ML20212A920

Forwards Addl Info Re 970515 Amend Request Which Proposed Changes to Section 4.12 of TS That Would Allow Use of voltage-based SG Tube Repair Criteria Per NRC RAI

ML20212A920
Person / Time
Site:	Prairie Island
Issue date:	10/20/1997
From:	Sorensen J NORTHERN STATES POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20212A923	List: ML20212A920 ML20212B794
References
TAC-M98944, TAC-M98945, NUDOCS 9710270087
Download: ML20212A920 (38)

Text

. _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _

Northern States Power Company Pralrle Island Nuclear oenerating Plant 1717 Wakonado Dr. East Welch. Minnesota L'm October 20,1997 U S Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos. 50-282 License Nos. DPR-42 50-306 DPR-60 Response to Request for Additional Information Related to License Amendment Request Dated May 15,1997 incorporation of Voltage Based Steam Generator Tube Repair Criteria l (TAC Nos. M98944 and M98945) l The NRC Staff has requested additionalinformation regarding our License Amendment Request Dated May 15,1997 which proposed changes to Section 4.12 of the Prairie Island Technical Specifications that would allow the use of voltage based steam generator tube repair critoria. Our response to the NRC Staff request for additional information is provided in the attachments to this letter.

During the preparation of the response to Question 1 in Enclosure 1, Westinghouse identified an inconsistency in their dose calculation. The dose calculation inadvertently used 0.72 pCl/gm dose equivalent 1131 for the initial coolant activity instead of the Technical Specification value of 1 pCl/gm dose equivalent I 131. The methodology, as described in Enclosure 2, used to perform the dose assessment was not changed by the calculational error described above. The dose calculation has been subsequently revised utilizing 1.0 peilgm dose equivalent lodine. The responses to Questions 1,2 and 3 and the re evaluation of the Westinghouse main steam line break allowable leak rate analysis (Enclosure 2) reflect the revised calculation.

Additionally, as a result of discussions with the NRC Staff, the dose calculation was revised to change the duration of the accident initiated iodine spike from 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The responses to Questions 1,2 and 3 and the re-evaluation of the Westinghouse main steam line break allowable leak rate analysis (Enclosure 2) reflect the revised calculation.

9710270087 971020 DR ADOCK 05000P 2 ( i t. k !

ILl:1,l[lhlOlllylijlll

a USNRC NORTHERN 8TATES POWER C2MPANY

~

October 20,1997 Page 2 i

+

The revisions to the dose calculation described above resulted in lower allowable l primary to secondary leakage values. The new allowable primary to secondary side 4

leakage is 1.22 GPM basou on a density of 62.4 pounds per cubic foot which is 1.66 i GPM at operating temperature (567.3 degrees F and 2250 psla). - All previous values i for allowable primary to secondary leakage provided in the May 15,1997 License

. Amendment Request and the August 29,1997 response to a request for additional

! Information, should be changed to the values specified above.

In this letter we have made no new Nuclear Regulatory Commission commitments.

Please contact Gene Eckholt (612-388-1121)if you have any questions related to this i

response.

j

/ oel J P Sorensen s,~

- Plant Manager -

l Prairie Island Nuclear Generating Plant 1

i c: Regional Administrator - Region Ill, NRC

Senior Resident inspector, NRC

NRR Project Manager, NRC J E Silberg f

Enclosures:

I Affidavit Enclosure 1 - Response to NRC Request For Additional Information Enclosure 2 - Re Evaluation of Westinghouse Main Steam Line Break Allowable Leak Rate Analysis Enclosure 3 - Section 2.2 of Design Basis Accident Radiological Study for the Prairie Island Control Room

4 UNITED STATES NUCLEAR REGULATORY COMMISSION NORTHERN STATES POWER COMPANY PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NO. 50-282 50-306 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION RELATED TO REQUEST FOR AMENDMENT TO OPERATING LICENSES DPR-42 & DPR-60 LICENSE AMENDMENT REQUEST DATED May 15,1997 Northern States Power Company, a Minnesota corporation, by this letter dated October 20,1997, with Enclosures 1,2 and 3 provides a response to an NRC Staff request for additionalinformation in support of the subject License Amendment Request dated November 27,1996. Enclosure 1 provides the response to the NRC Staff Request for Additional Information. Enclosure 2 provides a re-evaluation of the Westinghouse main steam line break allowable leak rate analysis. Enclosure 3 provides Section 2.2 of the design basis accident radiological study for the Prairie Island control room.

