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t k N lC ,f AMENDMENT 31 y.
'.. DOCKET No. 50 293
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a . - . :- [ , '} AUGMENTED SYSTEMS FOR ;. ,
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- h. REDUCTION OF RADIOACTIVE ....' ..
i . MATERIAL IN EFFLUENTS
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- d. i BOSTON M ' PILGRIM '
ECVRSOD. c~anNUCLEAR POWER STATION i ' COMPANY j,;. a' h , .i , . . . . , ,;; . . i.. . .. . i I i'
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se o AMENDMENT 31 f'9 i ! AUGMENTED SYSTEMS FOR PEDUCTION OF RADI0 ACTIVE MATERIAL IN EFFLUENTS
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TABLE OF CONTENTS i I y Page Number j I. Introduction 3 e .
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3 II. Summary 5
) III. BWR Radioactive Material Sources 7 (a) Design Basis Source Term Development 7 (b) Design Basis Source Terms 8
- 1. Activation Products 8 1.1 Coolant Activation Products 8 1.2 Non-Coolant Activation Products 8
- 2. Fission Products 9 2.1 Noble Gas Fission Products 9 je 2.2 Halogen Fission Products 9 2.3 Other Fission Products 9 .
(c) Design Basis Source Term Calculational Models and Uncertainties 17 - IV. ' Identification of Radioactive Effluent Pathways '. to the Environment 19
-1 (a) Gaseous Effluents 19
- 1. Main Condenser Cas Removal System ,
19
- 2. Turbine Gland Seal Holdup System 19
- 3. Mechanical Vacuum Pump 20
- 4. Station Ventilation Exhausts 20
- 5. Primary Containment Atmospheric Control System 20 (b) Liquid Effluents 21
- 1. Clean Radwaste System 21
- 2. Chemical Radwaste System 21
- 3. Miscellaneous Wastes 21
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h .. h TABLE OF CONTENTS O Page Number V. Control of Gaseous Ef fluents 22 (a) Augmented Offgas System 22
- 1. Design Objective 22
- 2. Design Basis 22
-- 3. System Description 22
- 4. Safety Evaluation 30
- 5. System Operation 38 ,
- 6. Inspection and Testing . 40 ,
f t (b) Miscellaneous Gaseous Effluents 40 - I 1. Turbine Gland Seal Holdup System 40 .
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- 2. System Function 40 -
- 3. System Description 41 _.
- 4. Equipment Description 41 ..'.
- 5. System Operation 41 ~" ""
- 6. Instrumentation 42 -
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Safety Evaluation Inspection and Testing 42 42
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_ _: ,- . . an Design Objective
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- 2. Design Basis *
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- 3. Systam Description 45 . ~;.i '.
- 4. Safety Evaluation 53 . 3JJ .
- 5. System Operation 53 .Di j .M
j (a) Normal Operation 53 "
. (b) . Malfunction and Failure Mode Analysis 58 i 6. Inspection and Testing -
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- (b) Miscellaneous Wastes 67
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h 1. Design Basis 67
- 2. System Function 67
- 3. System Description 67 .,
. . 4. Equipment Description .. 67 .
- 5. System Operation 68 1
- 6. " Instrumentation 68 ',
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- 7. Safety Evaluation 70
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[ 8. Inspection and Testing 70 4 O s V a
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,m.._---- .. _ _ m --- -s ... a .w . u .__.n.a._.._-=..,,.. . i l .. . TABIE OF CONTENTS 4. Page Number VII. Environmental Effects of Radioactive Effluents 71 i (a) Site Boundary and Large Population Group Exposures from Gaseous Effluents . 71 I 1. Exposures Resulting from Effluent from Augmented Offgas System 71
- 2. Exposures Resulting from Miscellaneous Gaseous Effluents 75 l
(b) Liquid Effluent Exposures to Individuals Through the Aquatic Food Chain 78 l c . I. VIII. Proposed Changes to Technical Specifications and Means for Determining Compliance 86
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3. M!ENDMENT 31 AUGMENTED SYSTDiS FOR REDUCTION OF RADI0 ACTIVE MATERIAL IN EFFLUENTS G I. INTRODUCTION s
- This document is submitted in response to
- (1) a letter from Peter A.
Morris to James M. Carroll dated June 23, 1971 and, (2) in response to recomendations on radioactive material in liquid and gaseous effluents made by the Advisory Comittee on Reactor Safeguards in a letter from , Spencer H. Bush to Glenn T. Seaborg dated April 7,1971. This Amendment i 31 supersedes all information in Amendment 29 relative to radioactive material in effluents. In particular, this Amendment addresses the specific cc=e.nts as noted belcw: (1) Letter Peter A. Morris to James M. Carroll dated June 23, 1971 P (a) Itses 1. , 2. , 3. ,1+. , and 5. are addressed in Section V, Control of Gaseous Effluents,of this Amendment.
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(b) " Operational limits and procedures proposed for utilization " of existing equiInent during (the) interim period" (pricr to installation of the Augmented Offgas System) are as given in the Technical Specifications attached to the proposed , facility operating license.
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.(c) "Further information regarding facility design features to process expected radioactive liquid wastes considering .
various ways for the disposition of evaporator tottoms ,.
. including a discussion of the characteristics of a . -
d sea-side site and how this may relate to the design of f] the radwaste system" is provided in Section VI, Control , of Liquid Effluents of this Amendment. -1 S i (2) Response to recomendations on radioactive mater'ial b liquid and. gaseous effluents made by the Advisory Comittee on Reactor Safeguards in a letter from Spencer H. Bush to Glenn T. Seaborg t dated April 7,1971.
"The applicant proposes that the gaseous and particulate '
i radioactivity discharged through the stack will not , exceed 10 CFR 20 limits. The Comittee believes the applicant should set a much lower operating limit and should make such e } accomplish this." quipment changes as may be necessary to P 4 >b v . 1, l
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u t 4. "\ Refer to Section V, Control of Gaseous Effluents of this Amend.::ent. i . l "The applicant has not provided equipment for concen'c rating and separating radioactivity from liquid wastes, and he i states that the radioactivity concentration in the condenser circulating water discharge will not exceed that permitted by 10 CFR 20. During the first reactor shutdown for refueling, the applicant will install an evaporator designed to permit the holdup of liquid wastes and thereby reduce the gross radioactivity discharged. The Co=mittee believes that the design and operation of this evaporator system-should be such as to reduce to levels as low as practicable i- the amount af long-lived radioisotopes discharged. The l Regulatory Staff should review and approve the design and r - operating mode of this equipment. The Committee also
- believes that prior to the installation of thic equipment, l
effort should be made to reduce the radioactivity released." Refer to Section VI, Control of Liquid Effluents of this ' " Amendment. - - ' *~~
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({ lV t II. _SW MARY The applicant proposes to procure, install and operate an Augmented Offgas System and an Augmented Chemical Radwaste System. The Augmented y OffgasSystemconsistsofcatalytichydrogen/oxygenrecombinersandcharcoal l adsorber delay beds with their associated equipment. The Augmented Chemical Radwaste System consists of an Ultrasonic Resin Cleaner, a liquid
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Radwaste Concentrator and a Radvaste Solidification and Packaging System ] with their associated equipment. ' System descriptions (Sections V(a) 3. and VI(a) 3.) and safety evaluations (Sections V(a) 4. and VI(a) 4.) for the proposed modifications i are provided in this report. Schedules for completion of these modifications are given in Table II-2. This report describes the most recent source term estimates (more recent than those used in Amendment 16 of the FSAR) and the pathways to q the environment of radioactive material in effluents, proposes control systems for these pathways and finalAy, evaluates the environmental effects of the reduced radioactive effluent releases. These effluent release j estimates supersede those presented in Amendment 16 of the FSAR. Proposed i Technical Specifications for Pilgrim Nuclear Power Station, and methods of determining compliance with these Technicd. Specifications, to go into effect after the new equipment becomes operational' arc also included. p O Table II-l summarizes site boundary exposures and e :posures to individuals + through the aquatic food chain resulting from the operation of the Pilgrim Nuclear Power Station with augmented systems. f , The new equipent described is not cupplied in order to meet any I 'known demonstrated technical or biomedical needs but is provided in . accordance with Part 50 34a of the Regulations of the United States Atomic f Energy Co= mission which require licensees to reduce radioactive effluent 3l releases to unrestricted areas to " levels which are as low as practicable"; a phrase that we interpret to mean the installation of radioactive ', effluent control equipent that is within the state-of-the-art and available l for purchase and installation at this time. - In Amendment 29 to the Final Safety Analysis Report (FSAR) the appli-r cant proposed an offgas activity reduction system which utilized compressed 8as holdup. This system would have reduced exposures to the 4 million people within 50 miles of the station from 309 manrem/ year for the 30-minute holdup system presently installed (assuming an average annual release rate equivalent to 0.025 curies /second after 30-minute holdup) to 10.9 manrem/ year at an estimated cost of $2,300,000. The recombiner charcoal absorbtion l system proposed herein would further reduce this exposure to,0.48 manrem/ l
*, . year at an estimated installed cost of $8,500,000. The estimated installed cost of the augmented chemical radwaste system ,is $2,000,000.
