Text

Semiannual Effluent Release Rept 4,Jan-June 1986
	ML20214W853
Person / Time
Site:	Limerick
Issue date:	06/30/1986
From:	Alden W; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Murley T; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
	NUDOCS 8612100429
	Download: ML20214W853 (84)
	v • d • e

I PHILADELPHIA ELECTRIC COMPANY LIMERICK GENERATING STATION UNIT NO. 1 DOCKET NO. 50-352 I

I I

SEMI-ANNUAL EFFLUENT RELEASE REPORT NO. 4 JANUARY 1, 1986 THROUGH JUNE 30, 1986 I

Submitted to The United States Nuclear Regulatory Commission Pursuant to Facility Operating License NPF-39 I

l Preparation Directed By:

J.

F. Franz, Manager Limerick Generating Station l

1 I

I 8612100429 860630 PDR ADOCK 0500 2

R I

l

I

\\

I TABLE OF CONTENTS I

I I.

Introduction I

II.

Tables A.

Summary of Radioactive Liquid and Gaseous Effluents 1.

Liquid Effluents 2.

Gaseous Effluents B.

Solid Waste Disposition Report III.

Attachments A.

Supplemental Information E

B.

Changes to the Offsite Dose Calculation Manual (oDCM) and Support Documentation I

C.

"B" Residual Heat Removal Service Water Radiation Monitor Failure I

iI lI i

'I I

lI

l I

I.

INTRODUCTION This submittal complies with the format described in I

Regulatory Guide 1.21, " Measuring, Evaluating and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water I

Cooled Nuclear Power Plants", Revision 1, June, 1974 with data summarized on a quarterly basis following the format of Appendix B thereof.

I In accordance with LGS Technical Specification 6.9.1.8, this document summarizes the gaseous and liquid effluent releases for the period of January 1, 1986 through June 30, I

1986.

Where "0.00E+00" is used, it denotes the less than detectable level for the given isotope.

These values are available upon request.

The resultant offsite doses from the gaseous and liquid pathways to members of the public I

are not required (per the aforementioned Technical Specification) to be reported at this time; but will be evaluated in the July 1, 1986 through December 31, 1986 i

submittal.

Supplemental information - consistent with the format of Appendix B of R.G.

1.21 is also included.

Disposition of solid waste during the report period is described within, to include: the total activity shipped by waste type and an estimate of the error in the reported totals; the estimated composition of each type of waste by I

isotope; and the solid waste disposition, i.e. number of shipments, mode of transportation, destination, type of container, total container volume, and solidification agent.

I Revision 4 of the Offsite Dose Calculation Manual is included with all justifications.

There were no revisions to the Process Control Program (PCP) during the report period.

Per Technical Specification 3.3.7.11, this report also I

contains an explanation of why an inoperable Residual Heat Removal Service Water Radiation Monitor (B) was not returned to service within the time specified.

I I

lI I

1 e.

.. = =ew--

m.w.we-

.a..

m.-.

em..

a a

a x

I I

I l

I I

II.

TABLES A.

SUMMARY

OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENTS 1.

sie 1e ee<1. e.

g I

I I

l I

I

=

1 l

SITE: LIMERICK Page 1 of 1 UNIT: U1 I

USER: MART DATE:

8/14/86 14:56 EFFLUENT AND WASTE DISPOSAL REPORT I

LIQUID EFFLUENTS -- SUMMATION OF ALL RELEASES UNITS :

QUARTER :

QUARTER EST.

TOTAL:

I 1

2

ERROR, %:

A. FISSION AND ACTIVATION PRODUCTS l 1. TOTAL RELEASE (EXCL.: CI

0.4 80E-02 : 0.000Et00 0.326Ef01 :

TRIT., GASES, ALPHA):

I

! 2. AVERAGE DILUTED UCI/ML : 0.265E-08 : 0.000Ef00 :

CONC. DURING PERIOD :

l l

3. PERCENT OF

0.000Ef00 : 0.000Ef00 : CURIE LIMITS NCrr APPLICABLE LIMIT

APPLICABLE t

B. TRITIUM

1. TOTAL RELEASE

CI

0.974Ef00 0.000Ef00 : 0.164Ef01 :

I

2. AVERAGE DILUTED

UCI/ML : 0.560E-06 : 0.000Ef00 :

CONC. DURING PERIOD :

I

3. PERCENT OF

0.000Ef00 0.000Ef00 : CURIE LIMITS NOT APPLICABLE LIMIT

APPLICABLE C. DISSOLVED AND ENTRAINED GASES I

1. TOTAL RELEASE

! CI

0.136E-02 : 0.000Ef00 : 0.403Ef01 :

2. AVERAGE DILUTED UCI/ML ! 0.779E-09 : 0.000Ef00 :

I CONC. DURING PERIOD :

3. PERCENT OF

0.000Ef00 : 0.000Et00 : CURIE LIMITS NCfr APPLICABLE LIMIT

APPLICABLE I

D. GROSS ALPHA RADI0 ACTIVITY I

1. TOTAL RELEASE

CI

0.244E-04 : 0.000Ef00 : 0.978Ef02 :

E. VOLUME WASTE RELEASED l LITERS : 0.218Ef07 : 0.000Ef00 ! 0.100Ef01 :

(PRIOR TO DILUTION)

I F. VOLUME DILUTION WATER LITERS ! 0.174Ef10 : 0.000Ef00 : 0.100Ef01 :

USED DURING PERIOD I

PAGE 1 0F 5 SITE: LIHERICK UNIT: U1 I

USER: MART DATE: 8/14/86 14:57 I

EFFLUENT AND WASTE DISPOSAL REPORT LIQUID EFFLUENTS FOR RELEASE POINT:

1 LIQUID RAD WASTE DISCHARGE TO SCHUYLKILL RIVER I

CONTINUOUS MODE BATCH NODE E.

NUCLIDES : UNITS :

QUARTER : QUARTER

QUARTER : QUARTER :

RELEASED :

1 2

I

H3

CI

0.000Ef00 : 0.000Ef00 : 0.974Ef00 : 0.000Ef00 :

l

C14

CI

0.000Ef00 ! 0.000Ef00 0.000Ef00 0.000Ef00 :

i NA24

CI

! 0.000Ef00 : 0.000Ef00 0.971E-05 0.000Ef00 :

l P32

CI

0.000Ef00 : 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

I

CR51

CI

0.000Ef00 : 0.000Ef00 0.209E-02 : 0.000Ef00 :

l MN54

CI

0.000Ef00 0.000Ef00 ! 0.535E-05 : 0.000Ef00 :

I

! HN56

CI

0.000Ef00 0.000Ef00 ! 0.000Ef00 0.000Ef00 :

FESS

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

l

FE59 CI

0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Et00 :

I C058

CI

0.000Ef00 : 0.000Ef00 : 0.111E-02 0.000Et00 :

C060 CI

0.000Ef00 1 0.000Ef00 : 0.926E-04 ! 0.000Ef00 :

I NI63

CI

0.000Ef00 : 0.000Ef00 1 0.000Ef00 : 0.000Ef00 :

! HI65

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I

CU64 CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

i ZN65 CI

0.000Ef00 : 0.000Ef00 : 0.116E-02 0.000Ef00 :

I

ZN69

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 0.000Ef00 :

BR83

CI

0.000Ef00 : 0.000Ef00 i 0.000Ef00 : 0.000Ef00 :

i BR84

! CI 0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

BR85

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

1-ENTER [ C 3 TO ERASE SCREEN AND CONTINUE : C I

PAGE 2 0F 5 SITE: LIMERICK UNIT: U1 I

USER: MART DATE:

8/14/86 14:57 I

EFFLUENT AND WASTE DISPOSAL REPORT LIQUID EFFLUENTS FOR RELEASE POINT:

1 LIQUID RAD WASTE DISCHARGE TO SCHUYLKILL RIVER I

CONTINUOUS HODE BATCH MODE I

NUCLIDES : UNITS : QUARTER

QUARTER QUARTER :

QUARTER :

RELEASED :

1 2

I LIQUID EFFLUENTS (CONTD)

I

RB86

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

RB88

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I l RB89

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

SR89

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I SR90

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

SR91

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

SR92

CI

0.000Ef00 0.000Ef00 ! 0.000Ef00 0.000Ef00 :

Y90

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

Y91H

CI

0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

Y91

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

Y92

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 0.000Ef00 :

I I

Y93

CI

0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

ZR95

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Et00 :

ZR97

! CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

NB95

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

H099

CI

0.000Ef00 0.000Ef00 0.747E-04 0.000Ef00 :

l TC99M

CI

0.000Ef00 0.000Ef00 0.747E-04 : 0.000Ef00 :

TC101

! CI

0.000Ef00 0.000Ef00 : 0.000Ef00 0.000Ef00 :

l

[

ENTER C C ] TO ERASE SCREEN AND CONTINUE : C ll

PAGE 3 0F 5 SITE: LIMERICK UNIT: U1 I

USER: MART DATE: 8/14/86 14:58 EFFLUENT AND WASTE DISPOSAL REPORT LIQUID EFFLUENTS FOR RELEASE POINT:

1 LIQUID RAD WASTE DISCHARGE TO SCHUYLKILL RIVER I

s CONTINUOUS MODE BATCH NODE I

NUCLIDES

UNITS : QUARTER

QUARTER QUARTER : QUARTER :

RELEASED :

1 2

l LIQUID EFFLUENTS (CONTD)

I RU103

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

RU105

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Et00 :

I RU106 CI

0.000Ef00 0.000Et00 0.000Et00 : 0.000Ef00 :

AG110M

CI

0.000Ef00 0.000Et00 : 0.000Et00 : 0.000Et00 :

I TE125H

CI

0.000Et00 : 0.000Et00 : 0.000Ef00 : 0.000E+00 :

I TE127H

! CI

0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Et00 :

TE127

! CI

0.000Ef00 0.000Ef00 0.000EiOO : 0.000Ef00 :

I TE129H

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

i

! TE129

CI

0.000Ef00 : 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

! TE131H

! CI

! 0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Et00 :

I l TE131

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I TE132

CI 0.000Et00 : 0.000Et00 : 0.000Et00 : 0.000Et00 :

I130

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 !

I 1131

CI

0.000Et00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I132

CI

0.000Ef00 0.000Et00 : 0.000Ef00 0.000Ef00 :

I 1133

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

1134

CI

0.000Ef00 0.000Ef00 ! 0.000Ef00 0.000Ef00 :

I135

CI

0.000Ef00 0.000Ef00 0.000Ef00 0.000Ef00 :

r

\\

ENTER C C J TO ERASE SCREEN AND CONTINUE :

C' l

Il t

-,~, -

~.~w~.,- ~~ _ -

.-- -,~-~,-

PAGE 4 0F 5 SITE: LINERICK I

UNIT: U1 USER: MART DATE: 8/14/86 14:41 EFFLUENT AND WASTE DISPOSAL REPORT 8

LIQUID EFFLUENTS FOR RELEASE POINT:

1 LIQUID RAD WASTE DISCHARGE TO SCHUYLKILL RIVER CONTINUOUS MODE BATCH MODE I

! NUCLIDES : UNITS :

CUARTER :

GUARTER : QUARTER

GUARTER RELEASED :

1 2

LIQUID EFFLUENTS (CONTD)

I CS134

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

CS136 CI

0.000Ef00 0.000Ef00 0.000Ef00 0.000Ef00 :

I i CS137 CI

0.000Ef00 : 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

CS138 CI

0.000Ef00 0.000Ef00 ! 0.000Ef00 ! 0.000Ef00 :

I

! BA139

CI 0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

! BA140

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

E BA141 CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 0.000tf00 :

BA142

CI

0.000Ef00 : 0.000Ef00 ! 0.000Ef00 ! 0.000Cf00 :

I LA140

CI

! 0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

! LA142

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 0.000Ef00 :

CE141

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I CE143

CI

0.000Ef00 0.000Ef00 ! 0.000Ef00 0.000Ef00 :

CE144

! CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I PR143

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 ! 0.000Ef00 :

PR144 CI

! 0.000Ef00 0.000Ef00 : 0.000Ef00 ! 0.000Ef00 :

I

! ND147

CI

! 0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I

! W187 CI

0.000Ef00 : 0.000Ef00 0.000Ef00 ! 0.000Ef00 :

! NP239

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 0.000tf00 :

I ENTER C C J TO ERASE SCREEN AND CONTINUE :

C I

e PAGE 5 0F 5 l

SITE: LIMERICK UNIT: U1 USER: MART DATE: 8/14/86 14:42

,u EFFLUENT AND WASTE DISPOSAL REPORT LIQUID EFFLUENTS FOR RELEASE POINT:

1 LIQUID RAD WASTE DISCHARGE TO SCHUYLKILL RI')ER 7

CONTINUOUS H0DE BATCH N0DE

NUCLIDES : UNITS : QUARTER : QUARTER : QUARTER : QUARTER :

RELEASED :

1 2

LIQUID EFFLUENTS (CONTD)

NONE CI

0.000E+00 : 0.000Ef00 ! 0 187E-03 : 0.000Et00 :

-m-_

m.me-em._e.___e====mm______-m._=_.._e_.-_m

.=___-_=-we_-_-

_e l

TOTAL FOR :

PERIOD

CI

0.000Ef00 : 0.000Ef00 : 0.970EF00 : 0.000Ef00 :

(ABOVE)

._W-_M_--m--_-W.______.W_______-___-m______eme__m-m_____me___-___-m-_.

XE-133

CI

0.000Ef00 : 0.00,0Ef00 1 0.972E-03 : 0.000E+00 :

[

I"x'E'[$s'~~I~CI"'I'0!000EI00I'OI0OO$+'00I'055e'It-0$~I'0I00O$I00I

' ~ ~ ~

'~'

EFFLUENT RELEASE

SUMMARY

OPTIONS 1 -- TERMINATE 2 -- ACCUMULATE GASEOUS RELEASES 3 -- ACCUMULATE LIQUID RELEASES 4 -- PRINT WASTE

SUMMARY

REPORT ENTER OPTION SELECTION C 1-4 ] :

4

- As-76

I

'I I

I I

I II.

TABLES I

A.

SUMMARY

OF RADIOACTIVE LIQUID AND GASEOUS EFFLUENTS 2.

