ML20249A268

Proposed Tech Specs Re Revised Steam Generator Tube Rupture Analysis (Plar 3-98-4)

ML20249A268
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Site:	Millstone
Issue date:	06/05/1998
From:	NORTHEAST NUCLEAR ENERGY CO.
To:
Shared Package
ML20249A259	List: B17288, Application for Amend to License NPF-49,by Incorporating Attached Proposed Rev Into Chapter 15 of Millstone,Unit 3 FSAR ML20249A268
References
NUDOCS 9806160254
Download: ML20249A268 (51)

Text

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a Docket No. 50-423 B17288

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Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Revised Steam Generator Tube Rupture Analysis (PLAR 3-98-4)

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June 1998 9806160254 980605 PDR ADOCK 05000423 P PDR

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U.S. Nuclear Regulatory Commission i

817288\ Attachment 2\Page 1 MARKUP OF PROPOSED REVISION 1

) Refer to the attached markup of the proposed revision to the Final Safety Analysis Report (FSAR). The attached markup reflects the currently issued version of the FSAR.

The following FSAR changes are included in the attached markup:

1. Table 15.0 Initial NSSS Thermal Power Output Assumed for steam generator tube failure is being changed from 3500 to 3494. The analysis, as documented in Section 15.6.3.2.2.1 uses an initial power of 3493.5.

However, 3494 will be used for consistency since this column of Table 15.0-2 references' Table-15.61-in which the value has been rounded to 3494.

2. Table 15.0-8 : The Steam Generator Tube Rupture dose values for the thyroid except for low population zone with concurrent iodine spike are being updated.

3. Section 15.6.3.2.1 : Two changes are being made to this section. The first change deletes References WCAP-13002, WCAP-13056, and NEU-98-005.

WCAP-13002 and NEU-98-005 were specific Millstone 3 margin to overfill analyses. These references have been superseded by the most recent margin to overfill analysis and therefore are not appropriate. WCAP 13056 was the three loop margin to overfill analyses which is no longer valid and therefore not an appropriate reference. The second change to this section is incorporation of the new margin to overfill single failure of the atmospheric

. dump by l

, demand.# pass va ves on two ~ ~of

' the

~ " ~intact

~ ~ steam

" generators to open on

4. Section 15.6.3.2.2.1: Specific to the analyses, three assumptions are added to the list of important assumptions and the numerical values presented in the section are being changed where necessitated by the new analysis. The three assumptions being added are the flow capacity and its associated pressure used for the atmospheric dump valve, the atmospheric dump bypass valve and the pressurizer power operated relief valve. The numerical values modified only incorporate the new analysis. ,

5. Section 15.6.3.2.2.2 : Two changes are being made to this section. The word iodine was added so that the wording is changed from " .. maximum allowable Technical Specification limit for primary coolant activity..." to

" maximum allowable Technical Specification iodine limit for primary coolant activity.. " This is a non-intent clarification. The second change is that the

4 U.S. Nuclear Regulatory Commission B17288\ Attachment 2\Page 2 sentence "The initial noble gas concentrations in the reactor coolant are based upon 1-percent fuel defects" is being changed to read "The initial noble gas concentrations in the reactor coolant are based upon Technical Specification Limits". This changes makes it clear that the initial noble gas concentration is also based on Technical Specification Limits.

6. Section 15.6.7 : The references to WCAP-13002/13003, WCAP-13056/13057 and NEU-98-05 are being deleted. This is a reflection of the change discussed under item 3 above.

7. Table 15.6.3-2 : The times in the sequence of events were updated to incorporate the appropriate times based on the nev7eh'effsis. -~ ~ - -

8. Table 15.6.3-3 : The mass releases for the SGTR analysis were updated to incorporate the appropriate mass releases based on the new analysis.

9. Table 15.6.3-4 : The assumed noble gas activity was changed from " based on 1% fuel defects" to " based on Technical Specification limits". This change clarifies this wording as discussed in item 5 above.

10. Table 15.6.3-5 : The numerical values for the iodine specific activity in the secondary coolant have been changed. This change incorporates the appropriate specific activity based on the new analysis. This specific activity is consistent with Table 15.0-10.

-" NT.fhbTei5.65-7': ~TiiiIassiltn@d noble'dds' activity was cKa"nged from " based "' ~~

on 1% fuel defects' to " based on Technical Specification Limits". This change clarifies the wording as discussed in item 5 above

12. Table 15.6.3-10 : The calculated offsite doses for the SGTR analysis were updated. This change reflects the appropriate doses based on the new analysis.

13. Figures 15.6.3-1 through 15.6.3-12. All the figures associated with the SGTR-offsite dose analysis are being replaced. This change incorporates the appropriate figures based on the new analysis.

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98 MP3 .20 3C TABLE 15.0-8 POTENTIAL OFFSITE DOSES DUE TO ACCIDENTS Dose (rem) Dose (rem) 2 hr Exclusion Low Population Area Boundarv (524 m) , . Zone (3862 m)

Postulated FSAR Accident Section Thyroid Gamma Thyroid Gamma Steam Generator Tube Rupture .

15.6.3 g'.

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a. Preaccid'ent '

iodine spike g.o E-oi E,00 ( l.oE 01

- 2.1E : 0 1.0E 02 - .,C 01 1.25 00

b. Concurrent -

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iodine spike GEM'

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-1.1 03 3 LOCA 15.6.5 1.4E+ 02 9.4E + 01 3.0E + 01 Weste Gas System 1.7E+fo l h'.

