ML20217G509

Forwards RAI Re Revised SG Tube Rupture Analysis Removal of SG Repair Methodologies & Restoration of Previous Dose Equivalent Iodine Limits for Mentioned Plants

ML20217G509
Person / Time
Site:	Byron, Braidwood
Issue date:	10/03/1997
From:	Dick G NRC (Affiliation Not Assigned)
To:	Johnson I COMMONWEALTH EDISON CO.
References
TAC-M97315, TAC-M97316, TAC-M97317, TAC-M97318, TAC-M98070, TAC-M98071, TAC-M98072, TAC-M98073, NUDOCS 9710100274
Download: ML20217G509 (15)

Text

v rjc-Wlff5 lc

,p

f* k. UNITED STATES

^

.<;p.qg,lyS1 g

] NUCLEAR REGULATORY COMMISSION o  : WASHINGTON, D.C. seseH001

1.  %*****,

' October 3, 1997 4

[ .

LMs.' Irene M. Johnson, Acting Manager F

Nuclear Regulatory Services Commonwealth Edison Company i Executive Towers West III

1400 Opus Place, Suite 500

[ Downers Grove, IL 60515

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE REVISED STEAM-GENERATOR TUBE RUPTURE ANALYSIS; REMOVAL OF STEAM GENERATOR REPAIR. )

METHODOLOGIES; AND RESTORATION OF PREVIOUS DOSE EQUIVALENT IODINE l LIMITS - BYRON AND BRAIDWOOD STATIONS (TAC NOS M97315, M97316r 1 j M97317,M97318,M98070,M98071,'M98072ANDM98073) j i

i

Dear Ms. Johnson:

1

. On November 13, 1996, Commonwealth Edison Company. (Comed) submitted its-

revised steam generator tube rupture (SGTR) analysis for Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2. By letter dated r February 28, 1997, Comed also submitted a request for a. license amendment

to-revise the technical specifications regarding steam

! methodology and primary coolant dose equivalent iodine 3evel.generator repair Both submittals i-are-related to the replacement of the steam generators for Byron,, Unit 1, and ,

F Braidwood,: Unit 1. We issued requests for additional information (RAI)

-regarding the SGTR analysis on February ll, May 20 and July 18, 1997. Comed provided its_ responses to the RAls on March 20,_ June 24 and August 19,il997.

I During t..e course of our review, we have identified the need for further

! information for both of the subject requests.in order to permit us to perform i confirmatory calculations of the potential radioactive release in the event of an accident. The-information requested in the enclosed RAI-applies to both-

l. Byron, Unit 1, and Braidwood,-Unit 1. If there are differences in the answers-4

-for' Byron and Braidwood, provide the Byron information first, since the steam

- generators will be replaced first,- and indicate when we may expect to receive-

[ the comparable information for-Braidwood.

/

/

c 0 1721 f' R.;30 lllll!II.IIlli!I,1ll l

L e

0 ju188umgga PDR EC FRE CB!TB C8PY

'~

_. ._ _ :~~~ ~T:T ____

I ' -'

, I. Johnson .

- Please' provide your response to the request so that we may continue our review of your_subelttals.

Sincerely,_ l Georg . Dick, Jr., Senior Project Manager Project Directorate III-2

- Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456, STN 50-457

Enclosure:

RAI cc w/ enc 1: see next page 9

-a+mem- .m_ .-

I. Johnson - 2'-

Please-provide your response to the request so that we may continde our review of your submittals.

Sincerely, '

Original sigred by: ,

George F. Dick, Jr., Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket Nos. STN 50-4'54, STN 50-45$.

STN 50-456, STN 50-457-

Enclosure:

- RAI cc w/ enc 1: see next page Distribution:

Docket File

, PUBLIC-PD3-2 r/f E. Adensam, EGAl R. Capra G. Dick C. Moore C. Miller J. Hayes R. Emch S. Bailey D. Lynch T. Harris (e-mail only TLH3)

OGC, 015B18

.ACRS, T2E26 R. Lanksbury, RIII.