Gaseous Effluents I

I

.I

'I I

I I

=-

r SITE: LIHERICK Page 1 of 1 UNIT: U1 B

USER: MART DATE: 8/14/86 14:31 l

l EFFLUENT AND WASTE DISPOSAL REPORT r~

l GASEOUS EFFLUENTS -- SUMMATION OF ALL RELEASES l

UNITS :

QUARTER :

QUARTER

EST. TOTAL:

1 2

ERROR, %:

l A. FISSION AND ACTIVATION GASES g

E

1. TOTAL RELEASE

CI

0.366Ef00 0.153E-02 0.470E+02 :

2. AVERAGE RELEASE

UCI/SEC: 0.470E-01 : 0.194E-03 :

I RATE FOR PERIOD 2 3. PERCENT OF TECHNICAL:

0.000Ef00 0.000E+00 : CURIE LIMITS NOT I

SPECIFICATION LIMIT :

APPLICABLE f

'E B. 10 DINES g

l

1. TOTAL 10 DINE-131 CI

0.000Ef00 : 0.000Ef00 : 0.000E+00 :

f I

2. AVERAGE RELEASE

!UCI/SEC: 0.000Ef00 ! 0.000E+00 :

RATE FOR PERIOD l

l

3. PERCENT OF TECHNICAL!

0.000Ef00 1 0.000Ef00 : CURIE LIMITS NCfr SPECIFICATION LIMIT :

APPLICABLE C. PARTICULATES I

I HALF-LIVES 38 DAYS !

2. AVERAGE RELEASE UCI/SEC: 0.000Ef00 ! 0.000E+00 :

RATE FOR PERIOD

'I

3. PERCENT OF TECHNICAL:

0.000E+00 1 0.000E+00 : CURIE LIMITS NOT SPECIFICATION LIMIT :

APPLICABLE I

4. GROSS ALPHA CI

0.000Ef00 0.000E+00 :

RADI0 ACTIVITY I

D. TRITIUM I

1. TOTAL RELEASE

CI

0.000Et00 : 0.000E+00 : 0.000Ef00 :

2. AVERAGE RELEASE

!UCI/SEC: 0.000E+00 : 0.000Ef00 :

I RATE FOR PERIOD

3. PERCENT OF TECHNICAL:

0.000Ef00 : 0.000Ef00 :

I SPECIFICATION LIMIT :

I

-~

__z-J

I PAGE 1 0F 9 SITE LIMERICK I

UNIT: U1 USER: MART DATE: 8/14/86 14:31 I

GASEOUS EFFLUENTS FOR RELEASE POINT:

1 NORTH STACK CONTINUOUS H0DE BATCH NODE I

NUCLIDES : UNITS : QUARTER :

QUARTER QUARTER : QUARTER

RELEASED :

1 2

1. FISSION GASES I

AR41

CI

0.000E+00 : 0.000Ef00 0.104E-01 : 0.000E+00 :

KR83M

CI

0.000Ef00 0.000Ef00 0.000E+00 : 0.000E+00 :

I KR85M

CI

0.000E+00 : 0.000Ef00 : 0.313E-02 : 0.000E+00 :

I

KR85

CI

0.000Ef00 0.000E+00 : 0.000E+00 : 0.207E-03 :

KR07

CI

0.000E+00 0.000Ef00 ! 0.196E-01 : 0.000E+00 :

I

KR88 CI

0.000Et00 : 0.000Ef00 0.124E-01 : 0.000Ef00 :

KR89 CI

0.000Ef00 0.000Ef00 : 0.000E+00 0.000Ef00 :

I KR90

CI

0.000E+00 0.000E+00 : 0.000Et00 ! 0.000E+00 :

XE131M CI

0.000Et00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

I

XE133M

CI

! 0.000Ef00 ! 0.000E+00 0.000Ef00 0.000Ef00 :

I

XE133

CI

0.000Ef00 : 0.000Ef00 : 0.000E+00 : 0.000Ef00 :

XE135M

CI

! 0.000Ef00 : 0.000Ef00 : 0.515E-01 : 0.000Ef00 :

l XE135 Cs

0.000Ef00 0,.000Ef00 0.822E-02 0.000Ef00 :

! XE137 CI

0.000Ef00 : 0.000E+00 0.000Ef00 : 0.000E+00 :

I

XE138 CI

0.000Ef00 0.000Et00 : 0.258E+00 0.000Ef00 :

UNIDENT.

CI

0.000Ef00 ! 0.000Ef00 0.000Ef00 : 0.000E+00 :

I TOTAL FOR :

PERIOD CI

0.000Ef00 0.000Et00 ! 0.364E+00 : 0.207E-03 :

I (ABOVE)

ENTER C C 3 TO ERASE SCREEN AND CONTINUE : C I

I

l ll PAGE 2 0F P SITE: LIMERICK I

UNIT: U1 USER: MART DATE: 8/14/86 14:32 EFFLUENT AND WASTE DISPOSAL REPORT I

GASEOUS EFFLUENTS FOR RELEASE POINT: 1 NORTH SinCK CONTINUOUS MODE BATCH NODE I

NUCLIDES : UNITS : GUARTER : QUARTER : QUARTER : QUARTER

RELEASED :

1 2

2. 10 DINES I

I131 CI

0.000Ef00 : 0.000Ef00 : 0.000E+00 0.000E+00 :

1133 CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I TOTAL FOR :

PERIOD

CI

0.000E+00 0.000E+00 : 0.000E+00 : 0.000E+00 :

I (ABOVE)

I

. 3. PARTICULATES C14

CI

0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I CR51

! CI

0.000Ef00 i 0.000E+00 : 0.000Ef00 : 0.000E+00 :

HN54

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 ! 0.000E+00 :

I FE59

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

C058

CI

! 0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

! C060

CI

0.000E+00 : 0.000E+00 1 0.000Ef00 ! 0 000Ef00 :

ZN65

! CI

! 0.000Ef'00 : 0.000Ef00 0.000Ef00 : 0.000E+00 :

I SR89

CI

0.000E+00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I

SR90

CI

0.000Ef00 : 0.000Et00 0.000Ef00 0.000Ef00 !

ZR95

! CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

I SB124

! CI

0.000Ef00 : 0.000Ef00 1 0.000Ef00 : 0.000Ef00 :

CS134

! CI

! 0.000Ef00 0.000Ef00 : 0.000Ef00 0.000Ef00 I

CS136 CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I

CS137

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000E+00 :

ENTER C C 3 TO ERASE SCREEN AND CONTINUE :

C I

---.~v-r-%-

m_ _ ___ _ _

- - =

m---.-.-----

' I PAGE 3 0F 9 SITE: LIMERICK I

UNIT: U1 USER: MART DATE: 8/14/86 14:32 EFFLUENT AND WASTE DISPOSAL REPORT I

GASEOUS EFFLUENTS FOR RELEASE POINT:

1 NORTH STACK CONTINUQUS MODE BATCH MODE I

NUCLIDES

UNITS : QUARTER : QUARTER

QUARTER :

RELEASED 1

2 1

2

3. PARTICULATES (CONTD)

I

BA140

CI

! 0.000Ef00 0.000Ef00 0.000Ef00 : 0.000Ef00 :

CE141

CI

0.000Ef00 i 0.000Ef00 0.000Ef00 0.000Ef00 :

I

! CE144

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

I

UNIDENT.

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 0.000Ef00 :

TOTAL FOR

PERIOD CI

0.000Ef00 0.000Ef00 0.000Ef00 0.000Ef00 :

I (ABOVE)

I I

I

I PAGE 4 0F 9 SITE: LIHERICK I

UNIT: U1 USER: MART DATE: 8/14/86 14:32 EFFLUENT AND WASTE DISPOSAL REPORT I

GASEQUS EFFLUENTS FOR RELEASC POINT:

2 UNIT 1 - SOUTH STACK CONTINU0US MODE BATCH NODE NUCLIDES : UNITS : QUARTER :

QUARTER l QUARTER : GUARTER RE' EASED :

1 1

2 1

2

1. FISSION GASES I

AR41

! CI

! 0.000Ef00 : 0.000Ef00 : 0.000Ef00 0.132E-02 :

KR83M CI

! 0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

! KR85M

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

KR85 i CI

0.000Ef00 : 0.000Ef00 0.000Ef00 t 0.000Ef00 :

I KR87

CI

0.000Ef00 : 0.000Ef00 1 0.000Ef00 : 0.000Ef00 :.

I KR88

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 ! 0.000Ef00 :

KR89 CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I KR90 CI

! 0.000Ef00 1 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

XE131H

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I XE133H CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 t I

XE133 CI

! 0.000Ef00 : 0.000Ef00 : 0.185E-02 : 0.000Ef00 :

XE135M

CI

0.000Ef00 i 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

I

! XE135 CI

0.000Ef00 ! 0.000Ef00 1 0.119E-03 : 0.000Ef00 :

XE137

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 ! 0.000Ef00 :

I I XE138

CI i 0.000Ef00 : 0.000Ef00 : 0.000Ef00 'l 0.000Ef00 :

UNIDENT.

CI

0.000Et00 ! 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

I t

TOTAL FOR :

PERIOD

CI

0.000Ef00 : 0.000Ef00 : 0.197E-02 ! 0.132E-02 :

I (AB0VE)

ENTER C C 3 TO ERASE SCREEN AND CONTINUE :

C I

I

I PAGE 5 0F 9 SITE: LIHERICK

' I UNIT: U1 USER: MART DATE: 8/14/86 14:33 EFFLUENT AND WASTE DISPOSAL REPORT I

GASE0US EFFLUENTS FOR RELEASE POINT: 2 UNIT 1 - SOUTH STACK CONTINUOUS H0DE BATCH H0DE I

NUCLIDES

UNITS : QUARTER ! OVARTER ! GUARTER : QUARTER

RELEASED i 1

2 1

2

2. 10 DINES I

1131

CI

0.000Ef00 1 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

I133

CI

! 0.000Ef00 ! 0.000Ef00 t 0.000Ef00 : 0.000Ef00 :

I i TOTAL FOR :

8

PERIOD

! CI i 0.000Ef00 : 0.000Ef00 0.000Ef00 0.000Ef00 :

I (ABOVE) i i

3. PARTICULATES C14 1

CI

! 0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

! CR51 CI i 0.000Ef00 : 0.000Ef00 ! 0.000Ef00 1 0.000Ef00 :

! HN54

CI i 0.000Ef00 ! 0.000Ef00 ! 0.000Ef00 : 0.000Ef00 :

I i FE59 I CI

0.000Ef00 t 0.000Ef00 0.000Ef00 0.000Ef00 :

C058 8 CI i 0.000Ef00 ! 0.000Ef00 1 0.000Ef00 : 0.000Ef00 :

I

! C060

! CI

0.000Ef00 1 0.000Ef00 8 0.000Ef00 ! 0.000Et00 :

ZN65

CI

! 0.000Ef00 8 0.000Ef00 i 0.000Ef00 1 0.000Ef00 :

I SR89

! CI i 0.000Ef00 : 0.000Ef00 t 0.000Ef00 0.000Ef00 :

I

! SR90 t CI t 0.000Ef00 8 0.000Ef00 : 0.000Ef00 ! 0.000Ef00 i t

ZR95

CI i 0.000Ef00 0.000Ef00 ! 0.000Ef00 0.000Ef00 :

I I SB124 CI

0.000Ef00 t 0.000Ef00 1 0.000Ef00 ! 0.000Ef00 :

CS134 i CI i 0.000Ef00 0.000Ef00 1 0.000Ef00 : 0.000Ef00 1 I

CS136 i

CI t 0.000Ef00 1 0.000Ef00 1 0.000Ef00 ! 0.000Ef00 :

I I CS137 I

CI

0.000Ef00 ! 0.000Ef00 : 0.000Ef00 1 0.000Ef00 i ENTER [ C J TO ERASE SCREEN AND CONTINUE ! C I

Pcge 6 of 7 PAGE FROM TO REASON FOR CHANGE 20.

Environm:ntal On-sito or Near Map replaced with a Sampling Stations the Limerick more legible version.

Site Boundary Generating No changes to Station contents of map.

21.

Evironmental Aquatic and Intermediate map Sampling Station Terrestrial broken into 2 Intermediate Environmental maps (p 21. &

Distance Sampling Stations p 22.) when map at Intermediate replaced with more Distances from the legible versions.

Limerick Generating (p. 21 & 22)

(P 21.)

22.

Airborne and TLD Environmental Sampling Stations

(

at Intermediate Distances from the Limerick Generating Station (p 22.)

23.

Environmental Environmental Sampling Stations Sampling Stations

[

Distant Locations at Remote Distances

[

from the Limerick Generating Station 24.

The setpoints will 3.11.1.1 Corrections also assure that a concentrations listed on Tech.

Spec. Table 3.11.1.1-1 for dissolved or entrained noble gas is not exceeded 24.

5 = margin of Fi previously omitted safety factor including F uncertainty, to 29.

B3. Containment

... equal to Addition Purge Isolation 2.1 uC/cc, to a value per Technical less than or Specification equal to 2.1 Table 3.3.2-2 j'

uCi/cc. The total Primary W

Containment Isolation. The l

4

PAGE 6 0F 9 SITE: LIMERICK UNIT U1 I

USER: MART DATE:

8/14/86 14:34 I

EFFLUENT AND WASTE DISPOSAL REPORT I

GASEOUS EFFLUENTS FOR RELEASE POINT! 2 UNIT 1 - SOUTH STACK CONTINUOUS H0DE BATCH NODE NUCLIDES

UNITS : QUARTER :

QUARTER : QUARTER

QUARTER :

RELEASED 1

2 1

2

3. PARTICULATES (CONTD)

I l

BA140

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000E+00 :

CE141

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I CE144

CI

0.000Ef00 : 0.000Ef00 : 0.000E+00 0.000Ef00 :

UNIDENT.

CI

0.000Et00 : 0.000Et00 : 0.000Et00 0.000Et00 :

I TOTAL FOR :

PERIOD

! CI

0.000Ef00 : 0.000E+00 : 0.000Ef00 : 0.000E+00 :

I (ABOVE)

ENTER C C J TO ERASE SCREEN AND CONTINUE : C I

I I

4 PAGE 7 0F 9 5ITE: LIMERICK I

UNIT: UI USER: MART DATE: 8/14/86 14:34 I

EFFLUENT AND WASTE DISPOSAL REPORT I

GASEOUS EFFLUENTS FOR RELEASE POINT:

4 HOT MAINTENANCE SHOP CONTINU0dS MODE BATCH NODE I

NUCLIDES

UNITS : QUARTER

GUARTER

QUARTER : QUARTER :

RELEASED 1

2 1

2

1. FISSION GASES I

AR41

CI

0.000E+00 : 0.000Ef00 : 0.000Ef00 0.000E+00 :

KR83M

CI

0.000E+00 ! 0.000Ef00 : 0.000E+00 0.000Ef00 :

I

! KR85M

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

KR05

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

I

! KR07

! CI t 0.000Ef00 ! 0.000E+00 ! 0.000Ef00 : 0.000E+00 :

I

! KR88

CI

0.000E+00 0.000Ef00 : 0.000Ef00 : 0.000E+00 :

KR89

CI

0.000E+00 0.000E+00 0.000Ef00 : 0.000E+00 :

KR90

CI

0.000Ef00 : 0.UO0Ef00 : 0.000Ef00 0.000E+00 :

XE131H

CI

0.000Ef00 1 0.000Et00 : 0.000Ef00 : 0.000Ef00 :

I

! XE133M

CI

0.000E400 : 0.000Ef00 0.000Ef00 0.000E+00 I

XE133

CI

0.000Ef00 : 0.000E+00 : 0.000Ef00 : 0.000Ef00 :

XE135M

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 0.000Ef00 :

XE135

CI

! 0.000E+00 : 0.000Ef00 : 0.000Ef00 1 0.000E+00 :

XE137

CI

0.000E+00 0.000Ef00 0.000E+00 0.000Ef00 :

I l

XE138

CI

0.000Ef00 i 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

UNIDENT.