This letter contains no restricted or other defense information.

NORTHERN STATES POWER COMPANY By ttA14v J6til P Sort nsen' Plant Manager Prairie Island Nuclear Generating Plant On thisk y of ' /ff'/before me a notary public in and for said County, personally appeared Joel P Sorensen, Plant Manager, Prairie Island Nuclear Generating Plant, and being first duly sworn acknowledged that he is authorized to execute this document on behalf of Northern States Power Company, that he knows the contents thereof, and that to the best of his knowledge, information, and belief the l statements made in it are true and t it is not interposed for delay.

2Ad .

) ,

y M- MV

\

I 140TARfPU0004 M E90R HtHNEMCouNTY ,

3.WNDEW L

y>%~,----e-r-,- ,,- - -- y - em-------,-e.- - e r - - - _ - ---er--r-----

ATTACHMENT 1 l l

Response to NRC Reauest For AdditionalInformation The following information is provided in response to an NRC Staff Request for Additional Information regarding our License Amendment Request Dated May 15,1997 which proposed changes to Section 4.12 of the Prairie island Technical Specifications j that would allow the use of voltage based steam generator tube repair criteria:

F.ESPONSE TO NRC STAFF QUESTIONS 1

, Question 1:

Provide a summary ol' how the dose calculation was done.

Response

A summary of the dose calculation is provided in Attachment A (Calculation Note Summary) of Enclosure 2 to this submittal.

Question 2:

On page 2 of Attachment 2,"The duration of the accident initiated lodine spike is 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. What is going on during this 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />? Is the lodine value constant or changing during this 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and at what value? Why is 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> used instead of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />?

1

Response

As a result of discussions with the NRC Staff, the duration of the accident initiated iodine spike has been changed to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. As described in the attached Westinghouse Calculation Note Summary (Attachment A to Enclosure 2), the lodine appearance rate is now terminated at 8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

Question 3:

Does the 109,155 lbm on Page 4 of Attachment 1 include liquid and/or steam?

Response

The 109,155 lbm secondary side liquid volume was a conservative estimate of the liquid volume / mass. The best estimate masses in the steam generator at 100% power are 107,000 lbm liquid and 5700 lbm steam. The value used bounds the actual best estimate mass in the steam generator since the initial iodine remains in the liquid space i

and not in the steam space. A partitioning factor of 0.01 could be applied to the steam I

,r.. - - - , . _ , -.

, ,--e , , , - - - -- r - , c -

r

= - - - . - . -_ _ - _ - . _ - - - .. __ - - - _ _ _. - - -

~

space. An initial secondary activity of 0.1 pCilgm is applied to the 109,155 lbm mass in the analysis.

Question 4:

On page 2 of Attachment 2

Does any release occur from the faulted steam  ;

generator after the Initial 15 minute boll off and, if so, what is the value?  !

Rosbones:

Release cont' 1s from the faulted steam generator at the allowable main steam line break leak rate u. 4 led from this calculation. The allowable primary to secondary side leakage is 1.22 GPM based on a density of G2.4 pounds per cubic foot which is 1.66 d

GPM at operating temperature (567.3 degrees F and 2250 psla). This release occurs for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after the event begins. All of the iodine in the primary coolant leaking through the steam generator tubes is assumed to be released directly to the environment without any partitioning.

Question 5:

Attachment 2 Table 2: Please describe the various Modes of operation of the 4 control room ventilation system, or provide reference to the PI USAR.

What are the individual flow rates in Modes 1 and 2? Does the 12,000 SCFM flow rate occur in both modes? Clarify that Mode 2 recirculation of 3600 cfm is filtered air.