The design basis for the augmented offgas system and the augmented j 1 r~ ' chemical radwaste system is an offgas release rate of 100,000 microcuries/ k -
, second af ter 30-minute holdup. The Technical Specifications attached to 1 .
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l the proposed operating license limit the offgas release rate (as measured af ter the nominal 30-minute holdup time) to 100,000 microcuries/ second averaged over the preceding three calendar months. This restriction has the effect of limiting the annual average offgas release rate to levels on the order of 50,000 microcuries/second. However, it is anticipated that the long-term average release rate over a' period of several years will be on the order of 25,000 microcuries/second. Tables presented in this report have been developed on the basis of 100,000 microcuries/second and 25,000 microcuries/second. The basis for each table is noted at the beginning of l each table. Tables for which offgas release rate bases are not noted are not direct functions of such release rates. . 3 N95 4., y
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/N TABIE II-l
SUMMARY
OF RADIATION EXPOSURES RESULTING FROM OPERATION OF THE PILGRIM NUCIEAR POWER STATION WITH AUGMENTED SYSTEMS Millirem / Year 1 at Site Gaseous Effluents Boundary (l) Effluent From Augmented Offgas System 0.023 Miscellaneous Gaseous Effluents - Effluent From Gland Seal Holdup System 0.15 All Other Miscellaneous Gaseous Effluents 0.50 -- Total Annual Gaseous Effluent Exposures 0.67 f Site Boundary Dose due to inhalation 0.6 (Thyroid dose, of I-131 released from Station ventila- < limit is 1500 tion exhausts. . millirem / year) ,
'th Exposures to Individuals through the 0.67 (Whole body dose,
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Aquatic Food Chain - limit is 500 .; (Assumes all station liquid effluents millirem / year) are released to the discharge canal - - -- r after processing through the Augmented . . . , 5.3 (To gastro- ..l. m;g f Chemical Radwaste System and the intestinal M Miscellaneous Radwaste System) ~ tract, limit is 1500 =
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- (1) Assumes an average annual source equivalent ec 0.025 curies /second ,
af ter 30-minute holdup. ~
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6-A. l TABLE II-2 TARGET SCHEDUE FOR INSTALLATION OF AUGMENTED SYSTEMS
,(Assumes Timely Approval of Systems as Proposed) i Cffgas System 1
! . 1. Begin modifications to in-plant piping and l electrical systems and structure modifications i to turbine building. November, 1971 l 2. Begin construction of auxiliary building. March, 1972 l I 3. Shipment of major hardware complete l (orcered September, 1971). March, 1973
- 4. Complete construction. September 1973 5 Complete startup tests. December, 1973
- 6. System Operational January, 1974 Liquid Radwaste System ^
- 1. Begin modifications to in-plant piping, a x- N p electrical systems and structures. November, 1971
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- 2. -Complete installation of radwaste '
1; I ~ concentrator (ordered March 1971). February, 1972
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3.. Complete installation of ultrasonic resin cleaner (ordered September 1971). d March, 1972 I 4. Complete installation of radwaate -- - solidification system (ordered November 1971) August, 1972 } s ;. i x l P s 4
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5 7 III. Boiling Water Reactor Radioactive Material Sources (a) Design Basis Source Term Development Design basis radioactivity levels in the reactor water, steam and offgas were estimated in the early stages of station design. These estimates provided guidance in plant equipment design, shielding design,
' system operation and performance evaluation, radioactivity measurement i
device specification and in estimating expected activity releases to the environment. The design basis radioactivity levels include all isotopes observed or predicted to be present which are significant in these design areas. . These design basis radioactivity levels have now been revised on the basis of operating data and measurements completed at several operating boiling water reactors during 1970. Emphasis was placed en observations ;
- at the Dresden II Nuclear Power Station of the Commonwealth Edison Company and the Kernkraftwerk RWE-Bayernwerk GmbH (KRB) Nuclear . Power Plant at Gundremmingen, German Federal Republic. .
.- Of primary interest in estimating expected activity releases to the environment is the amou.t of radiogas which may diffuse through the UO2 ~ .
fuel itself and a part of which then might leak through the zircaloy fuel 2 " " ~ : cladding barrier. The objective' of fuel fabrication is to assure a minimum
- - of penetrations initially present in the cladding barrier, and to provide _,
. - _~ a' barrier which will not deteriorate significantly during the several . . _ .
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years of fuel usage in the reactor. However, as no mechanical barrier 3
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;Gbarrier must be considered in the design and operation of the plant and , j
- :7 f its effluent control systems. The selection of a reasonable leakage t#5.7 J'.' ' level is essential as a " design basis". Experience in operating BWR's
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_ reactor water than other classification groups of fission products, so 3 that " design basis" leakage rates can conveniently be described in terms 'g of the noble radiogas group. This design basis source term is not to be ,
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confused with an emission rate in effluents, since its prime purpose is .J l -
- - ..to establish the input term to be considered in effluent treatment system i
- design for gaseous effluents and also to provide guidance for liquid
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; effluent treatment process design. .4
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- e For the Pilgrim Nuclear Power Station, a design basis noble radiogas release rate of 100,000 microcuries/second (as measured after a nominal >
~ i30 minute holdup time) was celected. For the purpose of estimating the ;
E expected release rate of radionuclides to the offsite environment a 4 long-term average value of 25,000 microcuries/second is used. _ 9 C ;' .1 :It is noted that the present Technical Specifications limit the offgas- j _ . .' , release rate (as measured after the nominal 30 minute holdup time) to
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{fA - ~ ' ' ~ 'd]R100,000 microcuries/secondaveragedovertheprecedingthreecalendar .N' ? .P. . months. This restriction has the effect of limiting the annual average 1j
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a offgas release rate to levels on the order of 50,000 microcuries/second. 7'? l
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[*. 4 8. [m V) (b) Design Basis Source Ter=s . I , Estimating offsite radiation exposures resulting from operation of ]' the Pilgrim Nuclear Power Station first requires an estimate of the equilibrium pri=ary coolant activity concentrations on an isotope by isotope basis. Two major sources of coolant activity exist. One of these sources is the activation product group which results
- from neutron activation of:
(a) structural materials in'the core, , f (b) primary system corrosion products carried thrcugh the core ( j . neutron flux, and l (c) activation products restilting fr'om neutrcn activation of the primary cooht (vater) itself and dissolved gases in the water. The other major source of coolant activity is the fission product group. q - This source results from leakage of these fission products frcm the fue' elements. It consists of ~ J (a) noble gas fission products, z g '
'(b) haloge fission products, and (c) other fissicn products. -- a
- n The remainder of this section discusses methods for arriving at equilibrium ]!'
, primary coolant activity concentration levels for each of these activity ,
sources and lists these estimates in detail. -
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- 1. Activation Products l
1.1 Coolant Activation Products N These activation products ara present in both the steam and the reactor water. The coolant. activation product activity in the steam is , presented in Table III-1. The activity in the reactor water is presented in Table III-2. These activity concentrations result from neutron _
, activation of the primary coolant (water) itself. These estimates are '
expected values based upon et.lculational models rather than experimental l' observations. 2i .