CI

0.000E+00 1 0.000E+00 : 0.000Ef00 : 0.000Ef00 :

I TOTAL FOR :

PERIOD

! CI

0.000EiOC : 0.000Ef00 0.000Ef00 : 0.000Ef00 :

I (ABOVE)

ENTER C C 3 TO ERASE SCREEN AND CONTINUE :

C l

ll

I PAGE 8 0F 9 SITE: LIMERICK I

UNIT: U1 USER: MART DATE: 8/14/86 14:34 I

EFFLUENT AND WASTE DISPOSAL REPORT I

GASEOUS EFFLUENTS FOR RELEASE POINT: 4 HOT MAINTENANCE SHOP CONTINUOUS MODE BATCH MODE II

NUCLIDES

UNITS : QUARTER : QUARTER :

QUARTER

QUARTER :

RELEASED 1

2 1

2

2. IODINES I

1131

CI

0.000E+00 0.000Ef00 0.000E+00 : 0.000Ef00 i 1133

CI

0.000Ef00 : 0.000Ef00 0.000E+00 : 0.000Ef00 :

TOTAL FOR :

PERIOD

CI

0.000Ef00 : 0.000E+00 : 0.000Ef00 : 0.000Ef00 :

(ABOVE)

I

3. PARTICULATES I

C14

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : 0.000E+00 :

l CR51

CI

0.000Ef00 : 0.000E+00 0.000Ef00 0.000Ef00 :

HN54

! CI

0.000Ef00 0.000Ef00 : 0.000Ef00 0.000Ef00 :

I

FE59

CI

0.000E+00 0.000Ef00 1 0.000E+00 : 0.000Ef00 :

C058

CI

0.000E+00 0.000E+00 : 0.000Ef00 : 0.000E+00 :

I

C060

CI

0.000Ef00 1 0.000Et00 : 0.000Ef00 0.000Ef00 :

ZN65

CI

0.000Ef00 0.000Ef00 : 0.000Ef00 : C.000E+00 :

I SR09 CI

0.000Ef00 : 0.000Ef00 : 0.000E+00 0.000E+00 :

I

SR90

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 0.000E+00 :

ZR95

CI

0.000Et00 : 0.000Ef00 : 0.000E+00 0.000E+00 :

SB124

CI

0.000Et00 : 0.000Et00 : 0.000Et00 : 0.000Et00 :

CS134

CI

0.000Ef00 : 0.000E+00 : 0.000E+00 1 0 000E+00 :

I

CS136 CI

0.000E+00 : 0.000Ef00 0.000E+03 : 0.000E+00 :

CS137

CI

0.000E+00 : 0.000Ef00 : 0.000E+00 : 0.000Ef00 :

I l

ENTER C C J TO ERASE SCREEN AND CONTINUE : C

PAGE 9 0F 9 SITE: LINERICK UNIT: U1 I

USER: MART DATE: 8/14/86 14:35 EFFLUENT AND WASTE DISPOSAL REPORT I

GASE003 EFFLUENTS FOR RELEASE POINT: 4 HOT MAINTENANCE SHOP CONTINUOUS MODE BATCH MODE NUCLIDES : UNITS : OVARTER : QUARTER

QUARTER

QUARTER :

RELEASED :

1 2

3. PARTICULATES (CONTD)

BA140

CI

0.000Ef00 : 0.000Ef00 0.000Ef00 : 0.000E+00 :

CE141

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000Ef00 :

CE144

CI

0.000Ef00 : 0.000Ef00 : 0.000Ef00 : 0.000E+00 :

UNIDENT.

CI

0.000Ef00 : 0.000Ef00 0.000E+00 : 0.000Ef00 :

TOTAL FOR :

PERIOD

CI

0.000Ef00 1 0.000Ef00 : 0.000Ef00 : 0.000E+00 :

(ABOVE)

ENTER [ C 3 TO ERASE SCREEN AND CONTINUE :

I I

' I I

I I

I II.

TABLES i

B.

SOLID WASTE DISPOSITION REPORT I

I I

{

I I

-- ~

1 M

M M

Page 1 of 2 SOLID WASTE AND IRRADIATED FUEL SHIPMENTS PERIOD 01/01/86 TO 06/30/86 A.

SOLID WASTE SHIPPED OFFSITE FOR DURIAL OR DISPOSAL (Not irradiated fuel)

I 1.

TYPE OF WASTE I UNIT I 6--mon t h I Est. Totai l i

I l

Period i

Error,

-A I

_ g _ _ _ _ __ g ---

I a.

Spent resins, filter sludges, I m3* 1 2.84E+2 I I

I evaporator bottoms, etc.

I Ci

! 1.51E+2 1 2.50E+1 g__

--______________g______3_________g_-

._____g i

b.

Dry compressible waste, I

m3 1 0.OOE+O !

I I

contaminated equipment, etc.

I Ci I O.OOE+0 1 0.OOE+O g

g _ _ _ _ __ g -

g I

c.

Irradiated components, I

m3 1 O.OOE+O I i

i control rods, etc.

I Ci

! O.OOE+O !

O.OOE+O I

g-

-g

-- -- g - -

____g i

d.

Other (describe)

I m3 I

I I

Ci i

I I

I l

Volume in cubic meters is total container volume shipped, i

Total curie content and principal radionuclides were determined by gamma isotopic analysis of each batch of resin. Non-gamma emitting isotopes were determined by using factors scaled to principal gamma g

emitters in accordance with "AIF/NESP Methodologies for Classification of Low-Level Radioactive Wastes From Nuclear Power Plants", November, 1983.

e

M M

M i

Page 2 of 2 2.