F"ilure 15.7.1 0.0E + 00 2.2E-01 (2) (2)

R dioactive Liquid -

Waste System Leak or F:.ilure (Atmospheric R: lease) 15.7.2 4.3E-01 4.7E-04 (2) (2) gi y Fuel Handling Accident 15.7.4 7.6E+ 00 5.1 E-01 (2) (2)

Sp:nt Fuel Cask Drop 15.7.5 (3) ,3) i ,

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(1) 1.6E-01 = 1.6 x 104 (2) 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />'sof release or less; a 30-day dose is not applicable. ~

(3) Not applicable-see Sections 15.7.5.2 and 15.7.5.3.

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issos.ues 2 of 2 l March 1994

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J The rupture b ufated activities released to the environment as a result of a steam generator t 1 on the parameters described in Tables 15.6-5 and 15.6-6 are shown i able 15.6-7. The t sted release values are for the two cases examined:

1. an assume eaccident iodine spike condition in the reac coolant, and

2. an accident initiate current iodine spike.

The releases together with the atmospheric rsi used to compute the doses presented in Table actors listed in Table 15.0-11 are

. -8 for the EAB (0-2 hr) and the LPZ (0-5 h.r).

The whole body and thyroid dos Iculated for the postulat ccident assuming a

. . preaccident iodina spike in reactor coolant is less than the do uideline values ~

described in 10CFR10 '.e.,300 Rem to the thyroid and 25 Rem to hole body.

For the assu condition of a concurrent iodine spike in combination with eq iodine rium entrations at full power, the analysis of the postulated accident resulte ~

dose val

- less than a small fraction of 10CFR100;i.e.,30 Rem to the thyroid and 2.5 Re

( ole body.

An SGTR resu

( ts-trr1ho b%c, af wntaminated reactor coolant into the secondary system and subsequent release of a portion of the activity to the atmosphere, and an analysis is typically performed to assure that the offsite radiation doses resulting from an SGTR are within the allowable guidelines. However, one of the major concerns for an SGTRis the possibility of steam generator overfill since this could potentially result in a significant increase in the offsite radiation doses. Therefore, to ensure that steam generator overfill O[

5 will not occur for a design basis SGTR for Millstone Unit 3, an analysis was performed to demqpstr pin to steam.annaratnr nyerfilj assuming the limitingjingle falhlte rela '

to Everfil (WCAP-13002p .!O gjgfr- f50Kd.hal A/E(/ T q[bkh The steam generator tube rupture analysis was ed for Millstone Unit 3 using the methodology developed in WCAP-10698. This analysis methodology was developed by the SGTR Subgroup of the Westinghouse Owners Group and was approved by the NRC in a Safety Evaluation Report dated March 30,1987. The LOFITR2 program, an updated version of the LOFlTR1 program, was used to perform the SGTR analysis for Millstone Unit 3. The LOFTTR1 program was developed as part oithe revised SGT analysis l methodology and was used for the SGTR evaluations in%VCAP-1069 . However, the LOFTTR1 program was subsequently modified to accommodate steam generator overfill and the revised program, designated as LOFTTR2, was used for the evaluation of the consequences of overfill in WCAP-11002. The LOFTTR2 program is identical to the LOFTTR1 program, with the exception that the LOFTTR2 program has the additional capability to represent the transition from two regions (steam and watert on the secondary side to a single water region if overfill occurs, and the transition back to two regions again depending upon the calculated secondary conditions. Since the LOFTTR2 program has

! been validated against the LOFTTR1 program, the LOFTTR2 program is also appropriate for performing licensing basis SGTR analyses.

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Nt h d.s c.% Ayyemseg f N lt.C5 i op r < h, .a y 4e Oc3' a /sg y,q.Mr.4, N a i5 operations personnel using the plant training simulator. Thus, the r/u SGTR Frp/(n -

on the application of the actual plant procedures and operator training. J

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g The limiting single failure was assumed to be onsistent with the methodolo WCAP-10698. The LOFTTR2 analysis to determine the margin to ov

! for the time period from the tube rupture until the primary and secondary ruptured steam generator was' calculated as a fu does not occur. The results of this analysis demonstrate that there is ma T enerator overfill for a design t-ds SGTR for Millstone Unit 3.

$8 h .6.4applic'eble Spectrum of BWR to Millstone 3. Steam System Piping Fellures Outside of Containm

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15.6.5within Loss-of-Coolant Accidents Resulting from a Spectrum the Reactor Coolant Pressure Boundary s

of Postula 15.6.5.1 Identification of Causes and Frequency Classification A LOCA is the result of a pipe rupture of the RCPB (Section 5.2). >For the reported here, a major pipe break (Istge break) is defined as a rupture with a tota an ANS Condition IV event, a limitirig fault, in that it

' lifetime of the plant but is postulated as a conservative design basis (S

(

Millstone 3 nucleer steam supply system (NSS) contains loop re Isolation v can be operated with one loop out of service. For the analyses presented ed ma}or pipe break (large breakl could occur in one of the active loops o vessel side of the Isolated (inactive) loop while in this (N-1) loop configuration. ,

A minor pipe break (small break) considered liereJs defined as a rupture of. Fedeactor" "d*#

u c^coolanfpresstfre b5uiYary with a@ total cross- sectfonal area less t normally' operating charging system flow is not sufficient to sustain pressurize pressure.

may occur duringThistheislife considered of the plant. a Condition lit event,in that it is an infrequent fault w The AppendixAcceptance Criteriafollows.

K of 10CFR501974)as for the LOCA are described in 10CFR50.46(10CF .

1. "

The calculated of 2,2OO

• F. peak fuel element clad temperature is below the requ

-2.

The amount of fuel ' element cladding that reacts chemically with water or steam reactor. does not exceed 1 percent of the total amount of Zirceloy in the 3.

The clad temperature transient is terminated at a time when the core ge try is still amenable to cooling. The localized cladding oxidation limit of 17 percent is not exceeded during or after quenching.

4.