Response

The Control Room Ventilation System is described in Section 10.3.3 of the Prairie

. Island USAR. There are two separate, t,ut interconnected systems - the normal ventilation system and the Control Room cleanup system. High radiation detected by the Control Room Radiation Monitors R-23/24 or a safety injection signal on either unit will start the Control Room clean up fans which are interlocked to also start the associated train Control Room Air Handler When the Control Room Clean-up fans are started, the outside air dampers are automatically closed and the 100% recirculation mode is suiomatically initiated. The following information provides more detail:

Control, Relay, and Computer Room Ventilation 1

The Control, Relay, and Computer Room Ventilation System, shown in the attached

- figures derived from USAR Figure 10.3-6, consists of two redundant ventilation trains with each train containing an air handling unit, a humidifier, a filter unit, and a cleanup fan Each train supplies one end of the control room distribution duct. Only one train is normally operated. Ductwork is also provided to direct air from the control room distribution duct to the Relay Room and to the Computer Room. A return fan in the Relay Room returns the ventilation air to the Control Room.

---

e __,w ,, , , -

, ,r,,-- -

l The Control, Relay, and Computer Room Ventilation System is a recirculation type system. The air handling unit in each train draws air from the Control Room with a small amount of makeup air supplied itom the outside atmosphere. The air flows through the air handling unit where it is filtered and cooled. Then a portion of the recirculating air is diverted to the humidifier which adds moisture and returns the air to the recirculation path. A minimum of one train is in operation at all times.

The control room air handler is operated either remotely from the control board using a

" Start Normal-Stop" control sw;tch or locally using a " Local-Remote" selector switch and " Start Stop" push buttons. The remote control switch also has a PULLOUT position to prevent air handler operation. The PULLOUT position is bypassed if the selector switch is placed in LOCAL.

With the selector switch in REMOTE, the air handler is started by placing the control board switch in START. The unit also starts automatically when the control room cleanup fan is started.

Starting the air handler opens the following dampers:

• The air handler discharge damper.

• The air handler outside air supply if no Control Room area high activity alarm condition exists.

• The air handler outside air roof damper if no Control Room area high activity alarm condition exists and if the damper control switch is not in the CLOSE position.

If the air handler was started automatically by cleanup fan operation, the air handler automatically stops when the cleanup fan is stopped.

Control Room Cleanup System Two control room cleanup fans are provided in the Control, Relay, and Computer Room Ventilation System, one for each train. The cleanup fan draws air through its associated particulate, absolute, charcoal (PAC) filter unit and discharges to the train's associated air handling unit. The fan can draw air either from the control room recirculation header or from the outside air supply header upstream of the air handler supply damper. Drawing air from outside requires operator action to realign the system. The automatic safety related lineup isolates the control room from outside air.

The cleanup fans are remotely operated using a " Start Normal-Stop" switch on the control board. The switch spring-returns to NORMAL and has a PULLOUT position to prevent fan operation. A cleanup fan can be started eithy manually by placing the control board switch in the START position or automatically by Control Room radiation monitor high activity or by a safety injection signal from either unit. The cleanup fans are interlocked such that the following actions occur on a fan startup:

. The associated control room air handler starts.

1 o The cssociat:d ci:cnup f n discharge damper opens.

, e The associated PAC filter supply damper from the recirculation h4ader opens.  ;

e The air handler outside air supply damper closes I e The control room outside air roof isolation damper closes.

. The control room exhaust steam exclusion control damper closes.

- The cleanup fans are stopped by placing the associated control switch in STOP. If the fan start was Initiated by an automatic signal, this signal must be cleared or the fan restarts when the control switch spring returns to NORMAL. When the cleanup fan is stopped, the associated outside air roof Isolation damper and PAC filter outside air supply damper receive close signals Placing the control switch in PULLOUT prevents fan restart if the auto start signal still exists.

in Mode 1 operation (the figure labeled NORMAL OPERATION),2000 cfm (including the 165 cfm of unfiltered air inleakage) of fresh outside air, mixed with 10,000 cfm of recirculated air is supplied to the control room by one of the Control Room air handling units, in Mode 2 operation (the figure labeled Si or CRM HIGH RADIATION), a total of 12,000 cfm of air is still recirculated by the Control Room air handling units and includes a minimum of 3600 cfm filtered air. Part of the remaining unfiltered air is assumed to be

the 165 cfm of unfiltered air inleakage coming into the control room. The Control Room cleanup fen takes a suction on the return duct from the Control Room and discharges to the inlet plenum of the air handling unit. The analysis assumes only one train is operating.