DOCUMENT NAME: G:\CHNTJR\ BRAID-BY\BB97315.RAI To receive e copy of this document, indi f ate in the box: "C" = Copy without enclosures "E" = Copy with enetosures "u" = No copy _ [ \ i l0FFICE PM:PD3-21 J\ l /E- A.A:PD3/2- l6? ERB . If D:PD3-2 , l F, lNAME GDICK 2P4R tM00RE ( J$ MILLER N M 1 RCAPRA 6] #w lDATE

. 89/ pf97 09/4/97 // W/A/97 W/cs/97

/v 0FFICIAL RECORD COPY l0

6

(

• I. Johnson .

Please provide your response to the request so that we may continue our review of your subrittals.

Sincerely, Original signed by:

George F. Dick, Jr., Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket Nos. STN 50-454, STN 50-455, STN 50-456, STN 50-457

Enclosure:

RAI cc w/ enc 1: see next page l

Distribution:

Docket File PUBLIC PD3-2 r/f E. Adensam, EGAl l

R. Capra G. Dick C. Moore C. Miller J. Hayes R. Emch S. Bailey D. Lynch T. Harris (e-mail only TLH3)

OGC, 015B18 ACRS. T2E26 R. Lanksbury, RIII DOCUMENT NAME: G:\CHNTJR\ BRAID-BY\BB97315.RAI To receive a copy of this document, Indise(e in the tems *C" = Copy without enclosures "Em a Copy with enclosures awa = No copy. [ T 1 0FFICE PM:PD3-FJ A l /E LA:P43/2 lDTERB .lf D:PD3-2 , l B-j!AME GDICK //M tBDORE ( #SNILLER N 2 C RCAPRA$]#w DATE EJ/ g(97 ~ 09/'@ 97 // 'f9/;2-/97 ' {G/cS/97

/v 0FFICJAL R; CORD COPY to

. I. Johnson Byron /Braidwood Power Stations Commonwealth Edison Company CCi Mr. William P.-Poirier,= Director George L. Edgar Westinghouse: Electric Corporation Morgan, Lewis and Bochius Energy Systems Business Unit 1800 M n reet, N.W.

r- Post Office Box 355, Bay 236 West- Washington, DC 20036 Pittsburgh, Pennsylvania 15230 Joseph Gallo 500 South Second Street ,

Gallo & Ross Springfield, Illinois 62701 '

1250 Eye St., N.W.

Suite 302 EIS Review Coordinator Washington,-DC 20005 U.S. Environmental Protection Agency 77 W. Jackson Blvd.-

l Michael 1. Miller, Esqu Pe Chicago, Illinois- 60604-3590 Sidley and-Austin One First National Plaza Illinois Department of Chicago, Illinois 60603 Nuclear Safety '

Office of Nuclear Facility Safety Howard A.-Learner 1035 Outer Park Drive-Environmental law and Policy Springfield, Illinois 62704 Center of the Midwest 203 North LaSalle Stteet Commonwealth Edison Company-Suite 1390 Byron Station Manager Chicago, Illinois 60601 4450 North German Church Road Byron-1111ncis 61010 U.S. Nuclear Regulatory Commission Byron Resident inspectors Office Kenneth Graesser, Site Vice President

-4448 North G9rman Church Road Byron Station Byron, Illinois 61010-9750 Commonwealth Edison Station 4450 N. German Church Road Regional Administrator, Region III Byron, Illinois 61010-U.S. Nuclear Regulatory Commission 801'Warrenville Road U.S. Nuclear Regulatory Commission Lisle, Illinois -60532-4351 Braidwood Resident Inspectors Office Rural Route #1, Box 79 Ms. Lorraint Creek Bracev111e, Illinois 60407 Rt. 1. Box 182