Estimate of major nuclide composition (by type of waste) i TYPE A I

TYPE D l'~~~~~~TEPE~C~~~~

l

~~~~~IEPE'D'~~~~l

_ g _ ____

g

-- g _

___g________________

g I Isotope ! Activity i Per Cent i Activity l Per Cent ! Activity l Per Cent i Actavity i Per Cent !

I I

(Ci) l I

(C1)

I I

(Ci) 1 I

(Ci2 1

8 g

-____g.

g___

g

.g-

.g g_-

_g___

_g-

=_____

I H3 8

I I

l l

l 1 C14 8

I I

I I CR51 I

4.95E+1 8 3.28E+1 1 1

I I

i i

I I MN54 1

1.18E+0 I 7.82E-1 1 I

I I

I I FES5 I

7.86E+0 1 5.2tE+0 I I

I I

.I I C058 1

4.47E+1 1 2.97E+1 1 I

I I

I FE59 I

4.30E-1 1 2.85E-1 1 I

I I

I I NI59 I

I I

I CO6O I

3.30E+0 1 2.19E+0 1 I

I I

I I NI63 I

I I

i 1

1 1

I I

I ZN65 1

4.36E+1 1 2.89E+1 1 I

I I

ND94 1

1 I

I I

I I SR90 1

I 8

8 I

i i

i I

I NB95 1

1.49E-2 1 9.85E-3 I

ZR95 i

1.26E-2 1 8.35E-3 I I

I I

I I TC99 I

I I

RU103 I I

I I

RU106 I

I I

I I129 I

a I

I I

I i

1131 1

2.24E-2 1 1.49E-2 I I

I I

I I CS134 I

I I

I CS135 I I

I I

I I CS136 I I

I I

I I CS137 I I

I I

I i

DA140 1 4.30E-2 I 2.85E-2 I I

I I

I I CE141 3

5.57E-3 8 3.69E-3 I I

I I

I CE144 I

I I

PU239 I

I I

i l

i PU241 I

I I

I AM241 1

I I

I AM243 I

I I

I CM242 I

I I

I l

CM243 I

I I

I i

1.

l I

l____

I I

3.

Solid Waste Disposition Number of Shipments Mode of Transportation Destination 15 Truck DARNWELL 2

Truck l{ANFORD

---,am,sas-a,.

s a

m.-

a

-a.a e

, aa u-,,,am 4

1 4

4 4

III.

ATTACHMENTS A.

SUPPLEMENTAL INFORMATION 4

4 4

g 4

4 4

Attachment A Page 1 of 3 III.

ATTACHMENTS A.

SUPPLEMENTAL INFORMATION 4

Facility: Limerick Generating Station - Unit 1 License:

NPF-27 1.

Regulatory Limits (Technical Specification Limits)

A.

Noble Gases:

1. < 500 mRems/Yr - total body

" instantaneous" limits per 5 3000 mRems/yr - skin Tech Spec 3.11.2.1 2.

< 5 mRads - air gamma

- quarterly air dose limits per

_ 10 mRads - air beta Tech Spec. 3.11.2.2 3

5 10 mRads - air gamma yearly air dose limits per

_ 20 mRads - air beta Tech Spec. 3.11.2.2 B.

Iodines, tritium, particulates with half life > 8 days:

1 5 1500 mRems/yr - any organ

" instantaneous" limits per (inhalation path)

Tech Spec. 3.11.2.1

2. 5 7.5 mRems - any organ quarterly dose limits per -

Tech. Spec. 3.11.2.3

3. < 15 mRems - any organ

- yearly dose limits per C.

Liquid Effluents:

1. Concentration < 10CFR20

" instantaneous" limits per Appendix B, Ta51e II, Col. 2 Tech. Spec. 3.11.1.1 2.

< l.5 mRems - total body quarterly dose limits per

< 5 mRems - any organ Tech. Spec. 3.11.1.2

3. < 3 mRems - total body

- yearly dose limits per

< 10 mRems - any organ Tech. Spec. 3.11.1.2 2.

Maximum Permissible Concentrations MPCs are not used to calculate permissible release rates and concentrations for gaseous releases.

J The MPCs specified in 10CFR20, Appendix B, Table II, Column 2 m

for identified nuclides are used to calculate permissible release rates and concentrations for liquid releases per LGS t

Technical Specification 3.11.1.1.

I l

4

Attachment A 3.

Averago Energy 5 determination based on gaseous effluent releases for the report period is 1.603 Mev.

4.

Measurements and Approximations of Total Radioactivity A.

Fission and Activation Gases The method used is the Canberra Series 90 Counting System; GS

- Gas Marinelli B.

Iodine:

The method used is the Canberra Series 90 Counting System; CH

- Charcoal Cartridge C.

Particulate:

The method used is the Canberra Series 90 Counting System; PT

- Air Particulate Sample, 47 mm filter.

D.

Liquid Effluents:

The method used is the Canberra Series 90 Counting System and the Radwaste Liquid Discharge Pre-Release Method with a 3.5 Marinelli.

5.

Batch Releases A.

Liquid Q1 02

of Batch Releases:

33 0

2381 0

Total Time period for batch releases, 96 0

1 Maximum time period for a batch release, 72 0

Average time period for batch release, 45 0

Minimum time period for a batch release, l

Average stream flow (Schuylkill River) during periods of release of effluents 1.37E06 7.lE05 into a flowing stream, gpm

= Minutes e

1 4

AL".acnfbtnt A Page 3 of 3 B.

Gaseous 01 02

of Batch Releases:

3 2

980 1440 Total Time period for batch releases, Maximum time period for a batch release, 487 1380 327 720 Average time period for batch release, 213 60 Minimum time period for a batch release,

= Minutes 6.

Abnormal Releases A.

Liquid None B.

Gaseous None 7.

Description of LGS Release Points Release Point 1 = North Stack, Common Release Point 2 = South Stack, Unit 1 Release Point 4 = Hot Maintenance Shop 0.

Liquid Dose Description of LGS Receptor 1 = LGS Liquid Radwaste Discharge Point Receptor 2 = Citizens Home Water Company Receptor 3 = Phoenixville Water Company Receptor 4 = Philadelphia Suburban Water Company Receptor 5 = City of Philadelphia Crew Course 4

4 4

4

4 4

4 III.

ATTACHMENTS B.

CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL AND SUPPORT DOCUMENTATION 4

i 4

4 14 4

~

l4 4

4 4

III.

Attachments B.

CHANGES TO THE OFFSITE DOSE CALCULATION MANUAL AND SUPPORT DOCUMENTATION 4

Revision 4 of the Offsite Dose Calculation Manual (ODCM) is submitted for the period in which the changes became effective.

The attached document cover sheet indicates the review and approval of the change by the Engineer-in-Charge, Nuclear and Environmental Section of Mechanical Engineering, and PORC approval by the Limerick Station Manager.

The following statements have been evaluated and determined to apply to the changes:

1.

These changes will not reduce the accuracy or reliability of dose calculations or setpoint

{

determinations.

Justifications are presented in tabular form to facilitate their interpretation.

2.

The most significant tachnical improvement to the ODCM is the inclusion of "SF",

the structural a

shielding and occupancy factor, into the Noble Gas g

Total Body dose calculation (ODCM, Rev. 4, Page 5; j

Technical Specification 4.11.2.1.1).

This addition is consistent with the guidance provided in Reg.

J Guide 1.109, Appendix E - Other Parameters.

5 Assuming the maximum individual spends about 50 percent of the time indoors, the overall shielding and occupancy factor is then approximately 0.7.

The factor of 0.5 is used directly for population dose calculations.

These factors are applicable for external gamma exposure from noble gas and for external exposure from contaminated ground surfaces.

Although "SF" was inadvertently omitted from the ODCM document, all Technical Specification Noble Gas Total Body dose calculations were performed using the factor of 0.7.

4 4

4' 4

4 4

Page 2 of 7-III.

ATTACHMENTS B.

CHANGES TO THE OFFISTE DOSE CALCULATION MANUAL AND SUPPORT DOCUMENTATION PAGE FROM TO REASON FOR CHANGE Table of Contents

a. Previous descriptions did not accurately reflect contents.

b. Add 2 sections, VI-B & VI-C 1.

.... Operating 39 5% Power License License NPF-27 (NPP-27) replaced with Full Power License (NPF-39) 1.

..in 10CFR20.106(a)

..in 10CPR20.106(a) correction for radionuclides for radionuclides other than the other than the dissolved or dissolved or entrained noble entrained noble gases and the gases.

For concentration dissolved or listed in Tech.

entrained Spec. Table noble gases,the 3.11.1.1-1 for all concentration dissolved or shall be limited entrained noble to the value gases as specified stated in Tech.

in Tech. Spec.

Spec. 3.11.1.1 3.11.1.1.

1.

Table 4.11-1 4.11.1.1.1-1 correction 2.

MPCi... listed in 3.11.1.1 correction Tech. Spec. Table 3.11.1.1-1 E

2.

The Ai values used Table II.A.1 correction for this of this calculation document are located in the Appendix, Table 1 2.

D DO reduce transcription errors 2,3 Ai AiO clarification and h

reduce transcription errors l

J' 2,3 tl dt1 reduce transcription B

errors l

l 2,3 to the total organ, O, from reduce transcription body or any organ, errors from liquid

Page 3 of 7 PAGE FROM TO REASON FOR CHANGE 3.

t dtl reduce transcription errors 4.

Ai Dose Factor AiO reduce transcription errors 4.

REFERENCE:

Provide source See B4SES - Note 1 of data 4

~

5.

.. 1975 to 1976...

.. 1976 which is correction

...most recent updated annually update was in 1983 5.

the location is the No change moved from below the site boundary,....

equation to the

... noble gas introduction releases 5.

DTB=i(Ki(X/Q)v Qiv)

DTB=

a. Addition of the 1(Ki(X/Q)v Qiv)SF structural shielding (lE06) and occupancy factor per R.G.

1.109, Appendix E -

Other parameters

b. Addition of unit ag conversion factor previously omitted.

6.

lE06 = unit Addition conversion, pCi/nCi 6.

a.

Isotopic SF = structural Addition Analysis Method shielding and occupancy factor; 0.7 (maximum individuals) and 0.5 (general population. Taken from R.G.

1.109, Appendix E - Other Parameters 6.

lE06 = unit Addition of unit conversion, pCi/uCi conversion factor previously omitted 6.

b. GROSS RELEASE No change moved to the METHOD - The introduction on pg. 5 location is the site boundary 790 NE from noble gas releases 4

4

Page 4 of 7 PAGE FROM TO REASON FOR CHANGE 6.

b. Gross SF = structural Addition Release Method shielding and occupancy factor; 0.7 (maximum individuals) and 0.5 (general population. Taken from R.G.

1.109, Appendix E - Other Parameters 7.

The location is 790m NE

a. correction the site boundary,

b. moved to the 762 ESE from introduction the vents E

8.

The location is 790m NE

a. correction the site boundary,

b. moved to the 762 ESE from introduction

... noble gas releases 8.

(X/Q)V (X/Q)v

a. corrected to Oiv Qiv become consistent with equations 9.

b. GROSS RELEASE deleted included in METHOD - The introduction on pg. 8 location is the noble gas releases 9.

2. BETA AIR DOSE deleted included in The location is introduction on pg. 8 the noble gas releases E

10.

b. GROSS RELEASE deleted included in METHOD - The introduction on pg. 8 the location is E

the... noble gas releases J

10.

age groups.

deleted removes the 3

If the computer is restriction not available, for use the following expression will be used:

10.

Location is the no change relocated to critical pathway introduction for III C

=

dairy ll70m ESE from vents 4

4

Page 5 of 7 PAGE FROM TO REASON FOR CHANGE 10.

D=3.17E09 (CF)

D=3.17E08 (CF) correction Qiv)

.... Qiv) 14.

a. Surv. Req.

4.11.4.1 Tech. Specs.

3.11.1.2.a 3.ll.l.2a corrections 3.ll.2.b 3.11.2b 3.11.2.a 3.11.2.2a 3.11.2.2.b 3.ll.2.2b 3.11.2.3.a 3.11.2.3a 3.ll.2.3.b 3.ll.2.3b 14.

b. Sur. Req.

4.11.4.2 g

Tech. Specs.

g 3.11.1.2.a 3.ll.l.2a corrections 3.ll.l.2.b 3.11.1.2b 3.11.2.2.a 3.11.2.2a J

3.11.2.2.b 3.ll.2.2b a

3.11.2.3.a 3.ll.2.3a 3.ll.2.3.b 3.ll.2.3b 15.

V.A.

Unique Semi-Annual Title changed to Reporting Radioactive more accurately Requirement Effluent Release reflect contents (6.9.1.8)-Dose Report Calculations for the Radioactive Effluent Release Report I

15.

VI.A. Sur. Req.

VI. Radiological to become 4.12.1 Environmental consistent with l

Monitoring Table of Contents l

Program B.

Surv. Req.

Addition 15.

l 4.12.2 16.

VI.A. Sur. Req.

C.

Surv. Req.

Addition 4.12.3 4.12.3 18.

Birch Substation SH1 correction i

(Control) SN1 i

18.

Linfield Substation 17B1 correction 17D1 i

18.

2301 Market St.

13H4 correction Phila. PA (Control) 13N4 19.

Fish-16C3 16C5 correction 4

Page 7 of 7 PAGE FROM TO REASON FOR CHANGE 31.

B5. Hot Maint.

1.

IODINE entire sentence Shop Setpoint moved to the j

Determination introduction for W

l. The iodine high section B5 alarm setpoint is set to alarm in the event that 10CFR20 dose rates at the site boundary are approached or exceeded 32.

B4. Containment

3. PARTICULATE Consolidated into Purge the introduction

3. The particulate for section B5 high alarm setpoint is set to alarm in the event that 10CFR20 dose rates at the site boundary are approached or exceeded 35.

Site Specific AiO correction Data Ai 38.

The model Technical

... Specification Clarification Specification LCO (Limiting LCO for all Condition for Operation) for all 40.

Figure IX.A.1 No change Figure replaced with more legible version.

41.

Figure IX.A.2 Figure IX.A.2 1 Figure broken into 2 Figures when Figure IX.A.3 A.2 replace by (p. 43)

A.2 AND A.3 4

4 4'

4 4

4

OFFSITE DOSE CALCULATION MANUAL l

Revision 4 LIMERICK GENERATING STATION UNITS 1 AND 2

[

{

PHILADELPHIA ELECTRIC COMPANY Docket Nos. 50-352 & 50-353 9

PORC Approval:

[

Station Kanager t

A ro

/

E.

I.

I, 8['

LGS Health Physics' Representative

/

Nuclear and Environmental Representative:

8[

/h4 k

c_

I Table of Contents Page I

Purpose 1

II.

Liquid Pathway Dose Calculations A.

Surveillance Requirement 4.11.1.1.2 1

B.

Surveillance Requirement 4.11.1.2 2

C.

Surveillance Requirement 4.11.1.3.1 3

III.

Gaseous Pathway Dose Calculations A.

Surveillance Requirement 4.11.2.1.1 5

B.

Surveillance Requirement 4.11.2.2 8

C.

Surveillance Requirement 4.11.2.3 11 D.

Surveillance Requirement 4.11.2.5.1 12 l IV.

Annual Dose A.

Surveillance Requirement 4.11.4.1 14 B.

Surveillance Requirement 4.11.4.2 14 l

V.

Semi-Annual Radioactive Effluent Release Report l

A.

Surveillance Requirement 6.9.1.8 15 VI.

Radiological Environmental Monitoring Program A.

Surveillance Requirement 4.12.1 15 l

B.

Surveillance Requirement 4.12.2 15 l

C.

Surveillance Requirement 4.12.3 15 VII.

Effluent Radiation Monitor Setpoint Calculations 24 l

A.

Liquid Effluents l

B.

Gaseous Effluents VIII. Bases 33 IX.

Liquid and Gaseous Effluent Flow Diagrams 38 g

ssex re HP h b DATE h / M Np I

I I.

Purpose The purpose of the Offsite Dose Calculation Manual is to establish methodologies and procedures for I

calculating doses to individuals in areas at and beyond the SITE BOUNDARY due to radioactive effluent from Limerick Generating Station and establishing setpoints for radioactive effluent monitoring I

instrumentation.

The results of these calculations are required to determine compliance with Appendix A l

to Operating License NPF-39, " Technical Specification and Bases for Limerick Generating Station Unit No. 1.

II.

Liquid Pathway Dose Calculations A.

Surveillance Requirement 4.11.1.1.2 - Liquid Radwaste Release Compliance with 10CFR20 Limits I

Limerick Generating Station Units 1 and 2 have one common discharge point for liquid releases under normal circumstances.

In the event of heat I

exchanger leakage, additional release pathways are possible through the plant service water system and the RHR service water system.

The following calculation assures that the radwaste I

release limits are met.

The flow rate of liquid radwaste released from the site to areas at and beyond the SITE BOUNDARY shall be such that the I

concentration of radioactive material after dilution shall be limited to the concentration specified in 10 CFR 20.106 (a) for radionuclides other than the dissolved or entrained noble I

gases.

For dissolved or entrained noble gases, the concentration shall be limited to the value stated in Technical Specification 3.11.1.1.

Each I

tank of radioactive waste is sampled prior to release and is quantitatively analyzed for identifiable gamma emitters as specified in Table l

4.11.1.1.1-1 of the Technical Specification.

I From this gamma isotopic analysis the maximum permissible release flow rate is determined as follows:

