine pi e j l aused y RC trans'ents pr or t th SG .f

b. The noble gas activity in the reactor o t a ed on 1-percent failed fuel, as provided in Table 11.1-5.

c. The initial secondary : alar.t activity it beced en

- th : dos; ~uivilc.d. cf 0.1 pCi 'um J I-131. .r//e n/4. ,,//n, f

concerdnf,oog are ufam,) f '), ,,.

,. ggh The following assumptions jgy ,p gg,parameters and pny_y,j gw, are j ,d to'calculatemory use r,*/e concen/nf*rru.

Jh.g. , the activity released and the offsite doses following an SGTR:

%W f, rNfERT :2~

a. fUsitot 1 a unt f digcha eo rea or pool t o' ond t s oun s.

ysstem[hroghteru urfis 50, l

b. 'I t i assum d tha 11 pe cent of th rea or c olan tha flows to th affe ed a d int ct s am g nera or f shes stea and a imm diate y re ase to t e viron ent w le th stea gene ator atmos heri

, relief valve arc o en. o cre it h s bee tak.n for j acrub ing.

L rNTERT 3~

c. A 1-gpm primary-to-secondary leak is assumed to occur l to the unaffected steam generators.

Rev. OL-5 15.6-12 6/91

F, --

INSERT G The initial reactor coolant iodine activity corresponds to an isotope mixture that bounds Technical Specification allowable conditions for both tight and open fuel defects. The initial isotopic mix is based on the relative concentrations from Table 11.1-5. The concentrations are then changed to achieve a Dose Equivalent 1-131 (DEI) of 1.0 Cl/gm, while maintaining the isotopic ratios from Table 11.1-5. This provides conservative values for the longer lived iodines which contribute the majority of the calculated thyroid dose. The initial

- inventories of the shorter lived lodine isotopes, which can provide a significant  :

contribution to the calculated whole body dose, are then increased to conservatively bound isotopic mixes which may occur in the presence of open fuel defects. Case 1 then includes an accident initiated, spiked release rate that increases by a factor of 500 during the accident sequence.

INSERT H  ;

The initial reactor coolant iodine activity corresponds to an assumed pre-accident iodine spike which results in concentrations that are a factor of 60 higher than those used in Case 1.

1 I

INSERT I Break flow to the ruptured steam generator is conservatively assigned values that bound calculated break fiow rate values. The assumed values bound the break flow rates calculated by the RETRAN code. Break flow rate values are l

discussed in Table 15.6-4.

INSERT J The fraction of reactor coolant that flashes to steam after reaching the se condary i side, as assumed in the accident analysis, varies over time. Key events which l trigger changes in the assumed flashed fraction are reactor trip and closure of l the manual block valve to isolate the failed open SG atmospheric steam dump valve. Flashed fraction values assumed in the radiological analysis are described in Table 15.6-4. i i

j l 1 1

ro

• CALLAWAY - SP

d. All noble gas activity in the reactor coolant which g% I is transported to the secondary system via the tube va;.  !

rupture and the primary-to-secondary leakage is l assumed to be immediately released to the environment. l t

1

e. At 67.3 minutes after the accident, it is assumed l that the RCS and steam generator pressures are equalized and below the steam generator atmospheric relief valve set pressure. Break flow to the faulted steam generator and primary-to-secondary leakage to ,

the intact steam generators are conservatively assumed to continue until 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the tube rupture.

f. The iodine partition faction between the liquid and '~

steam in the steam generator is assumed to be 0.01.

g. The steam releases from the steam generators to the atmosphere are given in Table 15.6-4.

h. Offsite power is lost.

1. Five hours after the accident, the RHR system is l assumed to be in operation to cool down the plant.

Thus, no additional steam release is assumed.

j. Radioactive decay prior to the release of activity is considered. No decay during transit or ground deposi- ,

tion is considered.

k. Short-term accident atmospheric dispersion factor, breathing rates, and dose conversion factors are pro-vided in Tables 15A-2, 15A-1, and 15A-4, respectively.

15.6.3.3.1.3 Mathematical Models Used in the Analysis Mathematical models used in the analysis are described in the following sections: --

a. The mathematical models used to analyze the activity (gdQ released during the course of the accident are d rcribed sw-h '77-"^i 1% fores (en -l)e asayNont lish) dove.

b. The atmospheric dispersion factors used in the analysis were calculated based on the onsite meteorological measurements program, as described in Section 2.3 of the Site Addendum, and are provided in Table 15A-2.

c. The thyroid inhalation immersion doses to a receptor at the exclusion area boundary and outer boundary of the low-population zone were analyzed, using the models described in Appendix 15A.