• Manteno, Illinois 60950 Mr. Ron Stephens Illinois Emergency Services Chairman, Ogle County Board and Disaster Agency Post Office Boa 357 110 East Adams Street Oregon, 1111r.01s 61061 Springfield, Illinois 62706, Mrs. Phillip B. Johnson Chairman 1907 Stratford Lane Will County Board of Su>ervisors Rockford, Illinois C1107- Will County Board Court 1ouse

- Joliet, Illinois 60434

Commonwealth Edison Company

'Braidwood Station Manager Rt. 1. Box 84 Braceville, Illinois 60407 Ms. Bridget Little Rorem

. Appleseed Coordinator 117 North Linden Street Essex, Illinois 60935-Document Control Desk-Licensing-

, Commonwealth Edison Company.

4

.1400 Opus Place, Suite 400 Downers Grove, Illinois 60515 Mr. H. G. Stanley.

Site-Vice President

. Braidwood Station Commonwealth Edison Company '

RR 1, Box 84 Bracev111e, IL- 60407.

4 7

1

)

s, .m . , _ _

. RE0 VEST FOR ADDITIONAL INFORMATION'

B[y1 SED STEAM GENERATOR TUBE RUPTURE ANALYSIS AND

. RESTORATION OF .REVIOUS DOSE EQUIVALENT IODINE LIMIT 4

COMMONWEALTH EDISON COMPANY BYRON STATION. UNITS 1 AND 2 BRAIDWOOD STATION. UNITS 1 ANf' 2 DOCKET NOS. STN 50-454. STN 50-455. STN 50-456 AND STN 50-152-1 The Byron and Braidwood submittals of the revised : team generator tube rupture (SGTR) analysis and the amendment request to iner)ase the reactor coolant system (RCS) activity of dose equivalent ggI are under review. Both submittals were made to support the replacement of the steam generators (SG) at Byron, Unit 1, and Braidwood, Unit 1. As a result of th;; review, it ha.

4 been determined tnat there exists an insufficient amount of information to permit-the staff to perform confirmatory calculations of the proposed actions. .

- In order to complete these_ actions, the licensee is requested to provide

- information sufficient for the staff to model the conveyance and the release of radioactivity for the SGTR, main steamline break (MSLB), rod ejection and locked rotor accidents. If the replacement SG have no im)act at all on the -

releases of radioactivity to the environ.nent for any of tiese accidents, then data need not be provided for the accident involved. The licensee should supply the following information and any additional information that is

! necessary for the staff to accurately model the respense of the replacement

SG. For each of the accidents provide a time line for those aspects of the event relevant to the determination of releases to the environment.

1. For the MSLB accident, provide the following. information:

a. Mass of liquid released from the faulted SG as a function of time.

b. -Mass of steam: released from the intact SG as a function of time. As a minimum, releases should be designated as those within two hours and those-after two hours. '

c. ! Flashing fraction for primary to secondary leakage into the intact SG.

d. Scrubbing fraction for flashed portion of primary to secondary leakage into the intact SG.

1

e. Primary bypass fraction (liquid entrained in th'e flashing fraction) for intact SG.

f. Time to isolate faulted SG.

g. Duration of plant cooldown by the secondary side.

h. Additional information which should be provided .is contained in Attachment I and Attachment 2.

i ENCWSURE l

L. _ _ - - _ .-

s

=. '

a

]

-2. For the SGTR accident, provide the following information:

a. Mass of liquid and steam released from the faulted SG as a function of

_ time;.As a minimes,' releases should be designated as though within two ,

hours and those after two hours.

l p 4

b. Mass of steam released from the intact SG es a function of time. As a '

i minimum, releases >hculd be designated as tfough within two hours and i i those after two hours.

l

. c. Flashing fraction-in the intact and faulted SG. i i d. Scrubbing fraction in the intact and faulted Sli.

i

. e. Primary _ bypass fraction for the intact and faulted SG.

f f. Time to isolate faulted SG.

g. Duration of plant cooldown by the secondary side,

h. Primary to secondary release rate from the ruptured tube as a function-l of time. *

1. ' Indicate if overfill conditions do or do not exist. If.they do exist,

!- appro)riate mass release data should be provided as a function of-time for tie faulted SG.