I Determine a Dilution Factor by:

Dilution Factor (uCi/ml i)

I

=

1

(

MPCi

)

uCi/ml i = the activity of each identified gamma emitter in uCi/ml ENGR ffY

[khh(b HP DATE &/3-&

l I

L b

MPCi = The MPC specified in lb CPR 20, Appendix B, Table II, Column 2 for radionuclides other f

than dissolved or entrained noble gases or L

the concentrations listed in Technical l

Specification 3.11.1.1 for dissolved or entrained noble gases.

Any unidentified

[,

conce,ntration is assigned an MPC value of 1E-07 uci/ml.

Determine the Maximum Permissible Release Rate with this Dilution Factor by:

A l

Release Rate (gpm)

=

B (Dilution Factor)

A = The cooling tower blowdown volume which will provide dilution.

Maximum flow rate is 10,000 gpm.

B = margin of assurance which includes consideration of the maximum error fn the activity setpoint and the maximum error in the flow setpoint and the possibility of multiple release pathways.

B.

Surveillance Requirement 4.11.1.2 The primary method of calculating dose contributions from liquid effluents released to areas at or beyond the SITE BOUNDARY will be by using a computer-based calculational program developed using the equations and parameters of R.G.

1.109, Rev. 1, October, 1977 (see bases Note

4) for all organs and age groups.

The Ai values

[

used for this calculation are located in Table II. A.1 of this document.

Dose contributions from liquid effluents released to areas at and beyond

(

the SITE BOUNDARY shall be calculated using the L

equation below.

This dose calculation uses as a minimum those appropriate radionuclides listed in Table II.A.1.

These radionuclides account for

(

virtually 100 percent of the total body dose and bone dose from liquid effluents.

l DO

= )f Ri i(AiO 1('dtl) (Cil) (F1) i=1 l

DO

= the cumulative dose commitment to the total body or any organ, 0,

from liquid effluents for the total time period Ili= $m dtl, in mrem Ri

= reported release points l

dtl

= the length of the first time period over which Cil and F1 are averaged for the liquid release, in hours. ENGR. Je-

[

H.P., sf(p9C L

DATE 8 - / 2 Fb F

t-

I I

- Cil

= the average concentration of radionuclide, i, in undiluted liquid effluent during time period dt from any liquid release, (determined by the effluent sampling analysis program, Technical Specification Table 4.11.1.1-1), in uci/ml.

l AiO

= the site related ingestion dose commitment factor to the total body or organ, O, for I

each radionuclide listed in Table II.A.1, in mrem-ml per hr-uCi.

See Site Specific Data.**

I F1

= the near field average dilution factor for Cil during any liquid effluent release.

Defined as the ratio of the maximum undiluted liquid waste flow during release to the average I

flow from the discharge structure to the Schuylkill River.

II.C Surveillance Requirement 4.11.1.3.1 Projected dose contributions from liquid effluents I

shall be calculated using the methodology described in Section II.B.

To estimate expected concentration of the various radionuclides (Cil) in the undiluted liquid effluent, the duration of liquid release-(dtl),

I and the near field average dilution factor (F1), the expected plant operating status shall be reviewed.

If no operational changes are expected which would affect I

Cil, dtl, or F1 the same values as used to evaluate Section II.B may be used.

If any operational changes are expec.ted during the following 31 days which could affect Cil, dtl or F1, the values used shall be based I

on plant history.

During the initial stages of plant operation, the values for Cil, dtl, and F1 as given in LGS FSAR Section 11.2 and EROL Section 5.2 may be used.

I See Note 1 in Bases I

I I

I I ENGR. Jge H.P.. A1ALL DATE 8-/ 3 -Bo I

TABLE II.A.1 I

LIQUID EFFLUENT INGESTION DOSE FACTORS (Decay Corrected) l AiO Dose Factor (mrem-ml per hr-uci)

Radionuclide Total Body Bone I

Cs-137 3.42E+05 3.82E+05 Cs-134 5.79E+05 2.98E+05 P-32 5.11E+04 2.05E+05 Cs-136 8.42E+04 2.97E+04

'I Zn-65 3.32E+04 2.31E+04 Sr-90 1.35E+05 5.52E+05 H-3 3.29E-01 I

Na-24 1.35E+02 1.35E+02 I-131 1.16E+02 1.40E+02 Co-60 5.70E+02 I-133 1.23E+01 2.31E-01 I

Fe-55 1.06E+02 6.61E+02 Sr-89 6.36E+02 2.21E+04 Te-129m 1.70E+03 1.08E+04 I

Mn-54 8.34E+02 8.34E+02 Co-58 2.00E+02 Fe-59 9.26E+02 1.02E+03

'g Te-131m 3.88E+02 9.53E+02 3

Ba-140 1.33E+01 2.03E+02 Te-132 1.21E+03 1.99E+03 I

NO'E: The listed dose factors are for radionuclides that may T

be detected in liquid effluents and have significant dose consequences.

These factors are decayed for one I

day to account for the time between effluent release and ingestion of fish by the maximum exposed individual, an adult.

There is no bone dose factor given in R.G.

1.109 for these nuclides.

I g

l REFERENCE - See BASES - NOTE 1 I

I ENGR. Jfe I

H.P.

, }/1 fr/r' DATE B Y/3-5(o I

III.

Gassoun Pathway Dose Calculations I

The controling receptor locations for the gaseous pathway dose calculations are based on a land-use census performed in 1975 to 1976 which is updated annually.

A.

Surveillance Requirement 4.11.2.1.1 I

The dose rate in areas at and beyond the SITE BOUNDARY due to radioactive materials released in gaseous effluents shall be determined by the expressions below:

1.

NOBLE GASES The dose rate from radioactive noble gas releases shall be determined by either of two methods.

Method (a), the Isotopic Analysis Method, I

utilizer the results of noble gas analysis required by specification 4.11.2.1.1 and 4.11.2.1.2.

Method (b), the Gross Release Method, assumes that all noble gases released are I

the most limiting nuclide-Kr-88 for total body dose and Kr-87 for skin dose.

For normal operations, it is expected that method (a) will I

be used.

However, if isotopic release data are not available method (b) can be used.

Method (a) allows more operating flexibility by using data that more accurately reflect actual releases.

I The location is the site boundary, 790m NE from the vents.

This location results in tne highest calculated dose to an individual from noble gas releases.

a.

ISOTOPIC ANALYSIS METHOD b i(Ki (X/Q)v Qiv) SF (lE06) l DTB

=

l Ds

= [ i((Li + 1.lMi) (X/Q)v) (lE06)

DTB

= total body dose rate, in mrem /yr.

Ds

= skin dose, in mrem /yr.

Ki

= the total body dose factor due to gamma emissions for each identified noble gas radionuclide.

Values are listed3" Table III.A.1, mrem /yr per uCi/m I

(X/Q)v

= 1.lE-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).

I ENGR. JdSC _

H.P. t]/{ M DATE h -/3 - %

I

I Qiv

= the release rate of noble gas radionuclide, i, in gaseous ef fluents from all vent releases determined by isotopic analysis I

averaged over one hour, uCi/sec.

SF

= Structural shielding and occupancy factor; 0.7 (maximum individuals) and 0.5 (general population).

Taken from I

R.G.

1.109, Appendix E - Other Parameters Li

= the skin dose factor due to beta emissions for each identified noble gas radionuclide.

I Values are listed on Table III.A.1, mrem /yr per uCi/m3.

Mi

= the air dose factor due to gamma emissions for each identified noble gas radionuclide.

I Values are listed on Table III.A.1, mrad /yr per uCi/m3.

1.1

= unit conversion, converts air dose to skin dose, mrem / mrad.

l lE06 unit conversion, pCi/uCi

=

b.

GROSS RELEASE METHOD l

DTB

=K (X/Q)V (Onv) SF I

(L + 1. lM) (X/Q) hnv Ds

=

I I

DTB

= total body dose rate, mrem /yr.

Ds

= skin dose rate, mrem yr.

K

= 1.47E04 mrem /yr per uCi/m3; the total body dose factor due to gamma emissions I

for Kr-88 (Reg. Guide 1.109, Table B-1).

I (X/Q) v

= 1.18-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).

I (h) nv the gross release rate of noble gases in

=

gaseous effluents from vent releases determined I

by gross activity vent monitors averaged over one hour, in uCi/sec.

SF

= Structural shielding and occupancy factor; 0.7 (maximum individuals) and 0.5 I

(general population).

Taken from R.G.

1.109, Appendix E - Other Parameters L

= 9.73E03 mrem /yr per uCi/m3; the skin I

dose factor due to beta emissions for Kr-87 (Reg. Guide 1.109, Table B-1).

6.17E03 mrad /yr per uCi/m3;'the air I

M

=

dose factor due to gamma emissions for Kr-87 (Reg. Guide 1.109, Table B-1).,

i ENGR.

Ee" i H.P. t AM

!g DATE T/ 3-lb

I 2.

Tha primary mnthod of calculating dose contribution from Iodine-131, Iodine-133, tritium, and radioactive material in particulate form, other than noble gases, with half-lives greater than eight days will be by using a I

computer-based calculational program developed using the equations and parameters of R.G.

1.109, Rev.

1, October, 1977 (see bases Note 4) for all organs and age groups.

If the computer model is I

not available, the dose contributions from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form, other than noble I

gases, with half-lives greater than eight days will be calculated using the equation below.

The l

location is the site boundary, 790m NE from the vents.

DT

= (CF)1 Pi (WV (Qiv))

DT

= dose rate to the thyroid, mrem /yr.

CF

= 1.02; the correction factor accounting for the use of iodine-131 and iodine-I 133 in lieu of all radionuclides released in gaseous effluents.

I P

= 1.62E07 mrem /yr per uCi/m3; the inhalation I-131 dose parameter for I-131 inhalation pathway.

The dose factor is based on the critical individual organ, thyroid, and most restrictive I

age group, child.

(Reg. Guide 1.109 Tables E-5 and E-9).**

P

= 3.85E06 mrem /yr per uCi/m3; the inhalation I-133 dose parameter for I-133 inhalation pathway.

The dose factor is based on the critical individual organ, thyroid, and most restrictive age group, child.

(Reg. Guide 1.109 Tables E-5 and E-9).**

WV

= 1.00E-05 sec/m3; the highest calculated annual average relative concentration for any area at or beyond the SITE BOUNDARY for all vent releases (NE boundary).

hiv

= the release rate of iod'ine-131 and/or I

iodine-133 in gaseous effluents from all vent releases, determined by the effluent sampling and analysis program (Technical I

Specification Table 4.8.2) in uCi/sec.

'I ENGR. M H.P.

/AAC DATE 6-/ 3 - S (a

I III.B Surveillance Requirement 4.11.2.2 I

The air dose in areas at and beyond the SITE BOUNDARY due to noble gases released in gaseous effluents shall I

be determined by the expressions below.

The dose rate from radioactive noble gas releases shall be determined by either of two methods.

Method (a), the Isotopic Analysis Method, utilizes the results of noble gas analysis required by specification 4.11.2.1.1 and 4.11.2.1.2.

Method (b), the Gross Release Method, assumes that all noble gases released I

are the most limiting nuclide - Kr-88 for total body dose and Kr-87 for skin dose.

For normal operations, it is expected that method (a) will be used.

However, if isotopic release data are not available, method (b)

I can be used.

Method (a) allows more operating flexibility by using data that more accurately reflects actual releases.

The location is the SITE I

l BOUNDARY, 790m NE from the vents.

This location results in the highest calculated gamma air dose from noble gas releases.

I l

1.

GAMMA AIR DOSE a)

ISOTOPIC ANALYSIS METHOD l

DG

= 3.17E-08 i(Mi(X/Q)v Qiv)

DG

= gamma air dose, mrad.

3.17E-08= years per second.

Mi

= the air dose factor due to gamma emissions for each identified noble gas radionuclide.

Values are listed on Table III.A.1, mrad /yr per uCi/m3.

l (X/Q)v

= 1.lE-05 sec/m3; the highest calculated average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

l Qiv

= the release of noble gas radionuclides, i, in gaceous effluents from all vents as determined by isotopic analysis, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

I ENGR. JV2(#

I H.P. ) \\ 4 /. d[O DATE

&/p I

I b.

GROSS RELEASE METHOD DG

= 3.17E-08 (M

(X/Q)v Qv )

I DG

= gamma air dose, in mrad.

3.17E-08= years per second.

M

= 1.52E04 mrad /yr per uCi/m3; the air dose factor due to gamma emissions for I

Kr-88 (Reg. Guide 1.109, Table B-1).

(X/Q)v

= 1.lE-05 sec/m3; the highest calculated I

annual average relative coacentration from vent releases for any area at or beyond the SITE BOUNDARY.

Qv

= the gross release of noble gas radio--

nuclides in gaseous effluents from all vents, determined by gross activity vent I

monitors, in uCi.

Releases shall be cumulative over the calendar quarter or year as appropriate.

2.

BETA AIR DOSE a.

ISOTOPIC ANALYSIS

=3.17E-08[i(Ni (X/Q)v Qiv)

DB 3.17E-08 = years per second.

Ni

= the air dose factor due to beta emissions for each identified noble gas radionuclide.

l See Table III.A.1, mrad /yr per uCi/m3.

(X/Q)v

= 1.lE-05 sec/m3; the highest calculated annual average relative concentration from vent releases for any area at or beyond the SITE BOUNDARY.

Qiv

= the release of noble gas radionuclide, i, in gaseous effluents from all vents as determined by isotopic analysis, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

I I ENGR. M

/V)(L(,l H.P.

8-/Pfe DATE I

b.

G_ROSS REL_ EASE METHOD DB e 3.17E-08

$1(X/Q)vQv

-l DB

= beta air dose, in mrad.

3.17E-08 = years per second.

1.03E04 mrad /yr per uCi/m3; the air dose N

=

factor due to beta emissions for Kr-87 (Reg.

Guide 1.109, Table B-1).

(X/Q)v

= 1.lE-05 sec/m3; the highest calculated annual average relative concentration from I

vent releases for any area at or beyond the SITE BOUNDARY.

Qv

= the gross release of noble gas radionuclides I

in gaseous effluents from all vents determined by gross activity vent monitors, in uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

III.C Surveillance Requirement 4.11.2.3 The primary method of calculating dose to an individual from Iodine-131, Iodine-133, tritium, and radioactive materials in particulate form, other than I

noble gases, with half-lives greater than eight days in gaseous effluents released to areas at and beyond the SITE BOUNDARY, will be by using a computer-based calculational program developed using the equations and parameters of R.G.

1.109, Rev. 1, October, 1977 (see Bases Note 4) for all organs and age groups.

Location is the critical pathway dairy 1770m ESE from the vents.

l D = 3.172-08 '(CF)

(0.5)[I(Ri WV Qiv) critical organ dose, thyroid, from all D

=

pathways, in mrem.

l 3.17E-08.= years per second.

CF

= 1.00; the correction factor accounting for the use of Iodine-131 and Iodine-133 in lieu of all radionuclides released in gaseous effluents.

0.5

= fraction of iodine releases which are nonelemental.

I ENGR. M ll H.P.

((]()('

lB DATE M/3-/I7 1 I

R

= 9.51 Ell m2 (mrem /yr) par (uCi/sec); the doso I-131 factor for Iodins-131.

The dono factor in based on the critical individual organ, thryoid, and l

most restrictive age group, infant.

See Site Specific Data.**

R

= 8.13E09 m2 (mrem /yr) per uCi/sec; the dose I-133 factor for Iodine-133.

The dose factor is based on the critical individual organ, thyroid, and most restrictive age group, infant.

See Site Specific Data.**

WV

= 1.82E-9/m2 (D/Q) for the food pathway for vent releases.

Qiv

= the release of Iodine-131 and/or Iodine-133 I

determined by Technical Specification Table 4.11.2.1.2-1, uCi.

Releases shall be cumulative over the calendar quarter or year, as appropriate.

III.D Surveillance Requirement 4.11.2.5.1 I