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1.2 Non-coolant Activation Products '
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The design basis activity levels of ncn-coolant activation products j. , ' represent observed values in operating boiling water reactors. The non-coolant activation products are formed by activation of impurities y] 3 in the coolant or by corrosion of irradicted system materials. The activity '
/ T in the reactor water is presented in Table III-3 Carryover of these 3: '
isotopes into the steam is estimated to be less than or equal to 0.1% f(") 3L by weight of reactor water (60.001).
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! 2.1 Noble Gas Fission Products L 3 The design basis for noble gas fission product release rates is ; i 100,000 microcuries/second (after 30 minute holdup time). Emissions on l [ the order of 100,000 to 200,000 microcuries/second(at30minutesholdup ' time) can be tolerated for reasonable periods' of time. Operation in this i range would not be permitted to result in average offgas release rates in excess of the Technical Specification limits. For the purpose of estimating the expected site boundary and large population group exposure rate, a long-term average release rate of 25,000 a.icrocuries/second is used in Section VII(a) of this report. The isotopic composition of the design basis noble gas release rate ! is calculated from equation (1) in Part III(c The resultant isotopic } release rates are presented in Table III-4, as). released from fuel (t = 0) and after 30 minutes decay. While Kr-85 can be calculated using equation (1), ~the number of confirming e: perimental observations were limited by
'the difficulty of measuring very low release rates of this isotope.
Therefore, the table provides, as an alternate, an estimated range for Kr-85 based on actual measurements. ,
- q,-2.2 Halogen Fission Products
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.,, set at 700 microcuries/second I-131 from the fuel (t = 0). The halogen
- release rate is not directly related to noble gas release rate. The d-
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[Y'.. if (rates may be tolerated for reasonable periods of time, in-plant contamination
.~ and other operating restrictions suggest that long term operation above
,h this level would be undesirable. . . .c a
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.' ^ The halogen release rate from the fuel can be calculated from equatios "
.(2) in Part III(c). Concentrations in reactor water can be calculated from equation (3) in Part III(c). Observations at operating BWR's indicate
- . that the " carryover" of the radio-iodines is estimated to be less than -
g ,or equal to 2% by weight of reactor water (60.02). The halogen activity concentrations in reactor water are presented in Table III-5. ! t l ;
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.is reactor water are presented in Table III-6. Fission products in BWR l reactor water are not adequately represented by simple equations. For
, ' these radioisotopes, concentrations in reactor water have'been estimated based on experience. Carryover of these isotopes into the steam is ,j
> Cestimated to be less than or equal to 0.1% by weight of reactor water
~ .($0.001). In addition to carryover, decay of noble gases in the steam 3a i leaving the reactor will result in more production of some of the 1 L ,
. isotopes listed. 1 )
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' is expected to be negligible, however, the isotopes may be observed in some samples in equilibrium or approaching equilibrium with the parent isotope.
Except for Np-239 which is listed in Table III-6 trace concentrations of transuranic isotopes have only been observed in a few samples where extensive and complex analyses were carried out. The alpha activity present is predominantly from Cm-242 et an estimated concentration' of10-6microcuries/ccorless. Alpha activity from plutonium isotopes is more than one order of magnitude lower than the activity from Cm-242. 1 Pu-241 (a beta emitter) may also .be present in concentrations comparable k to the Cm-242 level. p - b , ._p. U Oe en . . -, , v. .i . 2 ~> s a ' fs e .,. . ~ e. - .o..,
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- . .," 7
~
. '., ., M -- e U ,.* __r , , ,'.N 1
_. .r. . . ,..g ,
' i . ^f.+5, Ny
,"
- e< ~3s ,, ,- +
> I, 4 .f y' ,*;3 .
7 '.;,,cytm y j_ f-en ,' - yf.rs n s < n ,' ,y.
.f! -
r.,3 m ..
-- t
(
~
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,,,f,-.f .y _. 9 ; -{ -[:? ,.Q ' ; , x , ',* ,.
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.r-.-- ,:, . _ _
, .u,-. - ,. ,x- m -- ---
{.. l- . .. 11. P l h .* t . i TABLE III-l 1 REACTOR STEAM - C00IAN1' ACTIVATION PRODUCTS I 1
\
f COICENTRATION I ISOTOPE HAIF-LIFE (MICROCURIES/ GRAM) N-13 9 99 min 7 0 X 10-3 N-16 7 13 see 1.0 X 102 N-17 4.14 see 1.4 X 10~2 l 26.8 see 19 _ 7.4 X 10-1 F-18 109.8 min 4.0 X 10-3 . Gn u - s8% > r-e et,'
, - e e 4
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1'. . . 12. E e3 ; i l TABLE III-2
. I
?
REAC'IOR WATER - COOLAUT ACTIVATION FR0 DUCTS CONCENTRATION l; , ISOTOPE HAIF-LIFE (uCi/ce) N-13 9 99 min 2 9 x 10-2 , i
~ '
N-16 7 13 .sec 5.8 x 10 1 .
' N-17 4.14 sec 4.9 x 10-3 . .
. ~ .
.. . m, .
3 _ ._. . o 19 .. _ .26.8 sec . . 6.3 X 10-1 . w
.4 9' 1 F-18' -
1098amin 4.0 x 10-3 ;
.5 " W.
r e - r .,2
.,,'_.~
s-
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g n .sg .,4
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TABLE III-3 p REACTOR WATER - li0 tic 00LAllT ACTIVATI0ft PRODUCTS
- C0tiCErlTRATI0t1 (pCi/cc) i~ ISOTOPE HALF-LIFE
.s , Na-24 15 hr - 2 x 10-3 ,
-5 P-32 14.31 day 2 x 10
!' 27.8 day 5 x 10-4 Cr-51 -5 l' -
. Mn-54 313 day 4 x 10 l
Mn-56 2.582 hr 5 x 10-2 i
-3 i Co-58 71.4 day 5 x 10 Co-60 5.258 yr 5 x 10-4 _
8 x 10-5 Fe-59 45 day
)
ili-65 2.55 hr 3 x 10-3 - j Zn-65 . J5'.. _f. m. 243.7 day 1 x 10-6 -
-~
S- --13.7 hr _ ... 3 x 10~5 1;- j Zn-69m
-~ -
^ ~
I.I t :i 253 day 6 x 10-5
.. p
. Ag-110m' ' '~ _-
'mM
. .- :n _
-- - 3 x 10-3 ,
- ~' " - .-- 23.9 hr ;;
~ W-187
*M. g's es e w 9 *w- he wo .m y
~~, <
3 y_...- u.2 . .
'c~
~ k'?;~4- :
. - a . .w. . . _.
- - - . . - . ~ ., -
a,1,,.. w ,s -
~ , * * * - ..**q,
. + . , .
. i '
w2 3
. ::s .
~ _ - -
i j 4 i I 4 - s _2 . 5 - u
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" t .7 - v -
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4 .