! 'j. Additional-information which should be provided is contained in Attachment 3.

} ' 3. For the locked rotor accident, provide the following information:

F L a. Liquid release from the SG as a function of time.

+

j. b .' Duration of plant cooldown by the secondary side.

i

c. A description of how the primary to secondary releases were modeled as releases to the environment.

I

d. Fraction of fuel rods experiencing cladding perforation and/or fuel melting.

4. For the rod. ejection accident, provide the following information:

a. The fraction of the fuel rods which have their cladding breached as a result of this accident.-

b. The fraction of the-fuel rods which reach or exceed the initiation j . temperature for fuel melting as a result of thit accident.

_c. For the release via the primary to secondary leakage pathway, a description of the assumptions which were utilized in the release of such activity.

. d. For the release via the containment pathway, a descri) tion of the assumptions which were utilized in the release of suc1 activity.

j._, , .

.-. ~ __

INPlii PARAMETERS FOR EVALUATION 0F MAIN STEAMLINE BREAK ACCIDENT

- -1. Primary coolant concentration for technigl specification's (TS) maximum instantaneous value_ for dose equivalent I.

Pre-existino Soike Value (uti/a) 131g ,

'32 -

1 133g ,

1M I .

issy ,

2. Volume of primary coolant and secondary coolant.

3 Primary Primary CoolantCoolant Volume (ft( ) 'F)

Temperature I

i Secondary Secondary Coolant Coolant Steam Liquid Volume Volume (ft (ft {-)

Secondary Secondary Coolant Coolant Steam Feedwater Temperature Temperature ( ('F) 'F)

3. TS limits for DE '3'I in the primary and secondary coolant.

Primary Coolant DE "'I concentration 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> TS value (fi/g)

, Secondary Coolant DE I concentration (#Ci/g)

4. TS value for the primary to secwdary leak rate.

Primary to secondary leak rate, any SG (gpd)

Primary to secondary leak rate, total all SG (gpd)

5. Iodine Partition Factor Faulted SG 1 Intact SG 0.1 Primary to Secondary Leakage 1.0

6. Steam Released to the environment Faulted SG (lbs) 0-2 hours

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Intact SG (lbs) 0-2 hours

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

7. Letdown Flow Rate (gpm)

ATTACHMENT 1

~ _ . . . . . _ _ _ _ _ _

8. Release Rate for 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> TS of Dose Equivalent *!

D .

my .

mg .. ,

mg__,-

mi

mg ,

4 I

9. Atmospheric _ Dispersion Factors EAB (0-2 hours)

LPZ (0-8 hours)

Control Room (0-8 hours) 4

10. Control Room I Volume (ft )

3 '

Normal Makeu) Flow (cfm)

.' Emergency Ma(eup-Flow (cfm) 1 Makeup Filter efficiency (%):

Unfiltered Inleakage (cfm)

Recirculation Filter Flow Rate (cfm)

Recirculation Filter Efficiency (%)

4 1

l 4

w-w., ~ ,

• m som emm w ww-=*e<< = ~ ~

. MAIN STEAMLINE BREAK-THYROID DOSE ASSESSMENT Pre-existina Soike EAS LEZ Control Room Calculated doses (rem)

Regulatory Guidelines 30 30 30

- (rem)

Accident Initiated Soike

. E6B LEZ Control Room Calculated doses (rem) ,

Regulatory Guidelines 30 30 30 (rem) l l

4 l

ATTACHMENT 2 l

4 c.