Rev. OL-5 15.6-13 6/91

r= 1

,' 'jg i CALLAWAY -'SP 1

."'"- 15.6.3.3.1.4 Identification of Leakage Pathways and Resultant W' td . Leakage Activity p - Neam amp For the purposes of evaluating the radiological consequendes due to a postulated SCTR, the activity released from the af-fected steam generator, prior to isolation, s released directly to the environment by the' atmospheric-relic _ valve. The un- l affected steam generators are assumed to continually discharge steam and entrained activity via the atmospheric eelief Avalvey l up to the time initiation of the RHR system can be M edm d, M*y#

accomplished. Since the activity is released directly to the environment with no credit for plateout or retention, the

,,,... results of the analysis are based on the most direct leakage

@R11 pathway available. Therefore, the resultant radiological consequences represent the most conservative estimate of the potential integrated dose due to the postulated SGTR.

15.6.3.3.2 Identification of Uncertainties and Conservatisms in the Analysis

a. Reactor coolant activities based on extreme iodine spiking effects are orders of magnitude greater than that assumed for normal operating conditions. .

l

b. A 1-gpm steam generator primary-to-secondary leakage, with a conservatively high density, is assumed which l is significantly greater than that anticipated during normal operation. This leakage continues for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, even though RHR operation is assumed to begin at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

c. Tube rupture of the steam generator is assumed to be a double-ended severance of a single steam generator tube. This is a conservative assumption, since the steam generator tubes are constructed of highly ductile. materials. The more probable mode of tube failure is one or more minor leaks of undetermined origin. Activity in the secondary steam system is JP subject to continual surveillance, and the accumula-pypIj h tion of activity from minor leaks that exceeds the

' limits established in the technical specifications would lead to reactor shutdown. Therefore, it is highly unlikely that the total amount of activity considered available for release in this analysis would ever be realized,

d. The coincident loss-of offsite power with the occur-rence of an SGTR is a highly conservative assumption.

In the event of the availability of offsite power, the condenser dump valves will open, permitting steam dump to the condenser. This will reduce the amount of steam and entrained activity discharged directly to the environment from the unaffected steam generators.

Rev. OL-5 15.6-14 6/91

r o '

i ,

CALLAWAY - SP

e. The meteorological conditions which may be present at  :

the site during the course of the accident are -

uncertain. However, it is highly unlikely that the l meteorological conditions assumed will be present during the course of the accident for any extended period of time. Therefore, the radiological consequences evaluated, based on the meteorological conditions as ume , are conservative.

rr re f 4 /g (/)f

f. The radiolog cal consequences have been bas ed on a worst case single failure in the open position of the faultcdtsteam generator atmospheric relieO' valve

-(,kRPP which isn't isolated until 30.4 minutes after tube rupture. SA/fEgr- g

q. The las d acti of te br ak f w to he f ite s am ener or con rvat ely esum to ve a ig th ca ulat , co tan value of 11",whi t fau ed s eam g nerat r AR is s ck o n (f r ef st h r of he t nsi t).

l

h. The flashed fraction of the primary-to-secondary leakage to the intact steam generators is conservatively assumed to be the same as in the

-faultcdA s eam enerator. - SNTER r- /_

""f h- '_ )

1. re f w to he aulte stea gene tor i I c ser ativ ya umed o hav ah er t n l alc ate , co tant alue 55 m/se while the I fau ed team ener or A is e uck o en (f th f st ur o the ransi t) a d 10 m/sec er fter hrou 8 ho rs, en th gh op .ation; s sume to b gin a 5 ho s end'ng th tra ient.)

j. Stea relea s tn ugh le fa ted eam ner or (ARVareteminatdwhe bloc valv clos re a 30.4 m utes solat s ste rele ses om t s at am Th/[Eg7~ f n k steam release from the intact steam generator rig j4fbr The ARVs during the 5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> cooldown to RHR cut-in conditions is conservatively assumed to occur in its j entirety during the 0-2 hour period of the transient. I

1. Whole body doses from the intact steam generator-ARVs-A fDs during the cooldown to RHR cut-in conditions are calculated using conservative primary side activities.