~
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l,Q p ,
- - . ;. . -1 %
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-if
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9 .,
-a
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! . . TAl_tLE III l+ i i EMI5510f1 RATES OF NOBLE GASES I These rates are based on sufficient fuel cladding defects to result in a I total Of f ge ri Tease cat.: of 100,000 pCf/sec after 30 minute decay. RELEASE RATE RELEASE RATE l! Ot=0 9 t = 30 MIN. ISOTOPE' HALF-LIFE (uti/sec) (uCi/sec)_. l 3 3 1.86 hr 3.4 x 10 2.9 x 10 l Kr-83m 3 3 Kr-85m 4.4 hr 6.1 x 10 5.6 x 10
- Kr-85 10.74 yr 10 to 20
- 10 to 20
- 4 4 j Kr-87 76 min 2.0 x 10 1.5 x 10 4 4 Kr-88 2.79 hr 2.0 x 10 1.8 x 10 2
i Kr-89 3.18 min 1.3 x 10 5 1.8 x 10 5 kr-90 32.3 sec 2.8 x 10 ,,,,, Kr-91 8.6 sec 3.3 x 10 5 ,,,,, Kr-92 1.84 sec 3.3 x 10 5 ----- ". 4 7;
# Kr-93 1.29 sec ,,
9.9 x 10 - --. 4 2.3 x 10
~
Kr-94 1.0 sec ~
~
3 [ Kr-95 .5: 'sec 2.1 x 10 -----
. ; .. i
(@f j . 6! 1.4 x 10
~
~~
. .Kr-97 . ,_
.1 ,
sec Xe-131m . __ .; 11.96 day 1.5 x 101 1.5 i 101 ~ J' l
~
2 - - 2 i g1 2.8 x 10
~
~ -
Xe-133m ~ 7 2.26 day 2.9 x 10 3 3 Xo-133 - -5.27 ' day 8.2 x 10 8.2 x 10 - $~. Xc-135m
~
15.7 min 2.6 x 10 4 6.9 x 10 3 .], f Xe-135 9.16 hr 2.2 x 10 4 2.2 x 10 4 5 5 2 Xe-137 3.82 min 1.5 x 10 6.7 x 10 _ , 4 Xe-138 14.2 min 8.9 x 10 4 2.1 x 10 2 Xe-139 40 sec 2.8 x 10 5 ,,,,, ,' 5 Xe-140 13.6 sec 3.0 x 10 ,,,,, 5 l Xe-141 1.72 sec 2.4 x 10 ,,,,, ' 1.22 sec 7.3 x 10 4 --.--
-l Xo-142 '
! Xe-143 .96 sec 1.2 x 104 --- - 1 5.6 x 10 2 ,,;,,, ; ! Xe-144 -9 sec .: ;
+
6 5 'j TOTALS
- 2.5 x 10 s 1.0 x 10
.
- estimated fmm experimental observations .
c osa
)', i -?.. -
.ib c
. .. . . g
- 17. gj 1
y . %. . c'.
&]m .
k____.__ k- , . 15. l; TABLE III-5 b
\
l i REACTOR WATER FISSION PRODUCTS - HALOGENS i i CONCENTRATION ISOTOPE HALF-LIFE (uci/cc) l i h Br-83 2.40 hr - 2.0 x 10-2 i Br-04 31.8 min 3.8 x 10-2 Br-85 3.0 min . 2.3 x 10-2 I
.'f
.5, 1-131 8.065 day 1.8 x 10-2 1-132 2.284 hr 1.7 x 10-j .
hr 1.2 x 10-I
~
- 'i 1-133 ' 20.8 52.3 min 3.3 x 10-I -l l-134 -
6.7 hr 1.8 x 10~j ~ A I-135 - m_ , y . ~. r ~ 7,. 7- ,
, . . ./- _
-~ ~ '
- - '"y 3 . - ,
4 m *'
* * . , er g 1 - -
- - ;61 i
i , i i . Q R
"g d ;
iti
- 1. . .
i- -
'~
L , 3 [ .
, .r -
, . . n..
e s k - a ~~
' ' L - m d. '
,u-,', ', 17, j <
.e < ,c .
- . , , .gs -
,j s . .. - ~
, .A
*"' "*"~~*'N.--- ,w a* ? -- ~ .,_, , ,_
.u : ;
TABLC III-4 b
- 16. 1 I
I REACTOR 1l ATE!: IISS10il PP.000 CTS - 0 tiler IS0 TOPES
. C0f1CEilTRATIOt1 (b ISOTOPE IIALF-LIFE (uCi/cc) i Sr-89 50.8 day 2.7 x 10-3
. Sr-90 28.9 yr 2.1 x 10-4 Sr-91 9.67 hr 6.8 x 10-2 l Sr-92 2.69 hr . 1.2 x 10-l l t ~Zr-95 65.5 day 3.6 x.10-5 1 Zr-97 16.8 hr 3.0 x 10-5 lib-95 35.1 day 3.7 x 10 -5 Mo-99 66.6 hr 2.0 x 10-2 i Tc-99m 6.007 hr 2.9 x 10-I Tc-101 14.2 min 1.8 x 10-I Ru-103 39.8 day 1.7 :< 10-5 , i Ru-106 368 day 2.3 x 10-6 , Te-129m 34.1 day 3.5 x 10-5 Te-132 - 78 hr - 4.4 x 10-2 Cs-134 _ 2.06 yr 1.4 x 10-4 Cs-136 -
.13 day 9.4 x 10-5
' ~ ~ ~ ' '
Cs-137 30.2 yr 2.1 x 10-4
.32.2 min 2.4 x 10-I
~~1
. Cs-138 1.9 x 10-l
'~
Ba-139 83.2_ min n,
~- 12.8 day 8.0 x 10-3 \
Ba-140 2.3 x 10-I
~
Ba-141 18.3 min Ba-142 10.7 min 2.3 x 10-I Ce-141 32.53 day 3.5 x 10-5 Cc-143 33.0 hr 3.2 x 10-5 ! Ce-144 284.4 day 3.1 x 10-5 Pr-143 13.58 day 3.4 x 10-5 j I!d-147 . 11.0G day 1.3 x 10-5 ilp-239 2.35 day 2.2 x 10-I
\.
~
- O a '
u
' ~
g
- < ~ . , e
1 )*' p 17 III. (c) Design Basis Source Term Calculational Models and Uncertainties 14 l p\ Equations (1),(2),and(3)givetheformofthecalculational
- codels used to estinate noble gas and halogen release rates for input
- to radwaste processing equipment.
1 [ Noble Gas Release Rate (1) R$
= 2.6x107 .y x 0.4 (i.e Aj T) (e Ajt)
Halogen Release Rate (2) Rj = 2.4x10 7 yx 0 5 (j_e A T) j (e-Ait)
, Halogen Reactor Vater Concer.tration (3) Cj =
a V (Ai +Y) I Where: Rg = Releaserateofisotope1(pCi/sec) _ ; yg - Fissionyieldofisotope1(atems/ fission) __ 1 1g = Decay constant of isotope 1 (sec~I) T = Fuelirradiationtime(sec)
.t - DecAytimefollowingreleasefromfuel(sec) s Cg - Concentration of isotope i in reactor water (pCi/co)
(density of water p 1.0 g/ce) 9 TV ce -
= Volume of water in operating reactor (/cc))
(densityofwaterp.074g . _;l 1
- p = Cleanupsystemremovalconstant(sec -') ?
j Cleanup system flow rate (g/sec) p ,_ Quantity of water in operating reactor (g) 5 T = Steam carryover renoval constant (sec"I) for halogens
, y , (0.02) (steam flow (g/sec))
{ Quantityofwaterinoperatingreactor(g) 1 4 t 1 . U F s Y . t . e, *1 s -
~
18.
\
Noble Gas Release Rates v The release rate of noble gases can be expressed by the simplified formofequati6n(1): R = K yA* c 9 9 L h The observed experimental data have shown a variation in individual
; noble gas isotopes with respect to each other that can be expressed in terms of variation in m, the exponent of the decay constant term (4).
[1 The average measured value of a was O.4 with a standard deviation of 1 0.07. With the iRi G 30 min set at 100,000 uCi/sec,thevalueofKg is calculated after selecting the value for m. Variations from any mixture selected as a reasonable design basis have been observed, and should be expected. Small defects from t a fairly new core will tend to give mixtures toward the equilibrium
' end of the spectrum. Tramp uranium on surfaces from prior operation with significant cladding defects will give mixtures toward the re-coil end of the spectrum. The mixture selected as a design basis is that which, on the design basis of current observations, may be repre-sentative over the years of operation of a plant. Variations in ~
noble gas emission around the selected composition in the range from j m=0.2 to m=0.8 may occur. u 4 .- a Halogen Release Rates ~' f, ,
?
~ The release rate of halogens can be expressed by the simplified g, ~formofequation(2): n
~~
a , Rh
- E hYA
.y The observed experimental data have shown a variation in individual n b
. halogen isotopes with respect to each other that can be expressed in -
r" terms of variation in n, the exponent of the decay constant term ( A.). U k- The average measured value of n was 0.5 with a standard deviation of ~< jl
-Is+ 0.19. With I-131 calculated release rate after selecting the set value at 700of n.uCi/sec, the value of ( a
' d? '
3 w 1, Halogen Carryover e . The observed experimental data on halogen carryover provides an I q average value of 1.2% by weight of reactor water with a standard devia-y tionofg0.9%. , Other Fission Products .<
- ~
The concentrations of other fission products in reactor water are predictable with an estimated accuracy of + 100% to - % , as a general rule, for a given fuel leak rate condition.