The projected dosec from releases of gaseous effluents to areas at and beyond the SITE BOUNDARY shall be calculated in accordance with the following sections of this manual:

a.

gamma air dose - III.B.1 b.

beta air dose

- III.B.2 c.

organ dose

- III.C The projected dose calculation shall be based on expected releases from plant operation.

The normal release pathways result in the maximum releases from the plant.

Any alternative release pathways result in lower releases and therefore lower doses.

To estimate the expected releases of noble gases and radioiodines in gaseous effluents, the expected plant operating status.shall be reviewed.

If no operational changes are expected which would affect the magnitude or type of releases the same values used to evaluate Sections III.B.1, III.B.2 and III.C may be used. lI 1

1

VII.

Efflu!nt Rndiation Monitor Sntpoint Calculations A.

Liquid Effluents Al.

Radwaste Discharge Line Radiation Monitor -

g Monitor alarm setpoints will be determined in

~E oraer to assure compliance with 10CFR20.

The setpoints will indicate if the concentration of radionuclides in the liquid effluent at the site I

boundary is approaching the concentrations specified in 10CFR20, Appendix B, Table II, Column 2 for radionuclides other than dissolved or entrained noble gases.

The setpoints will 8

also assure that concentrations listed on l

Technical Specification 3.11.1.1 for dissolved ~or entrained noble gases is not exceeded.

The i

following method applies to liquid releases from the plant via the cooling tower blowdown line when determining the high-high alarm setpoint for the Liquid Radwaste Ef fluent Monitor during all 5

operational conditions.

When the high-high alarm setpoint is reached or exceeded, the releases will be automatically terminated.

l a.

DETERMINE Ct i

Ct =

CiD (Fi) 5 (Ci/MPCi)

Ct = concentration at the liquid radwaste discharge I

line monitor prior to dilution (to assure 10CFR20.106 limits are not exceeded) ; uCi/cc, 1

Ci = total concentration of liquid effluent discharge discharge prior to dilution with cooling tower blowdown; uCi/cc 5=

margin of safety factor including Fi uncertainty, to asecre that the high-high alarm will terminate the discharge before I

10CFR20 limits are exceeded.

Ci

.= sum of the ratio of the isotopic concentrations MPCi divided by their respective MPC.

D=

dilution factor due to blowdown from the cooling tower; calculated by dividing the total flow (cooling tower blowdown plus radwaste discharge flow) by the radwaste discharge flow.

Fi = Ratio of MPC-weighted releases in the liquid radwaste effluent monitor flow path divided by the total MPC-weighted liquid releases; I

ENGR. M-H.P.

44 /K ' > pxtg [p43.M(o

If any operational changes are expected during the following 31 days which could affect the magnitude or I

type of releases, the values used shall be based on plant history.

During the initial stages of plant operation the values for releases expected as given in LGS FSAR Section 11.3 may be used.

See Note 3 in Bases I

I I

I

'I 'I

b TABLE III.A.1 DOSE FACTORS FOR EXPOSURE TO A SEMI-INFINITE CLOUD OF NOBLE GASES

(-

Nuclide B-air *(Ni)

B-Skin **(Li)

G-Air *(Mi)

G-Body **(Ki)

{

Kr-83m 2.88E-04 1.93E-05 7.56E-08 Kr-85m 1.97E-03 1.46E-03 1.23E-03 1.17E-03

(

Kr-85 1.95E-03 1.34E-03 1.72E-05 1.61E-05 Kr-87 1.03E-02 9.73E-03 6.17E-03 5.92E-03 Kr-88 2.93E-03 2.37E-03 1.52E-02 1.47E-02 Kr-89 1.06E-02 1.01E-02 1.73E-02 1.66E-02 Kr-90 7.83E-03 7.29E-03 1.63E-02 1.56E-02

{

Xe-131m 1.llE-03 4.76E-04 1.56E-04 9.15E-05 Xe-133m 1.48E-03 9.94E-04 3.27E-04 2.51E-04 l

b Xe-133 1.05E-03 3.06E-04 3.53E-04 2.94E-04 Xe-135m 7.39E-04 7.llE-04 3.36E-03 3.12E-03

[

Xe-135 2.46E-03 1.86E-03 1.92E-03 1.81E-03 Xe-137 1.27E-02 1.22E-02 1.51E-03 1.42E-03 Xe-138 4.75E-03 4.13E-03 9.21E-03 8.83E-03 Ar-41 3.28E-03 2.69E-03 9.30E-03 8.84E-03

[

mrad-m3 pCi yr

- mrem-m3

{

PCi yr

REFERENCE:

Regulatory Guide 1.109, Revision 1, October 1977

[ [

r

[

~ - ~ " ~ ~ ~

IV.

TOTAL DOSE A.

Surveillance Requirement 4.11.4.1 I

If the doses as calculated by the equations in this manual do not exceed the limits given in Technical Specificatlons 3.ll.1.2a, 3.11.25, 3.ll.2.2a, l

3.11.2.2b, 3.ll.2.3a, or 3.11.2.3b by more than two I

times, the conditions of Technical Specification 3.11.4.2 have been met.

B.

Surveillance Requirement 4.11.4.2 If the doses as calculated by the equations in this manual exceed the limits given in Technical Specifications 3.11.1.2a, 3.11.1.2b, 3.ll.2.2a, 3.11.2.26, 3.11.2.3a, or 3.11.2.3b by more than two times, the maximum dose or dose commitment to a real I

individual shall be determined utilizing the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",

Revision 1, October 1977.

Any deviations from the methodology provided in Regulatory Guide 1.109 shall be documented in the Special Report to be prepared in accordance with Technical Specification 3.11.4.1.

The cumulative dose contribution from direct radiation from the. two reactors at the site and from radwaste storage shall be determined by the following method:

Cumulative dose contribution from direct radiation =

Total dose at the site of interest (as evaluated by TLD measurement)

Mean of background dose (as evaluated by TLD's at background sites)

Effluent contribution to dose (as evaluated above).

This method is used only to evaluate the contribution from direct radiation dose.

The direct radiation dose is then added to the dose or dose commitment determined in accordance with the methods in ute first paragraph above to determine total dose from all pathways.

This evaluation is in accordance with ANSI /ANS 6.6.1-1979 Section 7.

The error using this method is estimated to be approximately 8%.

l ENGR. E&"

H.P.

/VFEC WE M Wp -

I l V.

SEMI-ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT l

A.

Surveillance Requirement 6.9.1.8 The assessment of radiation doses for the radiation dose assessment report shall be performed utilizing the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",

Revision 1, October 1977.

Any deviations from the methodology provided in Regulatory Guide 1.109 shall be documented in the radiation dose assessment report.

The meteorological conditions concurrent with the time of release of radioactive materials (as' determined by sampling frequency of measurement) or approximate I

methods shall be used as input to the dose model.

The Radioactive Effluent Release Report shall be submitted within 60 days after January 1 of each year.

l VI.

RADIOLOGICAL ENVIRONMENT MONITORING PROGRAM l

A.

Surveillance Requirement 4.12.1 The radiological environmental monitoring samples shall be collected pursuant to Table VI.A.1 from the I

locations shown on Figures VI.A.1, VI.A.2, VI.A.3 and VI.A.4, and shall be analyzed pursuant to the requirements of Table 3.12.1-1 of the LGS Technical Specifications.

l B.

Surveillance Requirement 4.12.2 I

Pursuant to Technical Specification 4.12.2, the land use census shall be conducted during the growing season at least once per 12 months using information that will provide the best results, such as door-to-door survey, aerial survey, or by consulting local agriculture authorities.

The results shall be included in the Annual Radiological Environmental I

Operating Report (pursuant to Technical Specification 6.9.1.7).

I I

~

I ENGR. 3/Ase' H.P.

/ML DATE

/7'13 '8[:r J L

I C.

Survaillance Rcquirement 4.12.3 l

Pursuant to Technical. Specification 4.12.3, the laboratory performing the radiological environmental analyses shall participate in an interlaboratory I

comparison program which has been approved by the NRC.

This program is the Environmental Protection Agency's (EPA's) Environmental Laboratory Intercomparison Studies (cross check) Program.

Our participation code is CJ.

Participation includes all of the determinations (sample medium-radionuclide combination) that are offered by the EPA and that are I

also included in the monitoring program.

The results of the analysis of these (cross check) samples will be included in the Annual Radiological Environmental Operating Report.

1 I

I I

I l ENGR. >w l

H.P.

(V]Q f)

DATE Bd3 -6[o I

i _ - ________

f%

E E

N O

N~

E I

e TABLE VI.A.1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM

EXPOSURE PATHWAY NUMBER OF SAMPLES AND STATION STATION DISTANCE AND/OR SAMPLE SAMPLE STATION NAME CODE SECTOR (MILES)

COMMENTS Direct 40 Locations (a) TLD sites were chosen in accordance Radiation (a)

INNER RING LOCATIONS with Limerick Generating Station's

1) Evergreen & Sanatoga Roads 3651 N

O.6 Technical Specifications Table

2) Sanatoga Road 351 NNE O.6 3.12-7, Item 1.

The inner ring

3) Possum Hollow Road SSI NE O.4 and outer ring stations cover

4) LGS Training Center 751 FNE O.5 all sectors.

5) Keen Road 1053 L

O.5

6) LGS Information Center 1151 ESE O.5 The control and special interest

7) Longvtew Road, SE Sector 14S1 SE O.6 stations provide information on Site Boundary population centers and other B) Longvlaw Road. SSE Sector 1652 SSE O.6 special interest locations.

Site Boundary

9) Railroad Track Along 1851 S

0.3 Longvlew Road IO) Impounding Basin. SSW 21S1 SSW O.5 Sector Site Boundary

11) Transmission Tower, WSW 2352 WSW O.5 Sector Site Boundary

12) SW Sector, Site Boundary 2551 SW O.5

13) Meteorological. Tower 2 Site 26S3 W

O.4

14) WNW Soctor Site Boundary 29S1 WNW O.5

15) NW Sector Site Boundary 3251 NW O.6

16) Meteorological Tower 1 Site 3452 NNW O.6 OUTER RING LOCATIONS

1) Ringing Rock Substation 35F1 N

4.2

2) Laughing Waters GSC 2E1 NNE 5.1

3) Netffer Road 4EI NE 4.6 l

4) Pheasant Road, Game Farm 7El ENE 4.2 Site

5) Transmission Corrider 10EI E

3.9

6) Trappe Substation 10F3 ESE 5.5

7) Vaughn Substation 13EI SE 4.3

8) Pikeland Substation 16F1 SSE 4.9

9) Showden Substation 1901 S

3.6

10) Sheader Substation 20F1 SSW 5.2 i

. i

I &

Y! &

Y Y} &

S Y) i OUTER RING LOCATIONS (Cont'd)

11) Porter's Mtil Substation 2401 SW 3.9

12) Transmission Corrider.

25D1 WSW 4.0 i

Hoffecker and Keim Streets l

13) Transmission Corrider.

2802 W

3.8 i

W.

Cedarville Road

14) Prince Street 29El WNW 4.9

15) Poplar Substation 3102 NW 3.9

16) Varnell Road 34E1 NNW 4.6 CONTROL AND SPECIAL INTEREST LOCATIONS l

1) Birch Substation (control) 5H1 NE 25.8

2) Pottstown Landing Field 6C1 ENE 2.1

3) Reed Road 9C1 E

2.2

4) King Road 13C1 SE 2.9

5) Spring City Substation 15D1 SE 3.2 l

6) Linfield Substation 1781 S

1.6

7) Ellis Woods Road 2001 SSW 3.1

8) Lincoln Substation 31D1 NW 3.0 5 LOCATIONS

2. Airborne

1) Mean Road' 1053 E

0.5 (b) These stations provide for

2) LGS Information Center 1151 ESE O.5 coverage of the highest annual Radiciodine and 3) Longview Road 14S1 SE 0.6 ground level D/Q. and a Particulates

4) King Road 13C1 SE 2.9 control location.

Radio-l (b)

5) 2301 Market Street.

13H4 SE 28.8 iodine cartridges which have Philadelphia, PA (control) been tested for performance by the manufacturer are used at all times.

3. Waterborne (c) 9 LOCATIONS (C) All surface and drinking stations have continuous samplers.

Surface

1) Limerick Intake (control) 2451 SSW 0.3

2) Linfield Bridge 1682 SSE 1.1 Ground

1) LGS Information Center 1151 ESE 0.5

2) South Sector Farm Hear Site 18A1 S

1.0 Drinking

1) Phoentaville Water Works 15F7 SSE 5.2

2) Pottstown Water Authority 28F3 WNW 5.9 (control)

3) Philadelphia Suburban Water 15F4 SE 7.8 Company

4) Citizens Home water Company 16C2 SSE 2.4 Sediment Form

1) Vincent Dam Pool Area 16C4 S

1.9 Shoreline ENCR. 36<'

ll. P.

(fhG(('

DATE R-/.7, M, i,_

v

S S'

Y b

4.

Ingestion 6 LOCATIONS Milk (d)

1) Control Station 22F1 (d) Milk samples are taken from 2) 17C2 several farms surrounding LGS.

3) 1081 These farms include those with the 4) 2581 highest dose potential from which samples are routinely available, as well as a control station.

The locations of the farms are not itsted herein due to a long-standing agreement with the farmers involved.

In return for being allowed to sample and analyze the milk. PECO has agreed not to divulge the location of the farmo.

l Fish (e)

1) Middle of Vincent Pool 16C5 SSE I.9 (e) Two species of recretionally im-Upstream to Pigeon Creek porant fish, sunftsh and brown bullhead will be sampled if available.

2) Upstream'of LGS. Keim Street 29C1 WNw 3.2 Bridge to Hanover Street Bridge (control)

Food Products

1) LGS Information Center 1151 ESE O.5 (f) Food products are to be samples (f) as part of the LGS Technical Specification Program only if milk sampilng is not performed.

The milk pathway, which results in a higher maximum dose to humans than the vegetation path-may is monitored at a location near the site, and is a better indicator than vegetation samples.

In addition, no crops grown in the vicinity of LGS are irrigated with water in which Itquid plant wastes have been discharged.

ENGR. Jga ki ll. P.

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2951 e

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1053 E

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23 S 2 24SI i1Si 24S2 ESE I

25 51 WSW I

14SI 21S1 16S2 I

SW Se SSE

{

si n eauno m SSW 5

1 36S1 EVERGREEN & SANATOGA RDS.

2151 IMPOUNDING BASIN 351 SANATOGA ROAD 23S2 TRANSMISSION TOWER 551 POSSUM HOLLOW ROAD 2451 LGS INTAKE 7SI LGS TRAINING CENTER 24S2 FRICKS LOCK I

10S3 KEEN ROAD 25S1 SW SECTOR 11S1 LGS INFORMATION CENTER 2653 MET. TOWER 92 1451 LONGVIEW ROAD 29S1 WNW SECTOR 1

16S2 LONGVIEW ROAD 32S1 NW SECTOR 18S1 RAILROAD TRACKS /LONGVIEW RD.

34S2 MET. TOWER N1 ENGR M _

HI] M.

DATE

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ENVIROfNENTAL SAMPLING LOCATIONS ON-SITE OR NEAR THE LIMERICK GENERATING STATION

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1081 FARM IN ESE SECTOR 17C2 FARM IN S SECTOR 15F4 PHIL. SUBURBAN WATER CO.

18A1 ANDERSON ROAD 15F7 PHOENIXVILLE WATER CO.

22F1 FARM IN SW SECTOR I

1682 LINFIELD BRIDGE 25B1 FARM IN WSW SECTOR 16C2 CITIZENS HOME WATER CO.

28F3 POTTSTOWN WATER AUTHORITY 16C4 VINCENT POOL 29C1 KEIM STREET BRIDGE 16C5 VINCENT POOL i

DATE h/3 M I

AQUATIC AND TERRESTRIAL ENGR. M ENVIR0tNENTAL SAMPLING HP M

STATIONS AT INTERMEDIATE l

DISTANCES FROM THE LIMERICK GENERATING STATION t

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2E1 LAUGHING WATERS 1901 SNOWDEN SUBSTATION 4E1 NEIFFER ROAD 2001 ELLIS WOODS ROAD 6C1 POTTSTOWN AIRPORT 20F1 SHEEDER SUBSTATION I

7El PHEASANT ROAD 24D1 PORTERS MILL SUBSTATION 9C1 REED ROAD 2501 HOFFECKER S KEIM ST.

10E1 ROYERSFORD ROAD 28D2 W. CEDARVILLE ROAD 10F3 TRAPPE SUBSTATION 29El HIGH SUBSTATION 13C1 KING ROAD 31D1 LINCOLN SUBSTATION 13E1 VAUGHN SUBSTATION 31D2 POPLAR SUBSTATION 15D1 SPRING CITY SUBSTATION 34E1 YARNELL ROAD I

16F1 PIKELAND SUBSTATION 35F1 RINGING ROCKS SUBSTATION 1781 LINFIELD SUBSTATION AIRBORNE AND TLD I

ENGR _4fe# ENVIRONMENTAL SAMPLING g}{

STATIONS AT INTERMEDIATE HP DATE 8'/3-8[p DISTANCES FROM THE LIMERICK l

GENERATING STATION

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1 SH1 BIRCH SUBSTATION 13H4 2301 MARKET ST PHIL.

ENG. M HP.

//yJr DATE

df-B(0 EtN!RONMENTAL SN4PLING STATIONS AT REMOTE DIS-

- 2 3-TANCES FROM THE LIMERICK GENERATIr0 STATION

o.g.

Ci release of flow path of interest Ci I

all release flow paths MPCi b.