15.6.3.3.3 Conclusions 15.6.3.3.3.1 Filter Loadings The only ESF filtration system considered in the analysis which _.

limits the consequences of the steam generator tube rupture is Rev. OL-5 15.6-15 6/91

m

..- t ! .

INSERT K l

l The flashed fraction of the break flow to the ruptured steam generator varies over time. The values assumed in the radiological consequence calculation

. conservatively bound the calculated values. Key events which trigger changes in the assumed flashed fraction include reactor trip and closure of the manual block valve to isolate the failed open SG atmospheric steam dump valve. Specific values for the flashed fraction are listed in Table 15.6-4.

I INSERT L Break flow to the ruptured steam generator is conservatively assigned values that bound calculated break flow rate values. Break flow rate values are discussed in Table 15.6-4.

INSERT M i 1

i There are two steam release pathways from the ruptured steam generator that l are addressed by the radiological consequence calculation. These pathways are  !

l the steaming and flash pathways. The steaming pathway accounts for the boiling of the secondary side water inventory of the ruptured steam generator. The flash i l pathway accounts for the fraction of the leaked primary fluids which immediately flashes to steam after arriving in the secondary side. Release via the steaming pathway is temtinated by the SG atmospheric steam dump block valve closure at

!' 30.4 minutes. Ralease via the flash pathway is conse.rvatively continued ,

following block Valve closure. Release via this pathway is continued until the '

RETRAN resulb, indicate that no further flashing will occur.

, a; / ~

• CALLAWAY - SP TABLE 15.6-1 l TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH RESULT IN A DECREASE IN REACTOR COOLANT INVENTORY

!"N Time Accident Event (sec)

-Inadvertent opening of a j pressurizer safety valve Safety valve opens fully 0.0 )

i

^

Overtemperature AT reactor trip setpoint, reached 26.2 Rods begin to drop 28.2 Minimum DNBR occurs 28.4

' ' D' Steam generator tube rupture Tube rupture occurs 0.0 Reactor trip signal 613.3 Safety injection signal 613.3 Rod motion 615.3 Feedwater perminated 617.6 fup+ured SnultedVsteam generator atmospheric /reli-fxvalve opens 625.6 Seem dumjo Safety injection begins 628.3 Auxiliary feedwater injection 675.3 Operator isolates faulted steam generator by closing manual block valve 1826 Operator initiates RCS cooldown via intact steam generator atmospheric rcli;fA val es _2447 gg, b** "Y  !

g.g.g Operator completes RCS cooldown 3288

-e Operator initiates RCS depressurization via pressurizer PORVs 3469 Operator completes RCS depressurization 3558 Operator terminates safety injection 3739 l Operator equalizes primary-secondary pressure 4039 i RHR cut-in conditions reached 18000 Rev. OL-5 6/91

< *, : s l

CALLAWAY - SP 1

TABLE 15.6-4 ,,

PARAMETERS USED IN EVALUATING J'i THE RADIOLOGICAL CONSEQUENCES OF ,

A STEAM GENERATOR TUBE RUPTURE (SGTR) )

I. Source Data

a. Core power level, MWt 3,636'

b. Steam generator tube 1 -

leakage, gpm

c. Reactor coolant iodine F MI@ M  ;

activity: jy l v n .- j

1. Case 1 (Ini al ctiv ify e al dos equ alen of .0 pC /gm fI' wi a ssu d iod es ke at in cases e r te i dine leas in th eacto cool nt a J

fact of 5 0

2. Case 2 ran sum pre- cci nt ,

l odin spike, wh ch h a res ted i the dose '

e valen of 0 pC /gm o I-131 T rNSEW D \

d. Reactor coolant noble gas Based on 1-percent failed activity, both cases fuel as provided in Table 11.1-5

e. Secondary system initial _ Or c equivelmuL vf -

activity --.. ...,v... vf I-151 /d% o}? CLf, /

frithary ride ac-livi-ly

f. Reactor coolant mass, lbs '