A third mode of operation (the figure labeled POST ACCIDENT OPTION), provides the ability to bring in filtered fresh outside air while the cleanup fans are operating and is done under the direction of the Technical Support Center. This option is not analyzed since it requires deliberate operator action following evaluation of the radiation hazards.

The total control room volume includes the control room, the computer room, the relay room, and the two ventilation equipment rooms.

Question 6: ,

How do the filter efficiencies used in the calculation compare to the TS values?

Response

Technical Specification 4.14.B requires 299% removal for particulates, 299% removal for halogenated hydrocarbons, and 290% removal for radioactive methyl iodide.

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

In the calculation, the chemical form of iodine isotopes was assumed to be 100%

elemental lodine. Elemental lodine removal efficiency is the same as the Technical Specification value of 90%

The particulate value of 95%, assumed in the calculation, is more conservative than the Technical Specification value of 99% and was used to be consistent with a previous LOCA dose analysis.

Question 7:

Provide the bases for the X/Q value assumed in the control room dose calculation.

Response

The methodology for determination of the X/Q value used in the control room dose calculation is described in a letter from NSP to the NRC dated July 20,1981, which transmitted a design basis accident radiological study for the Prairie Island control room. Section 2.2 of that study, which describes the methodology for determination of the X/O value, is provided as Enclosure 3. The X/Q valuo utilized in the Westinghouse control room dose calculation is conservative with respect to the value determined by the 1981 design basis accident radiological study (Enclosure 3).

l

. s g

. 1-1: o ,7 6 m I: EH o

4 H

, . . . . ,/ . .

. . . . - -  :: 9

!:c ,

6 ,,

11 , .

m

. . - q !: -l m

. is . cJ_) .8 <'

o N ._1 -

@ ' 5.

l' A s e.  :  : --

g s a g; 4 4 .---

ij  ;

s W.- j'i . g. .s - . v .

m l

'~~il j o

'a O -  ?:E 7

• s I

.f g  % ,,

Z H I' E <,

l ";

9 .....,-

....  : ei---- e : e : ! P ,' ~ .

F',,

c' ,r' a

I, l' ip p

"-,,: : v :,: m t l e

o e s H g . ,,

,. y M

a 6 4 - - -

, #.,,<..gg : :g r y

g ; -

P P '

offI k g p -

,  ; ;- + o Z ilit "

N O e.

-c r. - . ...

.A"

,  !; si o g.

b

[-

is T- f? ',[,, $

o q -

Ep: a

, 1

'gg' E O 15 N A g  !: C*g"  !" if ,

l  :

H "h, g! e.1 I Z 1: Em <

. t jij Q b

.t U I

' ~

!:c I 6

E mm 3  : , . , ,

I iii E lI

! ei kI kE 4

e

, ...a

--en d _..-m-m--.+=----a ----_.m_.e- --.-----m---.a s a- z,.. .a. _,,..m.,,.4 . - - _,u_, J

• a

.- n p; E o I H

>: go. p A

H I: D. .

• :e: R : : : ; :,: : :;-; e := -

: : l

(_ 1: [*um o

{ge# e') , ,

6 ,,

y

' h I$ S4 Z e ij e

- _.)

.8 , . CD H

% }- -

o i: < >

EH e. t-

- 6 A N U) j! o m .  ! ,, ,

M lp v2 O', r[C l 9  :'+ = v: ':;, M 9

..  : n.'

i's **

Z O

I:I -  :

3 ,,.

'. /**>:.:i'r'+-'k's I N s y

o

,H y

f:l-..+;[ = .