DETERMINE C.R.

CR = Ct/E I

CR = the calculated monitor count rate above background attributable to the radionuclides; CPS E

= the detection efficiency of the monitor; uCi/cc/ cps.

I_

C.

The monitor high-high alarm setpoint above background should be set at the C.R. value.

The monitor high-high alarm setpoint will be calculated monthly.

The calculation will be I

based on isotopes detected in the liquid radwaste sample tanks during the previous month.

If there were no isotopes detected I

during the previous month then the annual average concentrations (EROL Table 3.5-3) of those isotopes listed in Table II.A.1 will be used to determine the setpoint.

If the I

calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to the new value.

If the I

calculated setpoint is greater than the existing monitor setpoint, the setpoint may remain at the lower value or increased to the new value.

A2.

Plant Service Water Monitor - Monitor alarm setpoint will be determined in order to be able I

to identify and rectify any potential problem due to excessive leakage of heat exchangers.

This setpoint results in concentrations at the site boundary far below 10CFR20, Appendix B, Table II I

limits.

The service water side of the fuel pool heat exchangers is kept at higher pressure than a

the shell side to prevent potential radioactive 5

c "t*=i"^ti a f '"* "*rvica "* tar-a.

DETERMINE CRS CRS = Z (CRB)

CRS = the calculated monitor setpoint I

count rate attributable to system leakage plus background; CPM I

Z

multiplier to establish monitor sstpoint count rate above background count rate C.R.B

monitor count rate attributable to background radiation; CPM b.

The monitor high alarm setpoint will be calculated monthly.

The calculation will be

'g based on the background count rate during

,g the previous month.

If the calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to I

the new value.

If the calculated setpoint is greater than the existing setpoint, the setpoint may remain at the lower value or i,ncreased at the new value.

A3.

RHR Service Water Monitor - Monitor alarm setpoints will be determined in order to be able I

to identify and rectify any potential problem due to excessive leakage of heat exchangers.

This setpoint results in concentrations at the site I

boundary far below 10CFR20, Appendix B, Table II limits.

The following method applies to liquid releases from the plant to the spray pond when determining the high-high alarm setpoint for the I

RHR Service Water Monitor during all operational conditions.

When the high-high alarm setpoint is reached or exceeded, the releases will be automatically terminated.

a.

DETERMINE CR CRS = (Z) CRB CRS = the calculated monitor count rate above background attributable to system leakage plus background; CPM Z

= multiplier to establish monitor setpoint count rate above background count rate.

CRB = monitor count rate attributable to background radiation; CPM E

= the detection efficiency of the monitor; uCi/cc/ CPM.

I

~

I I t

I b.

The monitor high-high alarm setpoint cbovo i

background should be set at the C.R.

value.

I The monitor high-high alarm setpoint will be j

calculated monthly.

The calculation will be based on the background count rate during I

the previous month.

If the calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to the new value.

If the calculated setpoint I

is greater than the existing monitor setpoint, the setpoint may remain at the lower value or increased to the new value.

B.

Gaseous Efflu.ents Bl.

North and South Stack Vent Radiation Monitors -

Monitor alarm setpoints will be determined in order to assure compliance with 10CFR20.

The setpoints will indicate if the dose rate at or i

beyond the site boundary due to radionuclides in the gaseous effluent released from the site is approaching 500 mrem /yr to the whole body and 3000 mrem /yr to the skin from noble gases, or I

1500 mrem /yr to the thyroid from I-131 and I-133 (inhalation pathway only).

The alarm setpoint for the gaseous ef fluent radiation monitors will be calculated as follows:

a.

North and South StacIr Vent Noble Gas Channel 1)

DETERMINE Ct CT = 2.12E-03 Ot I

CT =

the concentration at the vent noble gas l'

radiation monitor which indicates that the 10CFR20 dose rate limit at the site boundary has been reached; uCi/cc 2.12E-03 = unit conversion factor to convert uCi/sec/CFM to uCi/cc.

Qt =

the total release rate of all noble gas radio-nuclides in'the gaseous effluent (uci/sec) based g

on the lower of either the whole body exposure 5

limit (s00 mrem /yr) or the skin exposure (3 000 mrem /yr) Qt will be calculated as shown in Attachment 1.

anticipated maximum vent flow rate;

?

=

CFM I

I I

'~

~

I 2)

DETERMINE THE NOBLE GAS CHANNEL AI ARM SETPOINT (Sn)

Sn =

VFi (CT)

I VFi = fractional contribution to site boundary dose rate from the release point of interest; i.e. noble gas dose rate contribution from North Vent divided by the total I

noble gas dose rate contribution from the North and South Vents.

Normally the VFi values will be I

determined on a monthly basis but may be performed more often in response to plant conditions.

I b.

NORTH AND SOUTH STACK VENT IODINE CHANNEL 1)

DETERMINE CT CT =

2.12E-03 Ot E

CT = the concentration at the vent iodine radiation monitor which indicates that the 10CFR20 dose I

rate limit at the site boundary has been reached; uci/cc.

I 2.12E-03 =

unit conversion factor to convert uCi/sec/CFM to uC1/cc.

Ot = the total release rate of radiciodines in the I

gaseous effluents (uci/sec) Ot will be calculated as shown in Attachment 1.

F

= maximum antcipated vent flow; CFM.

2)

DETERMINE THE IODINE CHANNEL ALARM SETPOINT (Si)

Si = VFi (CT)

I VFi fractional contribution to site boundary

=

dose rate from the release point of interest; i.e.

iodine dose rate contribution from North Vent divided by 3

total iodine dose rate contribution from the North and South Vents.

Normally, the VFi values will be 8

determined on a monthly basis but may be performed more often in response to plant conditions.

I -

B2.

Tha monitor clarm s:tpoints will ba celculated monthly.

Th:sa routina calculctions will be based on isotopic analysis of the first scheduled I

sample of the month.

The monitor alarm setpoint calculations may be performed more often in response'to plant conditions.

If there were no I

isotopes detected in the sample, then isotopic concentrations calculated from the expected annual average noble gas and iodine-131 and 133 isotopic release rates (EROL Table 3.5-6) will be I

used to determine the setpoint.

If any calculated setpoint is less than the existing monitor setpoint, the setpoint will be reduced to I

the new value.

If the calculated setpoint is greater than the existing value, the setpoint may remain at the lower value or increased to the new value.

Due to the fact that I-131 and I-133 comprise 98.5% of the total dose based on expected ~ annual I

average releases (LGS FSAR Table 11.3-1) and particulates contribute a minor fraction of the total dose, a particulate channel setpoint will not be calculated for purposes of the ODCM.

B3.

CONTAINMENT PURGE ISOLATION I

a.

Monitor alarm setpoints will be determined for the North Stack Vent Wide Range Gas Monitor to initiate closure of the I

containment purge supply and exhaust lines in the event that high radioactivity releases are detected.

The setpoint will be determined to alarm and isolate containment i

at the minimum release rate from the North Vent which corresponds to a value less than or equal to 2.1 uCi/cc; per Technical i

Specification Table 3.3.2-2, Primary Containment Isolation.

The total effluent high alarm setpoint for the Wide Range Gas Monitor will be calculated as follows:

1)

DETERMINE Si Si =

Ci x F (min.) x 472 Si =

containment purge isolation setpoint (uci/sec)

Ci =

a value < 2.1 uCi/cc determined l

by the pTant staff.

l F(min) minimum anticipated vent flow

=

I rate during purge units conversion factor to convert 472

=

uCi/cc per CFM to uCi/sec I

ENGR. M g,p, jyg{f DATE R-L3-d

B4.

Containmtnt Purgn During Routine Conrations

,3 a.

Monitor alarm setpoints will be determined 5

for the North Stack Vent Wide Range Gas Monitor to indicate to Control Room personnel that unanticipated high I

radioactivity releases are detected.

The setpoint will be determined to alarm in the event that 10CFR20 dose rates at the site boundary are approached or exceeded.

The I

total effluent alert alarm setpoint for the Wide Range Gas Monitor will be calculated as follows:

I 1)

DETERMINE Sn S.n = VFiQt Sn =

Containment purge 10CFR20 alert alarm limit (uC1/sec)

Qt =

the total release rate of all noble gas radionuclides in the gaseous I

effluent (uCi/sec) based on the lower of either the whole body exposure limit (500 mrem /yr) or the skin exposure limit (3000 mrem /yr)

VFi = fractional contribution to site boundary dose rate from the release point of interest I

1.e. noble gas dose rate contribution from the north vent divided by total noble gas dose rate contribution from the North and i

South Vents Normally the VFi values will be determined on a monthly basis but may be performed more often in response to I

plant conditions.

b)

Prior to containment purge and venting, the I

monitor setpoint will be recalculated.

The calculations will be based on the noble gases detected by isotopic analysis of the containment atmosphere.

If the calculated I

setpoint is less than the exisiting monitor setpoing, the setpoint will be reduced to the new value.

If the calculated setpoint I

is greater than the existing value, the setpoint may remain at the lower value or increased to the new value.

5

~

I I 8 m

I B5.

Rot Maintenanca Shop Satpoint Datermination a.

The Hot Maintenance Shop Particulate and Iodine setpoints are based on a worst case isotope assumption.

Although the I

application of the worst case isotope results in a highly conservative setpoint, releases from the Hot Maintenance Shop are expected to be small by comparison.

In I

addition, a sufficient margin of safety factor is built in to the calculation, to preclude the application of a VFi for the I

releasa point.

The lodine and particulate high alarm setpoint is set to alarm in the vent that 10CFR20 dose rates at the site boundary are approached or exceeded.

l 1.

IODINE I

CT =

1500 mR/hr I

(1.0E-05 sec/m3) (7000CFM) (47 2) (1.62E07 mrem /yr) uCi/m3 I

CT = the concentration at the iodine monitor which I

indicates that the 10CFR dose rate limit at the site boundary has been reached, 2.8E-06 uCi/cc 1500mR/yr = 10CFR20 dose rate limit for iod ine,

I tritium and particulates with half lives greater than 8 days.

1.0E-05 sec/m3 = annual average depleted Chi /O 7000 CFM = maximum vent flow rate 472 = conversion factor to convert uCi/sec per CFM to uCi/cc I

1.62E07 mrem /yr = inhalation dose factor, I - 131 for uCi/m3 Ch ild, pe r Reg. Gu id e 1.109 I

I I

ENGR. Jr2 I

H.P.

N DATE M 434 -

2.

DStcrminn the Hot Ma,intenancs Shop high alarm sntpoint for iodino as follows:

.I Si = 0.01(2.8E-06 uCi/cc)

Si =

Hot Maintenance Shop Iodine high alarm setpoint; 2.8E-08 uCi/cc

.01 =

Margin of Safety Factor to encompass possible contribution from all other release points.

l 3.

PARTICULATE 3

Sp =.01(2.8E-06 uCi/cc) g 6.1 Sp =

Hot Maintenance Shop Particulate high alarm setpoint; 4.59E-09 uCi/cc

.01 =

Margin of Safety Factor to encompass possible contribution from all other release points 2.8E-06 uCi/cc =

The concentration at the iodine monitor which indicates that 10CFR20 I

dose rate limits at the site boundary have been reached.

I 6.1 =

ratio of the adult inhalation dose factor for Sr-90 x breathing rate for adult to the child inhalation dose factor for I-131 x breathing rate for child.

This ratio may be modified I

by plant personnel if the isotopes available for release are identified and a new ratio based on dose weighted averages is established I

I I

E ENGR. JJAC H.P.

ff!K DATE 8/3 4

ATTACHMENT 1 Ot Calculations I

1.

Ot(whole body) 500

=

(X/Q)v KiSi Qt

= the total release rate of all noble gas I

radionuclides in the gaseous effluent; uCi/sec.

(X/Q)v

= 1.lE-05 sec/m3; the highest calculated I

annual average relative concentration for an area at or beyond the site boundary for all vent releases (NE boundary).

Ki

= whole ' body gamma dose factors due to noble gases listed on Table III.A.1 (from Reg.

Guide 1.109, Table B-1).

Si

= the fraction of the total radioactivity in the gaseous effluent comprised by noble gas radionuclide "1".

2.

Ot(skin) 3000

=

(X/Q)v Zi((Li + 1.1Mi)Si)

(X/0)v

= 1.lE-05 sec/m3; the highest calculated annual average relative concentration for an I

area at or beyond the site boundary for all vent releases (NE boundary).

I Li

= beta skin dose factor due to noble gases, listed on Table III.A.1 (from Reg. Guide 1.109, Table B-1).

Mi

= air dose factor due to noble gases, listed on Table III.A.1 (from Reg.

Guide 1.109, Table B-1).

Si

= the fraction of the total radioactivity in the gaseous. effluent comprised by noble gas radionuclide "i".

I I

I

~

I I lI

3.

Ot(thyroid)=

1500 (X/Q)d (PiAi I

Ot

= the total release rate of radiciodines in the gaseous effluent; uCi/sec.

(X/Q)d

= 1.0E-05 sec/m3; the highest calculated I

annual average depleted concentration for an area at or beyond the site boundary for all Vent releases (NE boundary).

I Pi

= inhalation dose factor for child thyroid for radioiodines mrem-m3/uCi-yr; 1.62E07 for I-131 and 3.85E06 for I-133 Ai

= the fraction of the total radioactivity in the gaseous effluent (iodine channel) comprised by I

radionuclide "i".

I I

I I I

k VII.

BASES f

L Site Specific Data Note l':

Liquid dose factors, AiO, for section III.A were developed using the following site specific data.

The liquid pathways involved are drinking wcter and fish.

The maximum exposed individual is an F

adult.

L l

AiO =

(Uw/Dw + UF x BFi) KO x DFi Uw

= 730 liters per year; maximum adult usage of drinking water (Reg. Guide 1.109, Table 3-5).

{

Dw

= 85; average annual dilution at Phoenixville Water Authority intake.

('

UF

= 21 kg per year; maximum adult usage of fish (Reg.

L Guide 1.109, Table E-5).

BFi = bioaccumulation factor for nuclide, 1, in fresh-water fish.

Reg. Guide 1.109, Table A-1, except P-32 which uses a value of 3.0E03 pCi/kg per pCi/ liter.

KO

= 1.14E05 (lE06pci/uC1)(lE03 ml/Kg)8760 hr/yr units conversion factor.

(

DFi = dose conversion factor for nuclide, i, for adults in total body or bone, as applicable.

Reg. Guide 1.109, Table E-11, e'xcept P-32 bone which uses a

{

value of 3.0E-05 mrem /pci ingested.

The data for D was taken from data published in Limerick Generating Station Units 1 and 2 Environmental Report Operating License Stage, Volume 3.

All other data except P-32 BF and DFi were used as given in Reg. Guide 1.109, Revision 1, October 1977.

A P-32 BFi value was taken from

[

Kahn, B. and K. S. Turgeon, "The Bioaccumulation Factor for Phosphorus-32 in Edible Fish Tissue", NUREG-CR-1336, March, 1980.

A P-32 DF value was taken from Limits for Intakes of Radionuclides by Workers, International Commission on