. =
4 1
,,a l _ '
k1 f 4 s
*
- e t i
'?b
; 6pf p <> ,, 9 ,
) __ _
__'___ _ '.- , [<_.-. .s_ _ _ (_' ._ _tl,3 0
@d
19 IV.
~
Identification of Radioactive Effluent Pathways to the Environment U -(a) Gaseous Effluents Gaseous radioactive effluents from the Pilgrim Nuclear Power Station can be released to the environment through two types of pathways: Process Systemswhich and discharge radioac,are designed tive gaseous to routinely effluents. collect, Systems process in this category include the main condenser gas re= oval and offgas systems, turbine gland sealing and holdup systems and the ' , mechanical vacuum pump. Auxiliary Systems which in addition to their design function , have a potential for releasing airborne radioactivity resulting from equipment leakage, tank vents, su=p evaporation, spillage, and equipment testing. Systems in this category include the reactor and turbine building ventilation syste=s and the primary containment atmospheric control system.
- 1. Main Cond'ensar Gas Removal System The main condenser gas removal system is described in FSAR sections 11.4.3.1.1 and 9.4.4.2. The purpose of this system is to achieve and maintain condenser vacuum by removing non-condensible gases. These non-condensibles include radiolytic hydrogen and oxygen, water vapor, air and trace a=ounts of gaseous radioactive products. The radioactive 7 products re=oved by this system represent over 9% of the radioactivity available for release through gaseous effluents. The existing offgas system delays these non-condensibles for thirty minutes and then releases them to the environment through the offgas filters and the main station ,
stack. Subsequent to the installation of the augmented offgas system , ; (described in Section V.a of this Amendment) the non-condensibles will be processed to reduce their volume and increase the delay period before ' j filtration and release through the main station stack. ~i
- 2. Turbine Gland Seal Holdun The turbine sealing system is described in FSAR Sectior.a 9.4.4.3 and 11.4.3.2. The purpose of this system is to minimize condenser air in-leakage '
by maintaining a steam seal on the turbine packing glands. Thesteam/ air exhaust from this system passes through a gland seal condenser where the stesm is condensed and the non-condensibles are exhausted to the gland seal holdup line. The radioactivity processed by this system is proportional
, to the amount of steam utilized in the sealing process which is less than
.1% of the full power steam flow. The gland seal holdup cons * , s of a 1.75 minute delay before release through the main station st.,
/D \
k ) v , s 1 e 4 q
}
+
. Y 20.
,m 3. Mechanical Vacuu: Pumn f 1 k The mechanical vacuum pump is described in FSAR section 11.4.3.1.2.
The purpose of the pump is to supplement the main condenser gas removal system during startup when steam is not available to the main condenser gas re= oval system or the volume requirements exceed the capacity of the main condenser gas removal system. The radioactivity expected to be released from this pump is negligible, however, there does exist the potential for release, therefore, the discharge is directed to the gland seal holdup line before release through the main station stack.
- 4. Station Ventilation Exhausts The reactor building ventilation system is described in FSAR section '
10 9 3.3. The purpose of the system is to maintain desired ambient conditions within the reactor building, turbine building (below the operating floor at elevation 51') and the radwaste building utilizing once-through ventilation without recirculation. Normal discharge of radioactivity through this system is not anticipated, however, the potential does exist for airborne radioactivity resulting from sampling, process leakage, tank vents and su=ps to be discharged to the environment through the reactor building ventilation exhaust. The turbine building ventilation system is described in FSAR section 10 9 3.4. The purpose of the system is to maintain desired ambient conditions above the turbine building operating floor at elevation 51'
.g '
utilizing once-through ventilation without recirculation. Normal discharge - V> of radioactivity through this system is nct anticipated, however, the ._ potential does exist in certain areas of the turbine building for airborne radioactivity resulting from process leakage to be picked up by the turbinc ~ building ventilation system. Areas of low potential for airborne /l radioactivity such as the operating floor are ventilated by exhausting . directly to the environ =ent through roof exhausters. Areas below the
~
7l 6 operating floor which have a higher potential for airborne contamination are ventilated by exhausting to the Reactor Building Ventilation Exhaust System for discharge with continuous monitoring through the reactor i building vent stack. ' 5 Primary Continment Atmospheric Control System i The primary containment atmospheric control system is described in l FSAR section 5.2.3.8. The primary containment is normally a sealed volume inerted with nitrogen gas. Periodically, makeup gas is auto =atically added to the primary containment to maintain the desired nitrogen concentration. Under special conditions when personnel access to the containment is required, such as during refueling or maintenance periods, the primary containment is de-inerted by purging the containment atmosphere. During this period of de-inerting the potential exists for the release of airborne radioactivity to the environment. The systems are arranged such that the purge exhaust can be directed to either the reactor building ventilation exhaust or to the standby gas treatment system for release through the main station offgas atack. g , I ) I a c . _ _
21. IV. (b) Liquid Effluents All potentially radioactive liquid wastes are routinely collected and processed through the clean, chemical, or miscellaneous radwaste systems. Liquid radvaste effluents from the above systems can be released to the environment through only one pathway, the liquid radwaste discharge header.
- 1. Clean Radwaste System The clean radwaste system is described in section 9 2.4.1 in the FSAR. Clean liquid wastes consist primarily of equipment leakages and radwaste concentrator distillate and are normally low in conductivit'/ .
The purpose of the system is to minimize radioactive releases to the environment and to recycle liquid wastes for reuse within the plant. Clean liquid wastes may be discharged at a controlled rate and monitored via the discharge header after holdup, sa=pling, and analysis if it does
- not meet reactor feedwater quality or if there is an excess inventory of high quality water. The wastes will be released *into the circulating -
water discharge canal.
- 2. Chemical Radvaste System
.The purpose of the chemical radvaste system is to process chemical
- Q wastes and minimize radioactive releases to the environment. The primary sources of chemical wastes are the regeneration of the condensate demineralizers and floor drainage. The radwaste concentrator reduces che=ical vaste volumes which in turn increases holdup time and decreases
- .l activity releases. After holdup, sa=pling, and analysis the wastes may _{
be discharged at a controlled rate via the liquid radwaste discharge e{ header to the circulating water discharge canal. Subsequent to th'e installa- ' tion of the Augmented Chemical Radwaste System as described in Section VI, . .; releases will be further reduced. 1
- 3. Miscellaneous Wastes a The miscellaneous waste system is described in section 9.2.4.3 in the FSAR. Miscellaneous wastes consist of 1cw radioactivity decontamination washdown water (generally characterized by high detergent levels) used in the cleaning of various plant components. The miscellaneous vastes are held up, sampled and analyzed. These vastes are filtered prior to discharge to the circulating water canal via the radwaste liquid discharge header. Miscellaneous wastes can be transferred to the chemical waste system for further processing if required.
4
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22. i
% V. Control of Gaseous Effluents 4
) (a) Augmented Offgas System 4
- 1. Design Objective The design objective of the Augmented Offgas System is to control the release of gaseous radioactivity to the site environs so that the -
total radiation exposure to persons in unrestricted areas is as low as I e , practicable.
- 2. ' Design Basis .
The Aug=ented Offgas System is designed to limit offsite dcscs from rout $ne plant releases to levels which are as low as practicable and to stay within the limits established in the station operating license. The
~
Augmented Offgas System is designed to provide adequate time for corrective action to li=it the activity release rates should they approach established limits. As a design basis for the system, a noble gas input equivalent to
- an annual average offgas release rate (based on 30-mirute decay) of _; .
~100,000 microcuries/secondwasused. Table V-1 indicates the design ~~.
basic noble gas activity referenced to 30 minutes and the noble gas -
*" - activity after processing through the Augmented Offgas System. Also shown in Table V-1 are individual noble gas isotope activity reduction. factors
@. ,_ and the overall activity reduction factor provided by the system. The ,, ;E
- Augmented Offgas System receives the non-condensible gases discharged
- from the main condenser. These gases and their volumetric flow rates ~7 3
4 .. 1 are given in Table V-2. The air in-leakage design basis for the Pilgrim - 1., Nuclear Power Station Augmented Offgas System has been established at - - T, t
~
7 ft /=in 3 (at 1300F,1 atm.) per condenser shell. Ieakage from two . J'C condenser shells (corrected to standard conditions) gives a total of 12.3 3 standard cubic feet / min, the design basis air in-leakage for the system. j:1 7 N.