The core remains amenable to cooling during and after the break. ('

isse.ws 15.610 -

98 MP3 20 20 3 see-f D 1-(f3.t.s'.c.-oy s'/~f /f _

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1 15.6.3.2.2 Offsite Radiological Consequences ~

l An analy/is was also performed to determine the offsite radiological consequen SGTR using the methodology in WCAP-10698 and Supplement 1 to WCAP-10698, assuming the limiting single failure relative to offsite doses without steam generator l

overfill. Since stearn generator overfill does not occur, the results of this analysis represent  !

the limiting consequences for an SGTR for Millstone Unit 3. A thormal and hydraulic l

analysis was performed to determine the plant response for a design basis SGTR, and to I determine the integrated primary to secondary break flow and the mass releases from the l

ruptured and intact steam generators to the condenser and to the atmosphere. This '

information was then used to calculate the quantity of radioactivity released to the environment and sie resulting radiological consequences.

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_ 15.6.3.2.2.1 Thermal and Hydraulic Analy' sis The plant response following an SGTR was analyzed witIt the 1.0FTTR2 program for the time period from the tube rupture until the primary and secondary pressures are equalized and the break flow is terminated. The reactor protection system and the automatic .

actuation of the engineered safeguards systems were modeled in the analysis. The major operator actions which are sequired to terminate the break flow for an SGTR were also simulated in the analysis.

_ Analysis Assumptions .

t The accidelit modeled is a double-ended break of one steam generator tube located at the

,, top of the tube sheet on the outlet (cold leg) side of the steam generator. It was assumed that the reactor is operating at full power at the time of the accident and the secondary mass was conservatively assumed to be 10% less than the mass corresponding to full power operation at the' nominal steam generator water level. It was also assumed that a

" loss of offsite power _occigs.at thepe of.rdactorgaod.the highestanforth control asserfibTy wis assumed to be~s$ckin its fully withdrawn'5osition at reactor trip. .

Other important analysis assumptions include:

A. NSSS power = 3425 MWt* 1.02 (uncertainty) = 3493.5 MWt B.' Average RCS temperature = 587.1*F C. RCS pressure = 2250 psia - 53 psi (uncertainty) = 2197 psia D. Thermal design flow = 363,200 gpm

~

E. Pressurizer level = 62% -

F. SG tube plugging = 10%

G. Auxiliary feed flow = 635 gpm @ 1140 psia H. Auxiliary feed flow delay time = 60 seconds y- -

l.

Steam generator atmospheric dump valve flow = 820,000 lbm/hr/ valve @ 1140 psia \

l ,

J. Steam generator atmospheric dump bypass valve flow = 820,000 lbm/hr/ valve ~ l

@ 1140 psia L K. Pressurizer power operated relief valve flow = 210,000 lbm/hr/ valve @ 2500 psia

ls ,

m raa 98 MP3 2Q (h 1sw D sk p The limiting single failure was assumed to be the failure of the r 25 :=

M,,.fre"r .r, valve (idSPSS on the ruptured steam generator. Failure of this L'49V = ,g der,sf/er:c.4 in the open position will cause an uncontrolled depressurization of the ruptured stekn generator which willincrease primary to secondary leakage and t ass release to the atmosphere.

It was assumed that the ruptured steam n t ils open when th i ruptured steam generator is isolated, and that th is iso ate by lly closing 1 he associated block valve.

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The major operator actions required for the recovery fr R ed in Section 15.6.3.1, and these operator a e simulat s. The operator l action times used for the a e based on t ts o one Unit 3 plant trainin sim lato es, and the time l resente abt .6.31. It is noted that

, the S on the ruptured ste enerator ssume o fail open at the time the ruptured steam enerator is51M oriflirtree og with the recovery operations, l

the failed-open 'e. " en the red steam ge ator was assumed to be isolated by locally closing the associ ck valve. It s assumed that the ruptured steam

~

generator at inutes the valve was assumed to fall open.

After the ruptured steam gener tor was isolatedt an additional 8 minute delay .

(Table 15.6.3-1) was assumed b ore the operator initiates the RCS cooldown.

Transient Description 20 D

((M ,

The LOFTTR2 analysis results are described below. The sequence of events for this '

! transient is presented in Table 15.6.3-2.

Following the tube rupture, reactor coolant flows from the primary into the secondary side

' of the ruptured steam generator since.the primary pressure is greater than the steam generator pressure. In response to this loss of reactor coolant, pressurizer level decreases l

as shown in Figure 15.6.3-1. The RCS pressure also decreases as shown in W ~" Figure 15.6.3-2, as the steam bubble in the pressurizer expands. As the RCS pjr ssure.. . _ . _ ,

'tfecreassidue to tiilf'6ontinued primary 1o secondary leakage, auto

• Mictbr tno occurlF ~~

on art overtemperature AT trip signal at 109 seconds.

cdensph,te. /,,,p ;se /w.s -

After reactor trip, core poy<er rapidly decreases to cay heat levels. The turbine stop

~ valves close, and steam flow to the turbine is rminated. The steam dump system is designed to actuate following reactor trip limit the increase in secondary pressure, but the steam dump valves remain close ue to the loss of condenser vacuum resulting from the assumed loss of offsite powe t the time of reactor trip. Thus, the energy transfer from the primary system caus the secondary side pressure to increase rapidly after reactor trip until th and safety valves if their setpoints are reached) lift to dissipate the energy, as shown in Figure 15.6.3-3. The main feedwater flow will be ~

terminated and Al"W flow fror'n the two motor-driven AFW pumps will be automatically initiated following reactor trip and the loss of offsite power.