[

Radiological protection ICRP Publication 30, Supplement to Part 1, 1979.

Note 2:

To develop constant P(I-131) for Section III.A, the following data were used:

P(I-131)

K' (BR) (DFA)

=

10E06 pCi/uCi; unit conversion factor K'

=

3700 m3/yr; child's inhalation rate.

BR

=

4.39E-03 mrem /pCi; the thyroid inhalation DFA

=

I-131 dose factor for I-131 in the child.

ENGR. M H.P.

/7A'L(~ DATE W-/ 3'M r

Tha pathway is the inhalation pathway for a child.

All values are trkcn from Rtgulatory Guide 1.109, Revision 1, October 1977.

Note 3:

To develop constant R for section III.C, the following site specific data were used:

I RGi (D/Q)

K'QF (Uap) (Fm)(r) (DFLi)a fp(1-fs)(exp-itf)

=

,\\iew Yp K'

= 1E06pCi/uCi unit conversion factor QF

= 6Kg/ day; goat's consumption rate Uap

= 330 1/yr; yearly milk consumption by an infant hi (9.97E-07)/sec decay constant for I-131;

=

9.48,E-06 for I-133.

hw (5.73E-07)/sec decay constant for removal

=

of activity in leaf and plant surfaces.

I Fm

= (6.0E-02) day / liter, the stable element transfer coefficient for I-131.

r

= 1.0 fraction of deposited radioiodine retained in goat's feed grass.

DFLi = (1.39E-02) mrem /pCi - the thyroid ingestion dose factor for I-131 in the infant; 3.31E-03 mrem /

pCi for I-133.

fp

= 0.75; the fract' ion of the year the goat is on pasture (average of all farms).

fs

= 0.0; the fraction of goat feed that is stored feed while the goat is on pasture (average of all farms).

Yp

= 0.7 Kg/m2 - the agricultural productivity of pasture feed grass.

t

= 2 days - the transport time from pasture to goat, f

to milk, to receptor.

The pathway is the grass-goat milk ingestion pathway.

These data were derived from data published in Limerick Generating Station Units 1 and 2 Environmental Report I

Operating Stage, Volume 3.

All other data were used as given in Reg.-Guide 1.109, Revision 1, October 1977.

Similar data were used to develop the constant R for I-133.

. I

E Nota 4:

Th2 m thodology d;scrib:d herein will b0 implem nted B

vic computcr cod:s.

Thesa cod:s hava been vsrified as documented in:

1.

G.A. Technologies, RM-21A Computational Models, Document No. E-115-1241, June 1984.

2.

G. A. Technologies, Meteorological Monitoring, Display and Reporting System /RM-21A, Document No. 0375-9032, January, 1984.

Surveillance Requirement 4.11.1.2 Licuid Pathway Dose Calculations The equations for calculating the doses due to the actual release rates of radioactive materials in liquid effluents were developed from the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual, Doses to Man from Routine Releases of I

Reactor Effluents for the Purpose of Evaluating Compliance with 10CFRPart 50, Appendix I",

Revision 1, October 1977 and NUREG-0133 " Preparation of Radiological Effluent Technical I

Specifications for Nuclear Power Plants". October 1978.

Surveillance Requirement 4.11.2.1.1 and 4.11.2.1.2 - Dose Noble Gases The equations for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were I

developed from the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance I

with 10 CFR Part 50, Appendix I",

Revision 1, October 1977, NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", October 1978, and Regulatory Guide 1.111, " Methods for Estimating Atmospheric I

Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977 with site specific dispersion curves and disperion methodology.

The I

specified equations provide for determining the air doses in areas at and beyond the SITE BOUNDARY based upon the historical average atmospheric conditions.

The dose due to a noble gas release as calculated by the Gross Release Method is much more conservative than the dose calculated by the Isotopic Analysis Method.

Assuming the release rates I

given in Limerick Generating Station Units 1 and 2 Environmental Report Operating License Stage, Volume 3, the values calculated by the Gross Release Method for total body dose rate and skin I

dose rate are 4.8 times and 3.25 times, respectively, the values calculated by the Isotopic Analysis Method.

I

. I

For tha Gro;c Ralcase Mathod, Kr-87 and Kr-88 are uccd for tha liciting ckin end total body done fcctore resp;ctivaly, due to half life considerations.

Kr-89, the nuclide with the highest I

dose factors per Regulatory Guide 1.109 Table B-1 has a half-life of 3.2 minutes while the half-lives of Kr-87 and Kr-88 are 76 minutes and 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> respectively.

Therefore, by the time that gaseous effluents have been transported offsite, Kr-89 will have decayed enough so'that Kr-87 and Kr-88 are effectively the most limiting nuclides.

The model Technical Specification LCO, (Limiting Condition for Operation) for all radionuclides and radioactive materials in particulate form and radionuclides other than noble gases I

requires that the instantaneous dose rate be less than the equivalent of 1500 mrem per year.

For the purpose of calculating this instantaneous dose rate, thyroid dose from iodine-131 and iodine-133 through the inhalation pathway will be used.

Since I

the expected annual releases presented in LGS FSAR Table 11.3-1 indicate that iodine-131 and iodine-133 releases have the major dose impact this approach is appropriate.

The value calculated I

is multiplied by 1.02 to account for the thyroid dose from all other nuclides.

This allows for expedited analysis and calculation of compliance with the LCO.

Surveillance Requirement 4.11.2.2 and 4.11.2.3 - Dose Noble Gases The equations for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents were developed from the methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance I

with 10 CFR Part.50, Appendix I",

Revision 1, October 1977; NUREG-0133 " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants", October 1978; and I

Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977 with site specific dispersion curves and dispersion methodology.

The I

specified equations provide for determining the air doses in areas at and beyond the SITE BOUNDARY based upon the historical average atmospheric conditions.

MEMBERS OF THE PUBLIC, who may at times be within the SITE BOUNDARY, will be subject to lower annual average concentrations I

than those calculated at the SITE BOUNDART.

The maximum expected occupancy factor is a working year (or 25% of the year) along the railroad t:acks.

The maximum chi /q along the railroad tracks is 2.09 E-06 in the West sector.

Both chi /q and the occupancy I

factor are lower for this case than the NE sector SITE BOUNDARY.

The dose due to noble gas releases as calculated by the Gross I

Release Method is much more conservative than the dose calculated by the Isotopic Analysis Method.

Assuming the release rates given in Limerick Generating Station Units 2 and 3 Environmental Report Operating License Stage, Volume 3, the values calculated by the Gross Release Method for total body dose rate and skin dose rate are 4.8 times and 3.7 times, respectively, the values calculated by the Isotopic Analysis Method. DATE fd-/ h do I

l l Do'n, Iodinn-131, Tritium, cnd Rrdiorctiva Matorial in lW Pirticulnto Form (3

The equations for calculating the doses due to the actual release g

rates of radiciodines, radioactive material in particulate form, and radionuclides other than noble gases with half-lives greater than 8 days were developed using the methodology provided in l

Regulatory Guide 1.109, " Calculation of Annual Doses to Man from l

Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I",

Revision lg 1, October 1977; NUREG-0133, " Preparation of Radiological 5

Effluent Technical Specifications for Nuclear Power Plants",

October 1978; and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977 with site specific dispersion curves and dispersion methodology.

These equations provide for determining the actual doses based upon the h,istorical average atmospheric conditions.

MEMBERS OF THE PUBLIC, who may at times be witin the SITE BOUNDARY, will be subject to lower annual average concentrations

,I than those calculated at the SITE BOUNDARY.

The maximum expected occupancy factor is a working year (or 25% of the calendar year) along the railroad tracks.

The maximum depleted chi /q along the I

railroad tracks is 1.97 E-06 in the West sector.

Both depleted chi /q and the occupancy factor are lower for this case than the NE sector SITS BOUNDARY.

Compliance with the 10 CFR 50 limits for radiciodines, radioactivo materials in particulate form and radionuclides other than noble gases with half lives greater than eight days is to be I

determined by calculating the thyroid dose from iodine-131 and iodine-133 releases.

Since the iodine-131 and iodine-133 dose accounts for 99.97 percent of the total dose to the thyroid, the value calculated is not increased.

I I

~ I

m m

U O

O O

M O

O T

DATE L:NGH fjac h(O IIP L.D. Tank FE RL Laundry L.D*

Drain (LD)_

i1

Sample Substation Tank g

To Cooling

+

L.D. Tank wn Line 10 GPM (Nom.)

Sample point)

,hFromChemWasteTank FD FD FD

+ Collection Sample Demin Fi er Floor Tank (FDCT)

Tank #2 280 GPM Drain (FD)_

,To. Waste To FD (Nom.)

Subsystem

% Sludge Tank

Spent Resin Tank To Condensate FD e s e

Storage Tank

'\\ e ED

-* Sam e s,

-* Collection Tank Equipment Tank

ED ED Drain (ED)

Filter Demin s,

Subsystem' To Condensate go To Waste Samp e

E

+ Surge

.-+ Sludge Tank To ED Tank Tank Spent Resin Tank To FDCT Chem Waste (CW)

CW Subsystem Tank i

NOTE: Cooling' Tower Blowdown Line provides dilution Philadelphia Electric Company imeric enerating Station flow of 10,000 gpm for both units.

Units I&2 Liquid Effluent Flow Diagram Figure IX.A.l Rev. l EE E

m M

M M

O UNIT I UNIT 2 9

O L

L Particulate South Vent South Vent Particulate Iodine and Unit 1 Unit 2 Iodine and Noble Gas El. 411' El. All' Noble Gas Monitor (2)

Continuous Continuous Monitor (2) i (2)

M (2) s

~

y 234000 234000 cfm

'cfm Refueling Floor

~

Refueling Floor max max Exhaust M

t f

Exhatis t U (3)

(3)

Reactor Enclosure Reactor Enclosure Exhaust M

O s

Exhaust Q3)

T4000 54000 (3) D cfm cfm Equipment "I"

I"I" Equipment Compartment g

)

(

Compartment Exhaust (2)

(2)

Exhaust Philadelphia Electric Company Limerick Generating Station ENGR,Sf4d b

IIP Gaseous Effluent DATEg,f3, M

iagram i

41 South Stack

{

Figure IX.A.2 Rev. I

p i

i rm rm ro e

ra o

rm ro ra rm rm ro ro rT r

rm o

ENGR,-w

_ 42_

llP EM, DATE A

7 North Vent 1

Units I&2 Particulate & Iodine Elev. 4Il' filters. Low range gas monitor Wide I

A Range i

Particulate, Iodine, &

Continuous Acc{ dent

& {g range Monitor l

Noble Gas Normal Range Release gg gas monitor u

O O (3)

Turbine Eqclosure Exhaust 664000 cfm-2

.f Turbine Enclosure Exhaust (3) C V

Turbine Enclosure Equip.

X/

v Turbine Enclosure 4

g\\g(2) Comoar.tment Exhaust (2)

U Equip. Compartment Exhaust (2)

(2) g g.

(2)

Mechanical Vac. Pump 183000'cfm in W

, Battery Room Exhaust m

Steam Packing Exhauster 4

Steampacking exhauster D/W e

Standby Gas TreatmentSy.

(2) 9 5 adwaste Encl. Exhaust R

Radwaste Encl. Equip.

SGTS C (2) D

( Compartment Exhaust Fume Hood Exhaust (2) 4 l

I jUnit 1 Recombiner Control Area Exhaust m,

n (2) b (2) F N i

', Unit 2 Recombiner I **

I Philadelphia Electric Company Limerick Generating Station Units 1&2 tbhh Filter Gaseous Effluents Exhaust Fan

Unit 1 Charcoal System Flow Diagram - North Stack

- Unit 2 Charcoal System p ;,,,,., ty_a.3 p,,

i

-p l-sw-

'.S**

6

_,g y

I

p. '..

(.

n. '

.gygwys..-

,~~

6..;WW i*). ' -

.pr%g

-L...

?

a i

1 9

p R

,j s

w l

V o

~

i i,

\\. T v

. \\. /

i y

t

.s

? niitatu

' t~.

y '

j k,-(-

. 6.7 a.,

ua p".

o (neuma g

9 l

4.v c-v s

5 r

_ i 4(

}>l %

'N

~

T

~

q 3

y e.sg j s

s.,

o e ii

..,(.

i \\

A

-7(,

A.< r

'(o

.\\

g

' i t

p.

's 4

i

. r,. \\

~

.c fs,g @y.,s-c.-N.

A A+,

i u

-]

1 J

j 1

84AP DEFINING UNRESTRICTED AREAS AND SITE BOUNDARY FOR RADIDACTIVE GASEOUS AND

(

LSDUID EFFLUENTS 3-ricuas.IX.A:4 q

t t

+

=..

4 4

4 III.

Attachments C.

"B" Residual Heat Removal (RHR) Service Water Radiation Monitor Failure 4

4 4

4

'4 1

4

III.

Attachments C.

"B" RHR Service Water Radiation Monitor Failure

{

As stated in Surveillance Requirement 4.3.7.11, a Channel Function test for the RHR Service Water Radiation Monitors will demonstrate that an automatic isolation and control room annunciation will occur if c

L any of the following conditions exists:

1.

instrument indicates measured levels above the

{

alarm / trip setpoint.

2.

circuit failure.

[

3.

instrument indicates a downscale failure.

During a planned review of Surveillance Tests, it was

[

discovered that the downscale failure logic would not L

isolate the system and the alarm / trip setpoint and circuit failure logic would not annunciate in the control room due to the existing design.

These

[

Radiation Monitors were declared inoperable and the appropriate action was taken to comply with Technical Specifications.

A modification was completed and F

tested to include a downscale isolation and a circuit L

failure and alarm / trip annunciation.

The monitors were then declared operable.

The modification was completed in 30 days; however, the 30 day action was exceeded

[

since the original radiation monitors were not installed with the above capability until the modification was complete.

E E

[

E,_

PHILADELPHIA ELECTRIC COMPANY 23O1 MARKET STREET P.O. BOX 8699 P:llLADELPHIA. PA.19101 12151841-4000 August 29, 1986 Docket No. 50-352 Dr. Thomas E. Murley, Administrator U.S. Nuclear Regulatory Commission Region I 631 Park Avenue King of Prussia, PA 19406

SUBJECT:

Semi-Annual Effluent Release Report No. 4 January 1, 1986 through June 30, 1986 Limerick Generating Station Unit 1

Dear Dr. Murley:

Enclosed are two copies of the Semi-Annual Effluent Release Report No. 4 for the period January 1, 1986 through June 30, 1986 for Limerick Generating Station Unit 1.

This report is being submitted in compliance with the Technical Specification 6.9.1.8 of Operating License NPF-39 and Regulatory Guide 1.21 to fulfill the requirements of Regulatory Guide 10.1.

Very truly yours, W.

M. Alden Engineer-In-Charge Licensing Section Nuclear Generation Division PBB:vdw Enclosures Attachments l

cc: Document Control Desk, USNRC g[1k)Jf' l

E. M. Kelly, Senior Site Inspector See Attached Service List (w/o Enclosures) l

//

J

cc:

Troy B.

Conner, Jr., Esq.

Benjamin H. Vogler, Esq.

Mr. Frank R. Romano Mr. Robert L. Anthony Ms. Maureen Mulligan Charles W. Elliott, Esq.

Barry M. Hartman, Esq.

Mr. Thomas Gerusky Director, Penna. Emergency Management Agency Angus Love, Esq.

David Wersan, Esq.

Robert J.

Sugarman, Esq.

Kathryn S. Lewis, Esq.

Spence W.

Perry, Esq.

Jay M. Gutierrez, Esq.

Atomic Safety & Licensing Appeal Board Atomic Safety & Licensing Board Panel Docket & Service Section (3 Copies)

E. M.

Kelly Timothy R. S. Campbell July 23, 1986 l

l l

._

ML20214W853

Text

SUMMARY

SUMMARY

SUMMARY

SUMMARY

REFERENCE:

REFERENCE:

multiplier to establish monitor sstpoint count rate above background count rate C.R.B

SUBJECT:

Dear Dr. Murley:

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