" r eq o.

g d :o : 4 : e t n e::  ; : n: q[ 7" g) x O

5 [. . ..

s 5: c- g 5 ., , e - " 4

[ 6 eM, 5 4

• e' '.>e a, ' - ; : e; 1>

, ,+....

in

.) g I A Tp

' ,, ,5 $ 9 M 1*. *-+-

.. r. s i< ~ s m W,  !;o 1 o o o st- --

6 t o e o 3.o 4

O . -, -  ::e ," ;a a N O

O i'

~~

5

• " A--

- " " EE EE

/4 !:"

d

<, /( "

p.1 a

6g n ,,

[ -

, ; ^7 : ' ' ' '

,, 3i ,

O

'* 6

, [a [, , WW Z O [ ! ,, , ,,n N J I: Cm a  ! ,,' n't J'  !*E .

H ": :  : & : ; . g;;;-$: \ . :' v

,2i ir c1 "_ - E lJS *-+- ;EI i O c ,,

^

U ic g ."^

- l ra > r--

E ra,

$ I -1 I ll 1 t* 1-

. = .. .

.e e

. n E

E

. V,,

1: & + $';

• P  :::;t  ; $ + : : 'e  : : e =, O i:p = l; H 6 _ _ _ s. E4 3..... ...,...........,,--3 g

.  !:p ,

o a

[g.. m

. 7' gHh!: L;;i :d H Z '

ise n6 <>

, . ,21 Er.1 s ] g -

-> !e  : ,, M to ii +

e

s![

,6 Al a

H h

to J

l!I

, = ',

f/ .b.

e - - - 4'

,1 . .-

b .

l

@'m A .

i a'

4 .

i Z _ p+y ' :e : e'.,.,*.*v:

, ;e> , y h ,

g l O 1: Cv' =

[

i I o to H l, l

f i ,. O b  ::: @: .3:e:

: A 4

e

/ C '

>,* I

, C ,, ' '

A .

y v ,+ +a . 7, -

g xo

. .I .. ,. H

~ I: E== l 5 H y 6 +

s a

"V f  ;::= .--- <

.e $ M g D l. I $* l$ * $ 0 O

O g, r -4 lli g - ). .

A ,, ,, g g:

e I;t g

g,7, [:

6=

li gg g; a

O -

l6 ' I< Io LwJ Z

0 1,. n LvJ C,.4 = t,?n! '? I

-4

!6 E=1 ,

r .t 8

t [.ees 4 O ,

' {=i,r t

i y e 1: CM .I-t*

• l l

E

= 1 m , r- m,

{ l I n I

! { {

1* E-4

• ' **

• e & D

a Enclosure 2 Re Evaluation of

, Westinghouse Main Steam Line Break

Allowable Leak Rate Analysis j Westinghouse Letter Dated October 16,1997 i

f 1

F 4

l 2

i I

1 i

i d

. , , , - - - , n .- +- ~ --,..._~ , .,v--.v.~ +,

-- - , - - .- , , , , , . . . , . - - .,.. , - , . , ~ .

m___ . . _ _ _,___

.. C t

WESTINGHOUSE ENERGY SYSTEMS Nuciew somoes DMeion ELECTRIC CORPORATION Bou 158 l

Modeon Pomeylvania 1566SC158 NsD-E. TAP 00s6 I October 16,1997 Mr. Richard Pearson Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch,MN $5089

Reference:

(1) Northern States Power Company P.O. PJ6150SQ (2) NSD E-TAP-0032, Westinghouse (Lagally) to NSP (Pearson),

dated May 13,1997 (3) NSD-E-TAP-0079, Westinghouse (Lagally) to NSP (Pearson),

dated October 2,1997

Dear Mr. Pearson:

This letter transmits a re-evaluation of the Main Steam Line Break Allowable Leak Rate for the PINGP. The analysis was initially transmitted by Reference 2 and revised to address the NRC RAI and correct a discrepancy by Reference 3. The current revision incorporates the NRC methodology for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration of accident initiated iodine spike along with a 100%

iodine removal efficiency in the demineralizer beds.

The enclosure to this letter is Westinghouse intemal letter SAE CRA 97-273, which documents the analysis performed.

Should you have any questions regarding this transmittal, please contact me (412/ 722-5082) or Jill Monahan (412/ 274-4372).