/ i 4

1. In reactor vessel 2.1E+5

2. In total primary system 5.3E+5 p.7,3 6djh.)

rQ.,

g. Steam generator mass (each), lbs

1. Water 1.0E+5

2. Steam 7.7E+3

h. Offsite power Lost

1. Primary-to-secondary 67.3 minutes leakage duration II. Atmospheric Dispersion Factors See Table 15A-2 Rev. OL-5 6/91

, n .j i I

INSERT N 1

The initial reactor coolant lodine activity corresponds to an isotope mixture that bounds Technical Specification allowable conditions for both tight and open fuel defects. The initial isotopic mix is based on the relative concentrations from Table 11.1-5. The concentrations are then changed to achieve a Dose Equivalent 1-131 (DEI) of 1.0 Ci/gm, while maintaining the isotopic ratios from Table 11.1-5. This provides conservative values for the longer lived lodines which contribute the majority of the calculated thyroid dose. The initial inventories of the shorter lived iodine isotopes, which can provide a significant coitribution to the calculated whole body dose, are then increased to conservatively bound isotopic mixes which may occur in the presence of open fuel defects. Case 1 then includes an accident initiated, spiked release rate that increases by a factor of 500 during the accident sequence.

INSERT O The initial reactor coolant iodine activity corresponds to an assumed pre-accident iodine spike which results in concentrations that are a factor of 60 higher than those used in Case 1.

\

I 1

l l

E

_( ,5** $

CALLAWAY - SP 1

ddA i f f ., ?. TABLE 15.6-4.(Sheet 2)

Q?if III. Activity Release Data

a. Affected steam generator

1. Reactor coolant dis-charged to steam gener- -

y) ator, lbs 45jl2g00 ,

2. Flashed reactor coolant, 11  !

percent

3. Iodine partition factor 1.0 fk9,5;9$'r for flashed fraction of

- reactor coolant

4. Steam release to j atmosphere, lbs )

0-2 hrs 114,300 q 2-8 hrs 0 1

5. Iodine carryover. 0.01 factor for the non-flashed fraction of reactor coolant that mixes with the initial  ;

iodine activity in the steam generator

b. Unaffected steam generators

1. Primary-to-secondary 3,456(3) leakage, lbs

2. Flashed reactor coolant, Variable percent

3. Feedwater flow, lbs 0-2 hours 2.05E+6 2-8 hours 0

4. Total steam release, lbs

.U34 0-2 hours 1.43E+6 I4) 2-8 hours 0

$h{h 5. Iodine carryover factor 0.01(5) 1

6. RHR Cut-in time, hrs 5 4 f- ."2A/f687~I Notes:

1. Ass .s afd'ons vati ly h , c stan 55 1 /se rea w fo the irst our d 10 m/s the afte thro gh i 8 ho s, e n th gh RH oper tion s ass med beg at p/

5 urs.

2. co anc asne rac on of it .13/ assume dur g th r' fi t hou of th tra lent ile 4 faul d st am nerat atmo heri relie valve ope int t ed f ction con rvat ely f steam enera r fl (ass ed to e the ame a in t i' alt d ste mRev. ge. ry h l OL-5 i Zh/SGRf~ Q

F

+ , ., 4 4 INSERT P The noble gas release calculation assumes a conservatively high, constant 55 lbm/sec break flow rate for the first hour and 10 lbm/see thereafter through 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, even though RHR operation is assumed to begin at 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The iodine release calcdation is based on the conservative break flow rate of 55 lbm/sec until cooldown is completed.

INSERT Q Prior to reactor trip, the assumed flashed fraction is 15%. Immediately after reactor trip un';! closure of the SG atmospheric steam dump block valve, the assumed fissned fraction is 11% Following closure of the block valve, a variable flashed fraction is assumed which conservatively bounds the values calculated by the RETRAN code. The intact steam generator flashed fraction is i conservatively assumed to be the same as in the ruptured steam generator.

I .n i* [p A CALLAWAY - SP M EA TABLE 15.6-5

' ' Y,ffig.!

RADIOLOGICAL CONSEQUENCES OF A STEAM GENERATOR TUBE RUPTURE l 1,

I l Doses (rem)

1. Caee 1, a ecoNeor/- (n,-kf } f,),- ,g F .;lunon Area Boundary (0-2 hr) l

[h $3' Thyro:.d 22 7

-0. 050 g, g43 Whole body Low Population Zone Outer Boun(.ary (duration)

Thyrt.id 4.:;- 2.29 Whole body -0 000'-4.0/f.S

2. Case 2 3

n acaf},n-}-f,};n,3 ,-p, Exclusion A' tea Boundary (0-2 hr)

Tlyrroid 44,&- 34,3 Whole hudy -0. 07P 0,32f Low 1 ,pulation Zone Outer Boundary (duration)

Thyroid 2 . '. 0 - 'J.f3 Whole body J. 011 O,4347 Yi ?l l

i Rev. OL-5 6/91

_