. INPUT PARAMETERS FOR EVALUATION OF SGTR

-.- 1. Primary coolant-concentration of $35]R for dose equivalent 13'I .

4 Pre-existino Soike Value fuci/a) 1 131g ,

r 13a; ,

mg ,  ;

• I =

i 13sg ,-

i 2.- Volume of primary coolant and secondary coolant. ,

3 l

i Primary Primary Coolant Coolant Volume (ft( ) 'F)

Temperature '

! Secondary Coolant Steam Volume Total (ft3)

!- Secondary Coolant Mass Total (1bs)

, - Primary Coolant. Pressure (psia) i PrimaryCoolantMass(})bs)

Pressurizer Volume (ft Pressurizer Temperature (*F)

L Pressurizer Pressure (psia)

Secondary Coolant Liquid Mass /SG (Ibs) y Secondary Coolant Steam Mass /SG (lbs) i Secondary Steam Temperature ('F)

! Secondary Liquid Temperature ( F)

! 3. TS Limits for DE '3'I in the primary and secondary coolants:

Maximum Instantaneous in primary coolant (pCi/g)

48 Hour DE in primary coolant-(pCi/g)-

l Secondary Coolant (pCi/g) 4.. TS value for the primary to secondary leak rate (include reference

. temperatureand-pressure):

L Any SG (gpd)

Total. all SG (gpm) l S. Primary coolant activity (Ci) due to a pre-existing spike:

l *I.

! 1 sag ,

• I.

, *I.

135 j 1 1

j

-6. Primary coolant activity, levels (pCi/g) for accident initiated spike.

i-131} ,

• I -

[ . my ,

>- *I .

- 135; ,

1 I ATTACHMENT 3

.. n -_-- ~ . _ . . . - . - - - --

w .- paw .- -tr - w v. my --

- - 2' '

7. . Primary coognt concentration at maximum instantaneous value of dose equivalent I.

131g ,

tsay ,

tssg ,

tug .-

tssy -

8. Primary Coolant Activity (C1)-for Accident Initiated Spike 1"et g 13rg issg u4g

)- "'I

9. Iodine Partition Factor Faulted SG Intact SG Condenser

10. Steam Released to the environment as k function of time:

Faulted SG 0-2 hours

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Intact SG 0-2 hours

> 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />

11. Letdown Flow Rate (gpm)

. 12.-Atmospheric Dispersion Factors

EAB (0-2 hours)

LPZ (0-8 hours)

, Control Room (0-8 hours)

13. Control Room:

Emergency Makeup Flow (cfm)

Makeup Filter efficiency (%)

' Unfiltered In'ieakage (cfa) '

Recirculation Filter Flow Rate (cfm)

Recirculation Filter Efficiency (%)

Occupancy Factor (0-1 day) t d

4.- s~, -- , . - . _ , ;-. . . . _ .

~. _ . . _ _ - . _ . . - - . . . _ . _ _ _ . _ . _ _ _ _ _ . . . _ . _ . . _ . _ _ _ _ . .

5

.. -3

14. For.the Accident Initiated Spike Case-Release Rate (C1/hr) 500X Release Rate (Ci/hr) uty tug ing tug usg

15. Flashing Fraction, Primary Bypass and Scrubbing fraction as a function of time.

15. Mass release rate through the ruptured tube as a function of time, h .

4 I

i j

i o

- w ' h~ e a n4-rw,*-- . 4 w %u,., % -

sw-- y- c w -- ~ ~ - p

7 . STEAM GENERATOR TUBE RUPTURE THYROID DOSE ASSESSMENT Case Involvina Pre-existina Soike 168- W Control RQ25

- Calculated thyroid dose (rem)

Regulatory Limits (rem) 300 300 30 Case Involvina Accident Initiated Soike.

IAH- W Control Room Calculated thyroid dose (rem)

Regulatory Guidelines (rem) 30 30 30

)

ATTACHMENT 4

. . - - - . - - .