The RCS pressure and pressuriz,er level continue to decrease after reactor trip as energy j transfer to the secondary shrinks the reactor coolant and the tube rupture break flow continues to deplete primary inventory. The decrease in RCS inventory results in a low pressurizer pressure SI signal at approximately 381 seconds. After Si actuation, the Si flow rate initially exceeds the tube rupture break flow rate, and the RCS pressure and L

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8 MP3 20 h

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pressurizer level begin to increase and approach the equilibrium values where the Si flor rate equals the break flow rate. ~~

Since offsite power is assumed lost at reactor trip, the RCPs trip and a gradual transition to natural circulation flow occurs. immediately following reactor trip the temperature differential across the core decreases as core power decays (see Figures 15.6.3-4 and 15.6.3-5); however, the temperature differential subsequently incre6ses as natural circulation flow develops. The increase in the temperature differential slows the rate of the pressurizer level and pressure decrease as shown in Figures 15.6.31 and 15.6.3-2, respectively. The cold leg temperatures initially trend toward the steam generator temperature as the fluid residence time in the tube region increases. The RCS hot leg temperatures slowly decrease until the time when the ruptured steam generator MSPfW is assumed to:f ail open.

g g4 gg Maior Operator Actions

1. Identify and Isolate the Ruptured Steam Generator The ruptured steam generator was assumed to be identified and isolated when the narrow range level recovers to 29% on the ruptured 5 team generator or at 16.5 minutes ,~

after initiation of the SGTR, whichever is longer. For the Millstone Unit 3 analysis, the time to reach 29% narrow range levelis greater than 16.5 minutes, and thus the ruptured steam generator is assumed to be isolated at 1442 sepade wh . the level reaches 29%.

,g g The ruptured steam generator W ass ed fai open at this time.

However, the actual time used e analysis is sec ds nger because of the computer program nume ' requirements for s ulat g e operator actions. The failure of the auses the ruptured stes gen rato to rapidly depressurize as shown in Figure 15.6.3-3. The depressurizati of e ptured steam generator increases the break flow (Figure 15.6.3-6), a th ene y transfer from primary to c

_ seco.r}dary, results in a decrease in the ruptur lo te peratures as shown in _ e -

Figure 15.6.3-4. The intact steam generato loo te peratures also decrease, as ,

shown in Figure 15.6.3-5, until the failed-o en '

is isolated. The decrease in the RCS temperatures results in an initial decr ase in e pressurizer level and RCS pressure. However, the increased Si fio subseg ently causes the pressurizer level and RCS pressure to stabilize and begin to in ease a shown in Figures 15.6.3-1 and 15.6.3 2, respectively. It was assu ed t the time required for the operator to identify that t ptured steam genera r is open and to close the associated block valve is inutes. Thus, the was isolated at seconds which terminates th pressurization of the ruptured steam generator, and the ruptured steam generator ressure begins to increase after that time. -

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2. C6ol Down the R establish Subcooling Margin

' After the ruptured steam generator WmA.2 M nh*

block valve is closed, an 8 minute operator action time was imposed prior to initiation of cooldown. Since offsite power is lost, the RCS was cooled by dumping steam to the atmosphere using the McoRNc on the three intact steam generators. The cooldown was continued until RCS ocooling at the ruptured steam generator pressure is 20*F plus an allowance of 2*F for subcooling a fa,nph k h p $ b)YM6 N .

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_mn 98 MP3 20 I M-

&sa) e m uncertainty The depressurization o}the ruptured steam generator due to the f ailed-

&f open.MSPRV affects the RCS cooldo un target temperature since the target temperature is dependent upon the pdessure in the ruptured steam generator. Althoug the ruptured steam generator presst. o is increasing when the cooldown is initiated, the pressure is substantially below the setpoint. The lower pressure in the ruptured steam generator results in a lower RCS cooldown temperatur required for cooldown. The RCS cooldown was initiated at hich extends the time th ds by opening 4824 second(s.PRBV on the three intact staam nenerators, aand was comp p-

, ,4.,. Q , p & } y p. u a f , Jg26 The re uction in the intact steam generator pressure required to accomplish the cooldown is shown in Figure 15.6.3-3, and the effect of the cooldown on the intact loop temperatures is shown in Figure 15.6.3-5. The RCS pressure also decreases initially during the cooldown due to shrinkage of the reactor coolant as shown in Figure 15.6.3-2, and then begins to increase again as the SI flow increases. It is noted

- that the rupturedsteam genisiiT5riiressure continues to increase during th'e cooldown until the pressure approaches the MJpRY setpoint aaain.

i

3. Depressurize to Restore Invento m *sp4**se l*"r A

/i 48N D*JP T/Zylf After the RCS cooldown was completed, a 3 minute operator action i was assumed prior to depressurization. The RCS depressurization was initiated at econds to assure adequate coolant inventory prior to terminating Si flow. With t e RCPs stopped, normal pressurizer spray is not available and thus the RCS was depressurized by opening a pressurizer PORV. The depressurization was continued until any of the following conditions are satisfied: RCS pressure is less than the ruptured steam generator pressure and pressurizer levelis greater than the allowance of 16% for pressurizer level uncertainty, or pressurize.rievel is greater than 73%, or RCS subcooling is less than the 32 F allowance for subcooling uncertainty. The RCS depressurization reduces the break flow as shown in Figure 15.6.3 6, and increases Si flow to refill the pressurizer as shown in Figure 15.6.3-1. For this case, the

- - depressurization was terminated when the RCS pre surgis reducelto t_he, ruptured. ,.

steam generator press 6re (Figure 15.6.3-7) and tfi'e pressurizer levelis above 16%

(Figure 15.6.3-1). Although the pressurizer levelis less than 73% when the pressurizer PORV is closed, evel continues to increase and subsequently reaches' a maximum of approximately when Si flow is terminated.

e

4. Terminate Si to Stop rimary to Secondary leakage ce >lgM The previous actions have established adequate RCS subcooling, verified a secondary side heat sink, and restored the reactor coolant inventory to ensure that Si flow is no longer needed. When these actions have been completed, the Si flow m,ust be stogged to prevent repressurization.of the RCS and to terminate primary to secondary leakage.