. Generally, in-leakage varies from about 3 to 5 standard ft3/ min per shell, for large shells. Where special housekeeping is employed, and leaks are 'j .
detected and sealed, leakage can be reduced to 1 to 2 standard ft 3/ min ",;
] _
per shell and can be held at that level during extended station operation. , Air in-leakage at three operating boiling water reactors where condenser " in-leakage has a significant effect on offgas holdup time is given in Table V-3
~
[ 3.. 3._ System Description The offgas treatment system shown in Figure V-1 uses a high temperature I catalytic recombiner to recombine radiolytically dissocis.ted hydrogen and ' oxygen from the air ejector system. Non-condensible radioactive offgas
'is continuously re=oved from the main condenser by the air ejector during M plant operation. The air ejector offgas normally contains activation y q.'[a '
gases, principally, N-16, 0-19, and N-13 The N-16 and 0-19 have short half lives and quickly decay. The 10 minute N-13 is present in small i j 3 amounts which are further reduced by decay. The air ejector offgas also j i b contains the radioactive noble gas parents of biologically significant 2
, ;-q
],
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--.,l ,
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. q i
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l f .. .. [ 23. ( TABLE V-1 , i 5 DESIGN BASIS ESTIMATED OFFGAS RELEASE RATES i J (Release Rates in Microcuries/Second) i Discharge from Existing Discharge From Offgas Activity 30 Minute Charcoal Reduction Isotope Holdup Line Adsorbers Factors 3 -2 Kr-83m 2 9 X 10 5.1 X 10 57,000 i 1 Kr-85m 5.6 X 10 3 5,7 x 10 91 . Kr-85 10 - 20 10 - 20. 1 Kr-87 1.5 X 10b~ 1 7 X 10-3 8,800,000 b 1.8 X 10 b 1.2 X 10 1 Kr-88 1,500
~
2 g Kr-89 ~1.8 X 10 y,yy t,,g, g h Xe-13 h 1 5 X. lo. l
'2 4
~1 3.8 . . .
;f
.2.8 X l0
-Xe-133m -
2.6 X 10 - 1,100
$ 2 '
~ ~
2 y Xe-133 - ,
'6'.2 X 103 g,5 x 19 18 ~ ~ ~ ~4'
'Xe-135m 19 X ld3 ~ 0 very Iarge
' ' ' ].
~ Xe-135 2.2 X lob .
O Very Iarge .!fl' 1 - Xe-137 6.7 X 10 2 0 very Iarge [
;Xe-138 2.1 X 10 b 0 very Large Overall k.
TOTAL Approx. 100,000 Approx. 540 Offgas 185 ~
- Activity
' Reduction , Factor i $ e s . .1 7 . 4 ,1 1
.k h
l l
- h. -
. x 3
- 7. , .
- c. . ,-
. v;;.
.3
, d <-jy
l h, 2u. r TABLE V-2 t l .
, NONCONDENSIBLE GAS INFUTS TO AUGMENTED OFFGAS SYSTEM g.
e I. i Hydrogen 3 90ft/ minute l From Radiolytic f. 4
'0xygen . Decomposition of Water 45ft/ 3minute 1
Wat'er Vapor (to saturate) 25ft/ 3minute L Air . 12.3ft/ 3minute ( ', TOTAL 172ft3/ minute - e e ti
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- TABLE V-3 4 ,
. l l
f Number of f Condenser Type of Augmented Total Air In-Leakage We-Plant Shells Offgas System (Standard ft3/ min) ! KRB 250 1 Reco=biner/ Charcoal 4.1 l Tsuruga 342 1 Reccmbiner/CompressedGas 4.7 Fukuei=a 1 440 2 Recombiner/CompressedGas 70 I. , This data is consistent with leakages observed at conventional power plar.ts . . _ , , . i ivhereleakageratesrangefrom,2to11ft/minforunitsinthe 3 350-500 ~ W e range. .- Q
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27
.y ~ s Sr-89, Sr-90, Ba-140, and Cs-137 The concentration of these noble gases
{4 ; depends upon the amount of tramp uranium in the coolant and on the reactor k/ fuel cladding surfaces (usually extre=ely small), and the number and size of fuel cladding leaks. Afterhydrogen/cxygenrecombinationandchilling to strip the condensibles to reduce the volure, the remaining non-condensibles (principally kryptons, xenons and air) are delayed in a 30 minute holdup system before reaching the adsorption bed. Radioactive particulate daughters of the noble gases are retained on the HEPA filters and on the charcoal. The charcoal adsorption bed, cperating in a constant temperature vault,
. selectively adsorbs and delays the xenons and kryptons from the bulk carrier gas (principally air). This delay on the 'harcoal c permits the Xe and Kr to decay in place. The offgas is discharged to the environs via the main stack. The activity of the gas entering and leaving the offgas treatment systec is continuously =onitored. This system results in a reduction of the offgas activity (curies) released by a factor of approximately 185 relative to a 30 minute holdup system as shown in Table V-1.
The adsorption of noble gases on charcoal depends upon gas ficw rate, holdup time, rass of charcoal and a gas unique coefficient known as the
' dynamic adsorption coefficient. The pararetric ihterrelationships and governing equations are well proven from three years of operation of a similar unit at KRB in Germany.
Ihe design require =ents fer' the equip =ent of the offgas system are
. given in Table V-4. The Augmented Offgas System is designated ceismic Class II. This class includes those structures, equipment, and components h* which are important to reactor operation, but are not essential for preventing an accident which would endanger the public health and safety, and are not essential for the mitigation of the consequences of these accidents. A Class II designated item shall not degrade the integrity of arg item designated Class I. J S
The " front-end" components of the system are installed in the existing
. turbine building. The charcoal adsorbers and associated auxiliary equipment ,'
are installed in a new building whose access doors are-at elevation 23'. - As described in Amendment 19 to the FSAR, station structures at elevation 23' are not subjected to flooding. The system is not designed to be functional during or after a tornado. The system is not essential for the prevention of cecidents nor is it , essential for the mitigation of the consequences of such accidents. The Air-Ejector Offgas System is provided with flow, temperature, and radiation instrumentation to insure proper operation and control. Hydrogen analyzer instru=entation is also provided to insure that hydrogen
' concentration is maintained bcicw the flarrable limit. Table V-5 lists process cystem alar s and their location. *
~;
Air-ejector offgas radiation monitoring is divided into two subsystems. ' One subsystem takes a continuous sample from the offgas line just after the offgas system condenser. The other takes a continuous sample from the offgas system just before discharge to the main station stack. The I (7 former subsystem is described in section 712.2 of the FSAR. Additional ' monitoring and control action is now performed by the subsystem which . samples the offga's stream prior to release. # y 4,
-1,
28. cym TABLE V-4 . EQUIRENT DESIGN REQUIRE 4ENTS Principal Component Principal Construction Code (*)
- 1. Tanks D100 or API-650 & SR(a)
- 2. Heat Exchangers III-3 & TEMA-C (containingoffgas) .
3 Piping and Valves B31.1.0 (auxiliarysystems)
- 4. Piping and Valves III-3 (containingoffgas) 5._ , Pumps III-3 . _ . ...?
-6. ~ 3alves, Flow Control - III-3 ' ~ '~fk m7.1 Charcoal Adsorber Vessels / III-3 . .
.. 3;
- 7 ^
+, =
.eQ .y
.f~; 9
;w J H *~ Legend .
, n1 _
' ~ j,j c .D100 .. . , .. . AWA-D100, Standard for Steel Tanks, standpipes, '1.;
...~ ,
*~ ^ ~
2 Reservoirs, and Elevated Tanks for Water Storage .; API-650 , API 650, Welded Steel Tanks for Oil Storage
- . -= :.:.. ; .\
^
SR(a) . Nondest'ructive Tests Examination Requirements per _] _ , . m._ _ ASME Section VIII, Division 1 _.; j
.m .q
- III-3 _- .'