Sincerely, A m= ~

Hermann Lagally Principal Engineer SG Tube Integrity Management 1

cc:

Jill Monahan/ECE (w/o attachments)

T.A. Pitterle N. Kury/ ECE W.B. Middlebrooks

SAE-CRA 97 273 From : Containment and Radiological Analysis WIN : 284-4372 Date - October l$.1997 i

Subject:

Final Repon on NSP Allowable Leak Rate Keywords: NSP/ DOSES l- To  : H. O. Lagally WM MS 27 l

cc  : L. C. Smith G. Whiteman

! J. L. Grover l

t

References:

1) CN CRA 97 004 R0

2) CN CRA-97-004 R1

3) CN CRA-97 004 R2 Attachment A to this letter provides the Westinghouse steam line break radiological analysis calculation summary on the Prairic Island Allowable Leakage submittal. Attachment B

_ provides the Final Repon for the radiological potrion of the allowable leak rate calculation.

This repon incorporates the NRC's methodology for an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> duration of accident initiated iodine spike along with a 100% iodine removal efficiency in the demineralizer beds.

If you have any questions. please contact me, i

Qaf S.rncoajun J, S. Monahan Containment and Radiological Analysis -

Reviewed by:

U. Bachrach Containment and Radioiogical Analysis

ATTACHMENT A Calculation Note Summary Purpose The offsite and control room radiological consequences from a main steam line break (MSLB) are calculated to determine the maximum allowable primary to secondary leakage. The primary to secondary leak is limited by the maximum permissible dose, offsite or control room.

Approach The methodology that will be employed in performing this calculation is documented in NUREG 0800, Standard Review Plan (SRP) Section 15.1.5, Appendix A. The thyroid doses will be calculated with 1) a pre existing lodine spike and 2) an accident initiated lodine spike. The criteria defined in SRP 15.1.5, Appendix A for the offsite dose limits are: pre accident lodine spike = amits in 10CFR100 or 300 rom thyrold; and accident initiated lodine spike = "small frr.4 tion" (10%) of limits in 10CFR100 or 30 rem thyrold.

(

The criteria defined in SRP Lection 6.4 was used for the control room dose limits: 30 rem thyroid.

The MSLB dose analysis includes two sources of activities: initial secondary side

! lodine activity and the activity in the primary to secondary steam generator tube leakage.

Five Separate Cases were modeled to determine the radiological consequences of the i MSLB. They were:

1. Maximum Nominal RCS Coolant Technical Specification lodine Activity

2. Maximum Secondary Side Technical Specification lodine Activity in the intact Steam Generator

3. Pre accident lodine Spike Case

- 4. Accident initiated lodine Spike Case

5. Maximum Secondary Side Technical Specification lodine Activity in the Faulted Steam Generator .

The total radiological consequences from a pre accident lodine spike is the summation of the thyroid doses from cases 2,3 and 5. The total radiological consequences from

- an accident initiated iodine spike is the summation of the thyroid doses from cases 1,2, 4 and 5.

From sensitivity runs, the maximum allowable primary to secondary leakage was determined. The cases discussed on the following page assume the maximum allowable primary to secondary leakage of 4620 grams / minute to the faulted steam generator.

---

Y

Control Room Model The control model employed in each of the cases listed above is consistent with the values provided in the input assumptions.

Case 1 Maximum Nominal RCS Coolant Activity This case models the initial RCS coolant activity of 1.0 pCl/gm of D. E.1131 with transfers through steam generator tube leakage to the faulted and intact steam generators. The lodine in the primary to secondary leakage to the faulted steam generator is assumed to be directly released to the environment without partitioning.

The lodine in the primary to secondary leakage into the intact steam generator is mixed with the liquid in this steam generator and released in the sie n to the environment assc.ning a partitioning factor of 0.01 consistent with SRP 15.1.5 App. A.

The maximum coolant activities provided in the USAR were based on 1% fuel defects.

These values were converted to 1.0 pCl/gm of D. E.1131 based on the dose l

conversion factors (DCFs)in TID 14844 which are the bases for the Prairie Island Technical Specification for measurement of specific activity. The RCS volume is 5227.39 ft*, The RCS mass is calculated based on density at full power design conditions (2250 psia and 567.3 *F). The following initial coolant activities were assumed in this case.