The Si flow is terminated at this time if RCS subcooling is greater than the 32*F allowance for subcooling uncertainty, minimum AFW flow is available or at least one intact steam generator level is in the narrow range, the RCS pressure is increasing, and the pressurizer levelis greater than the 16% allowance for uncertainty. To assure that the RCS pressure is increasing, SI was not terminated in the analysis until the RCS pressure increases by at least 50 psi.

' MAY-29-98 FR1 12:03 PM SAFETY ANALYSIS BRANCH FAX N0. 860 832 4784 W P. 11 q 1,,_g.... ,m s ~~~DmM"_jcsf~~h

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After depressurization was completed, an operator action time of 3 minutes was

-  ?

assumed prior to Si termination. Since the above requirements are satisfied, S1 termination was performed at this time. After si termination, the RCS pressure begi t

andr crease as shown in Figure 15.6.3-2, and the differentist pressure between the RCS red steam generator decreases as shown in Figure 15.6.3-7. The intact steam generstof,.'0*E&vs are also opened to dump steam to maintain the prescribed RCS  !

temperature to ensure that subcooling is maintained. When the MGPRSVs are opened l the increased energy transfer from primary to secondary also aids in the ,

depressurization of the RCS to the ruptured steam generator pressure. As in she Figure 15.6.3 6, the primary to secondary leakage continues after the Si flow is terminated until the RCS and ruptured steam generator nrassures eaustin 4 aMsphsv.d.wplypesya/*eD The ruptured steam generator water volume is s[hown the water volume in the ruptured steam generator is significantly less tharLIha_ total ata 3 . . . . .

generator volume of 5850 ft when the break flow is terminated. The mass of water in the ruptured steam generator is also shown as a function of time in Figure 15.6.3-g.

_ Mass Releases , ,

The mass releases were determined for use in evaluating ~the exclusion area boundary "

low population zone radiation exposure. The steam releases from the ruptured and intact

- steam generators, the feedwater flows to the ruptured and intact steam generators, and primary to secondary break flow into the ruptured steam generator were determined for the

- period from accident initiation until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the accident and from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the accident. The releases for 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were used to calculate the radia exclusion area boundary for a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> exposure, and the releases for O - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> were used {

j to calculate the radiation doses at the low population zone for the duration of the accident. '

The operator actions for the SGTR recovery up to the termination of primary to secondar leakage were simulated in the LOFTTR2 analysis. Thus, the steam releases from the ruptured and intact steam generators, the feedwater flows to the ruptured and intact steam  !

generators, and the primary to secondary leakage into the Hiituredl steam generator were j

determined the leakage isfrom the LOFTTR2 results for terminated. '

the period from the initiation of the accident until .

~

Following the termination of leakage, it was assumedf actions are toren to' cool down the plant to cold shutdown conditions. The " Err.L . r the intact steam generators were assumed to be used to cool down the RCS to the RH tem operating temperature of 350*F, at the maximum allowable cooldown rate of the feedwater flows for the intact steam generators F/ hour. The steam rolesses and he period from leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were determined from a mass and energy balance using the calculated RCS and intact steam generator conditions at the time of leakege termination and at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. ~

The.RCS cooldown was assumed to be continued after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> until the RHR system in-service temperature of 350*F is reached. Depressurization of the ruptured steam generator to the RHR in-service pressure of 390 psia was then assumed to be performed via steam 3T. The RCS press'ure was a!co assumed to be reduced concurrently as the ruptured stea{m generator is dep assumed that the RCS cooldown and depressuri; ation to RHR operating conditions are completed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> af ter the accident since there is ample time to complete the

, operations during this time period. The steam rQeases and feedwater flows from 2 to 8 fen::: hr ff*ss b a

S

. A te W W L .sn r D , J L 98 MP3 20 la.

4 z; y q hours were determined for the intact and ruptured steam generators from a mass and 1 energy balance using the conditions at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at the RHR system in-service '

conditions.

Af ter 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, it was assumed that further plant cooldown to cold shutdown as well as long-term cooling is provided by the RHR system. Therefore, the steam releases to the atmosphere were terminated after RHR in-service conditions were assunied to be reached at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

For the time period from initiation of the accident untilleakage termination, the releases were determined from the LOFTTR2 results for the time prior to reactor trip and foliowing reactor trip. Since the condenser is in service until reactor trip, any radioactivity released to the atmo' sphere prior to reactor trip would be through the condenser air ejector. A'+er reactor trip, the releases to the atmosphere were assumed to be via the .'aNo and MSfmBVs The. mass releaseJatet.to the atmosphere from the t.OFTTR2 an ysis are .

presente Figures 15.6.3-10 and 15.6.3-11 for the ruptured and intact ste

~

generators, r pectively, for the time period untilleakage termination. The ma releases calculated fro the time of leakage termination until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and from 2 to 8 ho rs are also assumed to be re ased to the atmosphere via the MSPP.""s. The mass release for the SGTR event for the to 2-hour and 2 to 8-hour time inte is are presented in Table 15.6.3-3.

ogg ., A j g a we/ amp 15.6.3.2.2.2 Offsite Radiation Dose' Analysis --

The evaluation of the radiological quences of a steam generator tube rupture event assumes at the reactor ha en operating at the maximum allowable Technical m

Qg specification imit for p ' ary coolant activity and primary to secondary leakage for i sufficient time to ec blish equilibrium concentrations of radionuclides in the reactor coolant and in the seco ary coolant. Radionuclides from the primary coolant enter the steam N generator vi e ruptured tube and are released to the atmosphere through the MCPPVr d2 77 and tdSERBVs (and safety valves) and via the condenser air ejector exhaust.