' ASME Boiler and Pressure Vessel Code, Section III,
. . . . . . Class 3 :j
- s . c.;
~ . [ TB4A-C . ._
Tubular Dtchanger Manufacturers Association, Class C , i' B31.1.0 ,, ,
,US S B.31.1.0, " Power Piping" f
*jl
+: . .. .
\
-General Notes:
? i
-(l) This table applies to equiIn::ent' and piping procured for-the Augmented l E~~ Offgas System after July 1,1971.
d
- 3. . 3 i(2)' .The equipment for the Augmented Offgas System is seismic Class II.
d n : Site boundary doses resulting from failure of system components do : .), not exceed 10 CFR 20 whole body limits. a .
= -
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29 f i
- ' TABLE V-5
. .?
PROCESS INSTRUFENT AIARMS 3' . Control Roo:n Iccal 1 Indicated Recorded Indicated 4
- 1. Preheater Discharge temp.--low X
- 2. Recombinercatalysttemp.--high/ low ,
X --
- 3. Offgas Condenser Drain Well (dual) -- --
X ). Ievel--high/ low l+ . Offgas Condenser gas discharge -- -- X '
; . temperature--high 5 Hydrogen Analyzer (Condenser discharge) -- X _-
(Dual)--high ,
,6. Gas Flow (Offgas Condenser discharge)-- -- X' X "
high/ low - m - 1
; -, .m 7 Cooler--Condenser discharge --
X - ,C temperature--high/ low - J'.j
.,.'l.f .. ' - -
l _. _ : - -- _
. m.
~
'8..
. Glycol solution temperature-- .X X ' 1X _ ~i low .W
;- . hi - -
.~
..,.. .. ._ - , n- _.
'9
, ... ~
'__ . ,_ , 9 _
Gas reheater discharge humidity--high -- =X ,
-4
~
4 i
- >; 10. : Prefilter AP--high - ~.
X -:-- .
. Q'2
, - - , . , . . .. # -l, ..e)4
- 11. . _
Carbon bed. temperature--high -- 'X -- M
- L:4 l 12. .Carbonvaultte=perature--high/ low X X -- ~ .'j
-- , u - .a
'X ~I l
l _.
- 13. After filter A P--high l --
.X __s-_. ,
'a ,,
.e ,
Instrumentation ele:nents: "' f; a .
.. tj
- - Temperature--thermo. couple ~ 7. ,
; a - -
, _ ji l Level--differential pressure diaphragm ;j "1
- 4 _
, Hydrogen--thermal conductivity q a ,
.t
" ' ~ ^
7 Gas flow-flow orifice ' ' l
,;js
~
{ ,/- Differential Pressure--differential pressure diaphragm
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- 30.
The subsystem r'onitoring the offgas system upstreem of the main k' station stack has two instrumentation channels. Each channel consists
\
of a gamma-sensitive detector, a pulse preamplifier, a logarithmic radiation monitor with a power supply and a meter, and a strip chart recorder point. The monitors and the two-pen recorder are. located in the control room. Each logarithmic radiation monitor is powered from a different bus of the 24 volt d.c. system. The two gamma-sensitive scintillation detectors are mounted in two shielded sample chambers. The sample is drawn from the. offgas line through the sample chamber by the sample pump. Each monitor
. has three upscale trips and a downscale trip. An upscale trip indicates high radiation. A downscale trip indicates instrument trouble. 'Any one trip will give an alarm in the control room. Any one upscale.high radiation trip closes the carbon bed filter bypass valve, if open, and opens the offgas line to the carbon bed, if closed. Two upscale high-high-high radiation trips, one upscale high-high-high radiation trip and one downscale trip or two downscale trips isolate the offgas system outlet and drain ,
valves. .
^
The air ejector' offgas radiation monitors have monitoring characteristics
. sufficient to provide accurate indication of radioactivity in the air : D, ejector offgas. The system provides the operator with sufficient information '
to monitor the performance of the Augmented Offgas System. Sufficient _d redundancy is provided to allow maintenance and testing of the radiation ,j
, - . __ -monitori.ng system.3. Each channel can be .c.alibrated by analyzing a grab
. 7:f5 7 ,
7 ; - ; ,y- .g .g . a:. - . . - . . - - . . . . . - . h ,, . ,
," i . Figure V-2 shows the location and arrangement of the " front-end" '~ .-Q
] components of the Augmented Offgas System in the existing turbine building. .f '
. .. , Figure V-3 and V-4 show the arrangement.of the charcoal adsorbers and ~5
- 4 y ladsorber. auxiliary equipment in a new building located approximately 20 -3
- - - feet south of the existing turbine building between column lines 3 and 8. . . ei
^
,.
- The building will be approximately 68 feet by 72 feet in plan and extends ,y C 2O feet above and below grade. The elapsed time for procurement, installation -
J'
- .: ~ . 'a^ fand checkout ,
of the
;; - . Augmented
. . . - ~ ~ Offgas T System is estimated to be 26 months.
.33q
_ 4. _ . Safety Evaluation , ; , [
- n. .
.: m
- a- - - The decay time provided by the Augmented Offgas System permits ~ lM l
'significant radioactive decay of the activation gases and fission gases in the main condenser offgas prior to release. The adsorbers provide a ,[ ;
F
;_ T , i 22 day xenon and a 29 hour krypton holdup. The solid daughter products - ,
cof the decay of the nobic gases are removed by filtration and/or are retained ;
. ' t '"'- [on the charcoal. Final filtration of the charcoal adsorber effluent ;d f h precludes escape of charcoal fines. Particulate activity release is ^'d,
, 7 expected to be negligible.
s er L , Q'
<g
- t
_ r q - r "; , 5 .-, s Iodine' input into the offgas system is small because of its retention .j
; - 7J
., Vin reactor water and condensate. Additional iodine removal is provided by 1 ,
d Tateam condensation which occurs in the offgas condenser located downstream
~
. s: 1 -
, (of th'e hydrogen recombiner. Minute quantities of iodine entering the
~
- U
. charcoal absorbers are further absorbed.
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l r-- , TG s[~ . l.h me i - tw NmT 1 T-F0C [ y; - PRELIMINARY NOT TO BE USED FOR CONSTRUCTION P!LCRim HUCLEAR power STATION
' FINAL SAFETY ANALY5t$ REPORT EQUIPMDrr IDCATION AND ARRANCDfElfT ,
AUCMDITED OFTGLS SYSTEM ; CHARCOAL ADSORBERS TESSELS AND l ,~ AUZILIART EQUIPMDIT. ELEVATION 8'-O" l' l FIGURE T-3
m - m .. w' m---
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FLAN At0VE s PRELIMINARY NOT TO BE USED FOR CONSTRUCTION P!LGRim HUCLEAR POWER STATION FINAL $AFETY ANALY$l$ REPORT EQUTFMENT IDCATION AND ARRANCDG2fT AUCMENTED CFFCAS SYSTDI C11AR00AL ADSORBER TES$ELS AND AUZILIARY EQUIPMENT, FLAN ABOVE w FICURE V-4
+
. g '
l______._-______ _
31+. ym Radiation monitors at the recombiner outlet continuously monitor {Q S activity input levels to the charcoal adsorber beds. A radiation monitor
, is also provided at the outlet of the charcoal adsorbers to continuously monitor the release rate from the adsorber beds. This set of monitors provides diagnostic information on the perfomance of the charcoal bed -
holdup system. The charcoal bed adsorber radiation monitor also automatically isolates the offgas system in the event of high radiation levels in order to prevent treated gas of unacceptably high activity from discharging to the atmosphere. Shielding is provided for offgas system equipment to maintain safe radiation exposure levels for plant personnel. The equip =ent is principally operated from the control room. The charcoal adsorbers operate in a massive temperature controlled vault at 770 F so that upon system shutdown, radicactive Bases on the adscrbers will be subject to the same holdup time as during normal operation, even in the presence of continued air flow. The adsorbers are maintained at a constant temperature by an air conditioning system which removes the decay heat generated in the adsorbers. Failure of the air conditioning y
^
system will cause an alam in the control room. In addition, a radiation monitor is provided to monitor the radiation level in the charcoal bed ~ vault. High radiation will cause an alau in the control room. The hydrogen concentration / of the Eases from the air ejector is . maintained below the fhable limit by maintaining adequate steam flow ^ Q h for dilution at all times. This steam flow rate is monitored and alamed. The preheaters are steam heated rather than electrically heated in order - to eliminate the presence of potential 16nition sources and to limit the temperature of the gases in the event of cessation of gas flow. The recombiner temperatures are monitored and alamed to indicate any N deterioration of performance. A hydrogen analyzer dcwnstream cf the .$ recombiners provides an additional check. j The air ejector offgas system operates at a pressure of about 5 psig or less so the differential pressure which could cause leaka6e of radicactive , gases is small. To minimize the possibility of leakage of radioactive - gases) the system is welded wherever possible and bellows seal valve stems or equivalent are used. Operational control is maintained by the use of radiation monitors to assure that the release rate is within the established limits. Envircrzental monitoring is used to determine resultant dose rates and to relate these
- to the celease rates as a check on station perfor=ance. Provision is also made fcr sampling and periodic analysis of the influent and effluent gases for purposes of detemining their composition. This infomation is used in calibration of the monitors and in relating the release to environs dose. The operator is thus in full control of the system at all times. '
l Table V-6 contains a detailed malfunction analysis indicating consequences ~" of failure of various components of the system and design precautions taken to prevent such failures. . 3
~
3
.. A s
x a L
p;jyL; .. - x ;' ,
. ,p V 1
\ w.-- ;
TABLE V-6 EQUIINENT MAiFUNCTION' ANALYSIS l f MUIIMENT ITD4S MALFUNCTION CONSEQUENCES , DESIGN PRECAUTIONS
$
- A,' Preheaters 1. Steam leak Would further dilute process offgas. Steam Spare preheater.'