1131 77.38 Cl 1132 22.73 Cl 1133 93.52 Cl 1134 14.30 Cl 1135 51.34 Cl The removal processes assumed in this case include normal iodine decay and transfers to the secondary side. The transfer to the faulted steam generator via primary to secondary leakage is released directly to the atmosphere since the secondary side coolant is released within the first few minutes of tho transient. The maximum allowable leak rate from the primary to the secondary is 4620 gm/ min. The leak rate to the intact steam generator is 393.5 gm/ min with a steam rate from the intact steam generator as documented in the input assumptions. A partition factor of 0.01 is applied to the intact steam generator steam releases.

The thyroid inhalation doses are calculated using the equation provided in the Prairie Island USAR Appendix D Section D.8.4 using the dose conversion factors in ICRP 30.

~

Caso 2 Initial Secondary lodino Activity Intact Steam Generator This caso models the initialintact steam generator coolant activity of 0.1 pCl/gm of D.

E.1131. This activity is released in the steam to the environment assuming a partitioning factor of 0.01 consistent with SRP 15.1.5 App. A.

The maximum secondary intact steam generator activity is calculated by taking the initial RCS coolant activillos corresponding to 0.1 pCl/gm D. E.1131and multiplying by the mass differences in the intact SG and the RCS coolant. The SG liquid mass is 1091551bm. The initialintact SG activities are:

1131 3.5465 Cl -

l132 1.0420 Cl 1133 4.2860 Cl 1134 0.6555 Cl 1135 2.3530 Cl The removal processes assumed in this case include normal lodino decay and transfers to the atmosphore through steaming with a steam rate from the intact steam generator as documented in the input assumptions. A partition factor of 0.01 is applied to the intact steam generator steam releases.

l The thyrold inhalation doses are calculated using the equation provided in the Prairie 1

Island USAR Appendix D Section D.8.4 using the dose conversion factors in ICRP 30.

Case 3 Pre accident lodine Spiking This case models the initial RCS coolant activity of 60 pCl/gm of D. E.1131 with transfers through steam generator tube leakage to the faulted and intact steam generators. The lodino in the primary to secondary leakage to the faulted steam generator is assumed to be directly released to the environment without partitioning.

The iodine in the primary to secondary leakage into the intact steam generator is mixed with the liquid in this steam generator and released in the steam to the environment assuming a partitioning factor of 0,01 consistent with SRP 15.1.5 App. A.

The pre accident iodine spike is assumed to have occurred and raised the iodine concentration in the RCS to 60 times the initial value. The coolant activities at the start of the event are:

1-131 4.64E3 Ci 1132 1.36E3 Ci 1133 5.61 E3 Cl 1134 8.5852 Cl 1135 3.0BE3 Ci

The removal processes assumed in this case include normallodine decay and transfers to the secondary side. The transfer to the faulted steam generator via primary to secondary leakage is released directly to the atmosphere since the secondary side coolant is released within the first few minutes of the transient. The maximum allowable leak rate from the primary to the secondary is 4620 gm/ min. The leak rate to the intact steam generator is 393.5 gm/ min with a steam rate from the intact steam generator as documented in the input assumptions. A partition factor of 0.01 is applied

to the intact steam generator steam releases.

The thyroid inhalation doses are calculated using the equation provided in the Prairie l Island USAR Appendix D Section D.8.4 using the dose conversion factors in ICRP 30.

Case 4 Accident Initiated lodina Spiking This case models the initial RCS coolant activity of 1.0 pCl/gm of D. E.1131. An iodine spike is initiated by the accident, caused by either the reactor trip or the depressurization, This spike increases the RCS coolant lodine appearance to 500 times the normal equilibrium value (SRP 15,1.5, App. A).

The same transfers through steam generator tube leakage to the faulted and intact steam generators are modeled as in the normal lodine case. The iodine in the primary to secondary leakage to the faulted steam generator is assumed to be directly released to the environment without partitlening. The lodine in the primary to secondary leakage into the intact steam generator is mixed with the liquid in this steam generator and released in the steam to the environment assuming a partitioning factor of 0.01 consistent with SRP 15.1.5 App. A.