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The quantity of radioactivity released to the, environment, due to an SGTR, depends upon .

primary arid secondary coolant activity, iodine spiking effects, primary to secondary break flow, break flow flashing fractions, attenuation of iodine carried by the flash 6d portion of the break flow, partitioning of iodine between the steam generator liquid and steam phases, the mass of fluid released from the generators, and liquid-vapor partitioning in the turbine condenser hot well. All of these parameters were conservatively evaluated in a manner consistent with the recommendations in Standard Review Plan 15.6.3.

1. Design Basis Analytical Assumptions The major assumptions and parameters used in the analysis are itemiz'c'd.in Table 15.6.3-4.

MAY-29-98 FR1 12:03 PM SAFETY ANAL.YSIS BRANCH FAX N0.' 860 832 4784 P. 12 98 MP3 2.0 Q MNPS.3 FSAR '7

2. Source Term Calculations E* Y.b S4w/ QQ The radionuclides concentrations in the primary and secondary system, prior to and following the SGTR, are determined as follows:

a. The iodine concentrations in the reactor coolant will be based upon accident initiated and preaccident lodine spikes. '

l. Accident initiated Spike - The initial primary coolant iodine concentration is 1.0 pCilg of Dose Equivalent (D.E.) 1131. Following the primary system depressurization associated with the SGTR, an lodine spike is inklated in the primary system which increases the iodine release rate from the fuel to the coolant to a value 500 times greater than the release rate corresponding to the initial primary system lodine concentration. -- -

11. Preaccident Spike - A reactor transient has occurred prior to the SGTR and has raised the primary coolant lodine concentration from 1.0 to 60 pCi/ gram of D.E.1-131.

l

b. The initial secondary coolant lodine concentration is 0.1 pCi/ gram of D.E.1131 l

c. The chemical form of lodine in the primary and secondary coolant is assumed to be

' I elemental.

d. The ]nitia)_ noble gas concentrations in the reactor coolant are based u o nt

[ ;' s;ic p ,..-. .-

~

3. Dose Calculations C

dcd hGC C'

• l'd3 _-

The lodine transport model utilized in this analysis was proposed by Postma and TamI ".

The model considers break flow flashing,' droplet size, bubble scrubbing, stea -

partitioning. The model assumes that a fraction of the lodine carried by the break flow becomes airborne immediately due to flaghing and atomization. Removal credit can be .

' taken for scrubbing of iodine contained in the atomized coolant droplets as they rise from the rupture site to the secondary water surface. H6 wever, the ben fit associated with the scrubbing of the steam bubbles has been shown to be relatively insignificant, and has been neglected in this analysis. The fraction of primary coolant iodine which is l not assumed to becgme airboma immediately mixes with the secondary water and is assumed to become airborne at a rate proportional to the steaming rate and the lodine partition coefficient. This analysis conservatively assumes an lodine partition coefficient of 0.01 between the steam generator liquid and steam phases. Droplet -

i 1

• removal by the dryers is conservatively assumed to be negligible. ~

I The following assumptions and paremeters were used to calculate the activity released i to the atmosphere and the offsite doses following an SGTA.

I (1) Postma. A, K., and Tom. P. S., " lodine Behavior in a PWR coling System Following a Postulated Steam Generator Tube Rupture *, NUREG-0409.

L__---___-_-__-__-_-_-___

• ' 98 NPb 2Q D1 i MNPS-3 FSAR I

WCAP-10266 P-A, Revision 2 with Addenda (proprietary), and WCAP-10337-A (i Kabadi, J. N, et al., "The 1981 Version of the Westinghouse)ECC prietary),1987.

Using the BASH Code."

ht-tz.

Qgt WCAP-10444-P-A,1985. Davidsor., S. L., ed., et al., " VANTAGE 5 Fuel Assembly .

Reference Core Report."

WCAP-10444-P-A, Addendum 2,1988, and letter from W. J. Johnson (Westinghouse)

M. W. Hodges (NRC), Letter Number NS-NRC-88-3363,1988," VANTAGE SH Fuel Assembly."

~ _ _ _ _ y - -

N WCAP-13002 proprietary) and WCAP-13003 (non-proprietary),1991, " Margin / pgto Over Analysis for a ' Steam Generator Tube Rupture for Millstone Unit 3 - Four-Loop g Operati l

WCAP;13056 (proprietary) and WCAP-13.057 (non-proprietary),1991, " Margin to Ove (AF$$ nalysis for a Steam Generator Tube Rupture for Millstone Unit 3 - Three-L lb WCAP-10698 (proprietary) and WCAP-10750(non proprietary),1984,"SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill." '

Supplement 1 to WCAP-10698,1985," Evaluation of Off-site Radiation Doses for an SG Accident." '

I WCAP-11002 (proprietary) and WCAP-11003 (non-proprietary),1986, " Evaluation of Steam Generator Overfill due to a SGTR Accident."

)

J. F. Opeka to U.S. Nuclear Regulatory Commission letter B15028, dated December 14

(%'N Nh 1994, " Millstone Nuclear Power Station, Unit No. 3, Proposed Revision to Technical Specifications, Supplemental Collection and Release System."

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}

Event Time (sec)

SG Tube Rupture O Reactor Trip on OTAT"' 109 St Actuated 381 s

Ruptured SG isolated L t S t W S g d v A.

..,mm. 4442 y eest Ruptured SG hteep Fails Open g 1444

' At Ines PkitC D"*f g Ruptured SG leM!MYMWeek Valve Closed -

RCS Cooldown Initiated RCS Cooldown Terminated h p RCS Depressurization Initiated Q {

RCS Depressurization Terminated St Terminated Break Flow Terminated -

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"' The nominal trip setpoint specified in the Technical Specifications is modeled since earlier reactor trip is conservative for this analysis. Although the low pressurizer pressure reactor trip setpoint is not reached the analysis modeled the Technical Specification value with positive uncertainties.

MNPS 3 FSAR 98 MP3 20 h Table 15.6.3 3 Millstone Unit 3 SGTR Analysis Mass Releases Total Mass Flow (Pounds) 0 - 2 HRS 2 - 8 HRS Ruptured SG

- Condenser 117,300 0

- Atmosphere 1 r20

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, , . . . . . . . ~ . . 150,100 0

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MAY-29-98 FRI 12:04 PM SAFETY ANAL.YSIS BRANCH FAX N0. 860 832 4784 P. 13 l

MNPS-3 FSAR 29 /3y Table 15.6.3 4 Millstone Unit 3 SGTR Analysis Parameters Used in Evaluating Radiological Conser,cnces ~

1. Source Data A. Total steam generator tube 1.0 leakage, prior to accident, gpm B. Reactor coolant lodine activity:

1. Accident initiated Spike . .

The initial RC iodine activities based on 1.0 pCi/ gram of D.E.1131 are -[

presented in Table 15.6.3 5. The  !

ledine appearance rates assumed for

• the accident initiated spike are Nu presented in Table 15.6.3 6. Icg'eh p

. , r. [

2. Preaccident Spike Primary coolant lodine activities base

- on 60 pCi/ Oram of D.E.1131 are presented in Table 15.6.3-5.

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3. Noble Gas Activity The initial primary co activities based o presented in Table

. Ahh ra k

g E 3-/.

C. Secondary system initial activity Dose equivsfent of 0.1 pCilgm of 1-131, presented in Table 15.6.3 5.

D. Reactor coolant mass, grams 2.359 x 10' -

E. Initial Steam generator mass 4.316 x 10' (each), grams

• F. Offsite power .

l.ost at time of reactor trip G. Primary to. secondary leakage 8 duration for intact SG, hrs.

H. Species of lodine . 100 percent elemental' .

li

' hY-29-98 FR1 12:04 PM SAFETY ANALYSIS BRANCH FAX N0. 860 832 4784 P. 14 MNPS 3 FSAR y 98 MP3 20 4d Table 15.6.3 7 Millstone Unit 3 SGTR Analysis - - - - -

t. /.

' . . '#g Noble Gas Specific Based on d Activities

%%. in theg Primagoolant'S W 00 yg C ^' &

T gts p Nuclide_ Specific Activity (uCi/om)

Kr-83m 0.130 Kr 85m 0.499 Kt-85 0.00999 .

.. Kr.87 0.358 - - -

Kr-88 0.980 Kr-89 0.0308  !

Xe-131m -

0.00324 '

Xe-133m 0.177 Xe-133 .

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Offsite Radiation Doses Doses (Rem)

• Calculated Allowable Guideline Value Value

1. Accident initiated lodine Spike Exclus, ion Area Boundary (0-2 hr.)

Thyroid h.0 30 Whole Body Gamma l

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Ok 2.5 Thyroid 2.0 30 Whole Body Gamma < 0.1 -

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, Exclusion Area Boundary (0 2 hr.) 5/

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Thyroid 0 300

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1- y Docket No. 50-423 B17288 Attachment 3 Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Revised Steam Generator Tube Rupture Analysis (PLAR 3-98-4)

Description of the Change, Background and Safety Assessment

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June 1998

U.S. Nuclear Regulatory Commission B17288\ Attachment 3\Page 1 j

Description of Chance 1

The proposed revision to the Millstone Unit 3 licensing basis will address a

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recent Steam Generator Tube Rupture (SGTR) analysis. The SGTR analyses described in the FSAR include a offsite dose analysis and a margin to overfill analysis. Both of the analyses have been updated.

Backaround The offsite dose analysis was updated to reflect a larger capacity for the steam generator atmospheric dump valve and the margin to overfill analysis was updated to feflect a new, single failure. With respect to the Steam Generator Tube Rupture margin to overfill, a new single failure was discovered which-is-more limiting than the current analysis. The current analysis used the failure to open on demand of the atmospheric dump bypass valve associated with one of the intact steam generators. Recently it was discovered that there is a . single .

failure which will cause failure to open on demand of the atmospheric dump bypass valve associated with two of the intact steam generators. The dose consequences analysis postulates failure in the open position of the atmospheric dump valve for the steam generator associated with the ruptured steam generator tube. The current analysis used 600,000 lb/hr/ valve at 1140 psia for the flow capacity of the atmospheric dump valve. This is a minimum flow capacity. A flow capacity of 820,000 lb/hr/ valve at 1140 psia has been determined to be the associated maximum capacity. The larger capacity results in a larger release rate to the atmosphere from the failed open atmospheric dump valve.

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The Steam Generator Tube Rupture analysis for margin to overfill and the analysis for dose consequences both needed to be revised due to the errors.

This FSAR change incorporates the results of the revised analyses into the FSAR.

SAFETY ASSESSMENT The change contains the update to the FSAR incorporating the limiting single failure in the most recent Steam Generator Tube Rupture (SGTR) margin to overfill analysis. This limiting single failure results in credit for the,'use of thE atmospheric dump bypass valve associated with only one intact steam generator opening on demand. The current analysis credits the atmospheric dump bypass valve associated with two intact steam generators opening on demand. A larger minimum capacity for the atmospheric dump bypass valve is credited in the revised analysis. This larger capacity is still a conservative minimum capacity.

The Operator actions and their timing credited in the ' analysis is not being

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U.S. Nuclear Regulatory Commission B17288\ Attachment 3\Page 2 changed. The acceptance criterion of no steam generator overfill following an SGTR is still met. .