The initial RCS coolant activities from Case 1 are:

1131 77.38 Ci 1132 22.73 Cl 1133 93.52 Cl 1134 14.30 Ci 1135 51.34 Cl .

To reach an equilibia m concentration, the appearance rate must equal the removal rate. The removal of iodine from the RCS corsists of two mechanisms, normallodine decay and removal by the letdown flow through the mixed bed demineralizers. The removal rate due to letdown purification is:

Ayna = letdown flow /RCS mass where: letdown flow = 40 gpm normal at 127'F,15 psig Aynam = 2.297E 05/sec

The decay constants for each lodine isotope are:

1131 9.97E 7/see I132 8.42E 5/sec l133 9.25E 6/sec l134 2.20E 4/sec l135 2.92E 5/sec The activity appearance rate equals the activities times the removal rate. The appearance rate at the normal equilibrium lodine concentration were calculated as:

l l131 0.1112 Cl/ min 1132 0.1464 Cl/ min L l133 0.1806 Cl/ min 1134 0.2082 Cl/ min 1135 0.1608 Cl/ min i

i The accident initiated spike is 500 times these values:

1131 55.6 Cl/ min 1132 73.2 Cl/ min 1133- 90.3 Ci/ min 1134 104.1 Cl/ min 1135 80.4 Cl/ min f

The totaliodine appaarance rate is 403.6 Cl/ min. There is no initial RCS coolant activity in this casa. This is modeled as a transfer from the fuel to the RCS coolant at the rates above. The iodine appearance rateis terminated at 8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />.

The removal processes assumed in this case it clude normal lodine decay and transfers to the secondary side. The transfer to the faulted steam generator via primary to secondary leakage is released directly to the atmosphere since the secondary side coolant is released within the first few minutes of the transient. The maximum allowable leah rate from the primary to the secondary is 4620 gm/ min. The leak rate to the intact steam generator is 393.5 gm/ min with.a steam rate from the intact steam genarator as documented in the input assumptions. A partition factor of 0.01 is applied to tne intact steam generator steam releases.

The thyroid inhalation doses are calculated using the equation provided in the Prairie Island USAR Appendix D Section D.8.4 using the dose conversion factors in ICRP 30.

Case 5 Initial Secondary lodine Activity Faulted Steam Generator d

This case models the initial faulted steam generator coolant activity of 0.1 pCi/gm of D.

E.1 131. This activity is released entirely to the atmosphere without partitioning through the MSLB.

The maximum secondary faulted steam generator activity is calculated by taking the initial RCS coolant activities corresponding to 0.1 pCl/gm D. E.1131 and multiplying by the mass differences in the faulted SG and the RCS coolant. The SG mass is 109155 lbm. The initial faulted SG activities are:

1131 3.5465 Cl 1132 1.0420 Cl 1133 4.2860 Cl 1134 0.6555 Cl I135 2.3530 Cl Th9 caamnts in the faulted steam generator are released over a 15 minute time Interval to p* ovide the control room ventilation system a time dependent release. There is no

dechv procer 3 assumed in this case.

The thyroid inhalation doses are calculated using the equation provided in the Pralrie Island USAR Appendix D Section D.8.4 using dose conversion factor in ICRP 30.

RESULTS The site boundary, low population zone and control room thyroid doses for each individual case are provided below.

CASE 1 2 3 4 5 SB (rem) 0.1159 0.02387 6.951 5.414 1.045 LPZ (rem) 0.1221 0.01807 7.321 22.91 0.2849 CR (rem) 0.1762 0.02701 10.57 28.50 1.072 The total pre accident and accident initiated lodine spiking radiological consequences l are:

Pre accident lodine Spike (rem)

Site Boundary = 8.02 Low Pop Zone = 7.62 i

Control Room = 11.67

Accident initiated lodine Spike ( rem)

Site Boundary = 6.60 Low Pop Zone = 23.34 Control Room = 29.78 The Control Room thyroid dose for the accident initiated lodine spike case is the limiting case with respect to the maximum allowable leakage, l

1 l

.