The change also contains the updates to the FSAR that incorporate the most recent analysis performed for determining the design basis dose consequences of a Steam Generator Tube Rupture. Tha updated dose analysis, as well as the current FSAR analysis, postulates the failure in the open position of the steam generator atmospheric dump valve associated with the steam generator with the ruptured tube. Revising the analysis does not impact the failure probability of the steam generator atmospheric dump valve. The SGTR analysis credits closure of the atmospheric dump valve block valve to isolate.ibe failed open.

atmospheric dump valve. The revised SGTR dose analysis uses a larger flow rate through the valve. The higher flow is within the capabilities of the block valve to close and, therefore, the failure probability of the block valve to close is not increased by the change.

The revised dose analysis uses a larger flow capacity for the failed open atmospheric dump valve as well as a shorter credited time for closure of the block valve. The faster isolation time,20 minutes instead of 30 minutes, more than compensates for the larger capacity assumed for the atmospheric dump valve. Therefore, the revised dose analysis does not increase the consequences of an SGTR. In addition, changing the margin to overfill and dose analyses and their descriptions can not cause a steam generator tube rupture.

The change is to the analyses and FSAR description of thataanalyses. The --.~m change to the margin to overfill analysis credits a larger minimum atmospheric dump bypass valve capacity. The important changes in the dose analysis are the increased capacity of the atmospheric dump valve and the shorter time utilized for isolation of the failed open atmospheric dump valve. The only change in equipment credited in the analyses is the crediting of the block valve to close when there is a larger flow through the valve. The block valve can close under the postulated accident conditions. Therefore, the change does not create the possibility of an accident of a different type.

The revised dose analysis reduces the time available to the Operators to isolate the failed open atmospheric dump valve from 30 minutes to 20 minutes. The actions required are unchanged. The twenty minutes allows sufficient time for the Operators to both recognize the failure of the atmospheric dump valve and to close the block valve. However, reducing the available time to the Operators from 30 minutes to 20 minutes represents a reduction in the margin for error

U.S. Nuclear Regulatory Commission B17288\ Attachment 3\Page 3 available to the Operators and, thus, represents a reduction in the margin of safety. This means that the change is considered an Unreviewed Safety Question.

The only adverse impact of the change is this small reduction in the margin of error available to the Operators. This reduced time results in a small increase in the probability that the Operators will fail to mitigate an SGTR. However, the change results in a decrease in the consequences of an SGTR. The decrease in consequences is judged to offset the increase in probability and, therefore, the change is safe.

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Docket No. 50-423 -

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l Millstone Nuclear Power Station, Unit No. 3 Proposed License Amendment Request Revised Steam Generator Tube Rupture Analysis (PLAR 3-98-4)

Significant Hazards Consideration and Environmental Considerations l-l i

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U.S. Nuclxr Regulatory Commission B17288\ Attachment 4\Page 1 Significant Hazards Consideration NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC).

The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does not involve an SHC because the revision would not:

1. Involve a significant increase in the probability or consequence of an accident previously evaluated.

The FSAR Steam Generator Tube Rupture offsite dose analysis is being

., m updated.lo.raflect a larger capacity for the steam generator atmospheric dump

- valve. ~ The updated analysis, as well as the current FSAR analysis, postulate the failure, in the open position, of the steam generator atmospheric dump valve associated .with the steam generator with the ruptured tube. Revising the analyses does not impact the failure probability of the steam generator

atmospheric dump valve; The SGTR analyses credit closure of the atmospheric dump valve block valve to isolate the failed open atmospheric dump valve. - The revised SGTR analysis uses a larger flow capacity for the atmospheric dump

! valve. A larger flow capacity, without other changes being made, would increase the consequences associated with this failure. However, the time credited for closure of the block valve is being reduced to 20 minutes after the atmospheric dump valve fails open, instead of 30 minutes after the atmospheric dump valve fails open. A shorter isolation time, without other changes being made, would decrease the consequences associated with the atmospheric dump fail failing open. This faster isolation time more than compensates for the larger capacity assumed for the atmospheric dump valve. Therefore, the revised analyses does not~ increase the consequences of a Steam Generator Tube Rupture. The "

l: change is a revision to the analyses for a steam generator tube rupture and the description of the analyses in the FSAR. Changing the analyses and its description can not cause an increase in the probability of a steam generator tube rupture.

Therefore, the proposed revision does not involve a significant increase in the

.. probability or consequence of an accident previously evaluated.

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. Create the possibility of a new or different kind of accident from any accident previously evaluated.

The change is to the analyses and FSAR description of that analyses. The important changes in the analyses are the increased capacity of the atmospheric dump valve and the shorter time utilized for isolation of the failed open atmospheric dump valve. The only change in equipment credited in the

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,, U.S. Nuclear Regulatory Commission B17288%ttachment 4\Page 2 analyses is the crediting of the block valve to close when there is a larger flow through the valve. The block valve can close under the postulated accident conditions.

Therefore, the proposed revision does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

I, ..alve a significant reduction in a margin of safety.

The revised analyses reduces the time available to the Operators to isolate the failed open atmospheric dump valve from 30 minutes to 20 minutes. The actions required are unchanged. The twenty minutes allows sufficient time for the Operators to both recognize the failure of the atmospheric dump valve and to close the block valve. However, reducing the available time to the Operators from 30 minutes to 20 minutes represents a reduction in the margin for error available to the Operators and thus represents a reduction in the margin of safety. The reduction in the margin of safety is not~significant since the twenty minutes allowed by the analysis is still significantly above the typical ten minute minimum assumed response time for Operator actions performed in the control room. In addition, Operator training provides assurance that the twenty minute time limit is met.

Therefore, the proposed revision does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria of 10CFR51.22 for environmental considerations. The proposed revision does not involve an SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures.

Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusigp from the requirements for environmental review.

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