consumption would increase.
- 2. Iow pressure Recombiner performance would fall off at low Iow temperature alarms
>- steam supply power level and hydrogen content of recombiner on preheater exit and gas discharge would increase, eventua14 to a recombiner inlet, combustible mixture. ' B. Rtcombiners 1. Catalyst Temperature profile changes through catalyst. Temperature probes in
, gradually Eventually excess hydrogen would be detected recombiner and hydrogen deactivates by meter. Eventually the gas could become analyzer. Spare recombiner.
combustible. .
- 2. Catalyst gets Hydrogen conversion falls off and hydrogen Condensate drains, temperature wet at startup is detected by downstream analyzers, probes in recombiner. Air Eventua]Jy the gas could become combustible. bleed system at startup.
Recombiner thermal blanket, L - , spare recombiner and heater. Hydrogen analyzer. 4 ; C. Offgas 1. Cooling water The coolant would leak to the process gas none. Condenser leak (shell) side. This would~be detected if drain well liquid level increases. Moderate leakage would be of no concern from a process standpoint. D. Drain Well 1. I/iquid level If both drain valves fail to open, water will Two separate drain systems, instruments build up in the condenser and pressure drop each provided with high and 4 fail will increase. low alarms. The high Ap, if not detected by instrumentation could cause pressure buildup in the main condenser and eventually a reactor scram, w
?;d:. ' '
s ' , - Y . .;.:.a h..u e .sa..u :a: e _.J. s~ : -
,1; ..
-6 a ,. .
If a drain valve fails to c3nse, gas will recycle to the main condenser increasing the load on the SJAE and causing back-pressure on the main condenser, eventually causing a reactor scram, i E. Water separator 1. Corrosion of High quantity of water collected in 30-minute Stainless steel'esh m wire mesh holdup line and routed to radwaste. specified. element i F. Cooler: 1. Corrosion of Glycol-water solution would leak into process Stainless steel finned tubes candensers finned tube (shell) side and be discharged to clean radwaste. specified. The inventory
- It would be detected at radwaste prior to of glycol solution can be being discharged to the reactor condensate observed in tank. Spare system. cooler condenser provided.
- 2. Icing up of Shell side of cooler could plug up with ice, Design glycol water solution finned tube gradually building up pressure drop. If this temperature of 35 40 F.
happens, the spare unit could be activated. Spare unit provided. Complete blockage of both units would increase Temperature alarms. main condenser pressure leading to a reactor
- trip.
G. Moisture 1. Corrosion of Increased moisture would be retain'e d in process Stainless steel mesh ceparators wire mesh gas routed to charcoal adsorbers. Over a long specified. Relative humidity element period of time, the charcoal performance would ' instrumentation provided. deteriorate due to moisture pickup. Spare unit provided. H. Gas reheater 1. Tube leak Steam would leak into the process gas and Stainless steel tube. Iow increase gas relavive hamidity. Eventually, pressure steam. Moisture the high moisture would cause deterioration recorder and high moisture of charcoal perfomance, alarm. Pr; filters 1. Hole in filter More radioactivity would deposit on the op instrumentation provided. I. media charcoal in the first adsurber vessel of the Spare unit provided. train. This would increase the radiation
~
level in the charcoal vault, making maintenance,' i more difficult. l l I . _ . ! w M X:..GJU :i . ' a '. ., .2. ,, ~j. . A
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v . . High 4 instrumented, l
; Charcoal performance will deteriorate O?J.1 Charcoal. 1.- Charcoal gets
. fadsorbers wet gradually as charcoal gets wet. Holdup times mechanically simple gas
[ ~ l'~ for krypton and xenon will decrease and plant dehumidification system
. emissions will increase. with redundant equipment.
L*5* !n .
- K.'; Vault air
; 1.' Mechanical s If ambient temperature exceeds about 80 F, Spare refrigerator unit failure : increased emission could occur. provided. Vault temperature
"- conditioning .,
.%;' , / units ,
alarm provided. QS q 6
~ 'i-1 '
!If ambient temperature is below about 60 F, Vault temperature alarm
]p[,;M / . ; charcoal could pick up additional moisture. provided. Q (-g '.
- m. j L.7 After filters 1. Hole lin filter Probably of no real consequence. The charcoal dp instrumentation and high l ,3 7 . media ' media themselves should be a good filter at the op alarm provided. Sparc unit provided.
M"C y , 1
- low air velocity.
- _r-c' Spare refrigerator provided.
t, N M.IGlycol . 1. , Mechanical If spare unit fails to operate, the glycol fannre solution temperature will rise and the Clycol solution temperature
. j ! refrigeration d M , machines. . dehumidification system performance will alarms provided.
' deteriorate. This will cause gradual buildup
& 4 ' MJ6 - of moisture on the charcoal, with increased
' plant emissions. '
LM. .in g M~N.L. .. .f.#j. ..
~
.~
Steam jet
~
- 1. Iow flow of When the hydrogen and oxygen concentrations Alarms provided for low
%w . tejectors =. motive high exceed 4 and 6 volume percent, respectively, steam flow and low steam the process gas becomes combustible. pressure. y& N7 . pressure steam 6 Steam flow to be held at
.- ~5 Inadequate steam flow will cause overheating and deterioration of the catalyst. constant maximum f3cw regardless of plant power level.
- 2. Wear of steam Increased steam flow to reccabiner. This jfy supply nozzle 'could reduce degree of receabination at low
>/ .. .;- of ejector power levels. e e,, , w'. .+: ek i M' ,, ce'4/ T , ,
, ,, y, , , ,,; , ,3 ; s lid NiidM, : ;.M d ,l h ;;h O'9 ',~'< j ,b y 4
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38. 5 System Operation
\ Non-coridensible gas re=oved from the main condenser, including air in-leakage, is diluted with steam to less than I %t (by volume) hydrogen -
concentration in the steam . jet compressors (See Figure V-1). The diluted
.offgas is superheated and then passed through a catalytic recombiner to convert the hydrogen and oxygen into water. The offgas effluent from the recombiner, containing only traces of hydrogen, is passed through a condenser (cooled by reactor feedwater) to remove the bulk moisture, and
. then to a 30 minute holdup for the decay of the N-13, N-16, 0-19, krypton and xenon isotopes. Decay daughters and iodine are removed by condensation on the walls of the holdup pipe and further removal of the decay daughters
~
is effected by filtration. The offgas is processed by a cooler-condenser to remove additional moisture and iodine, a de-entrainer and reheater to reduce the relative humidity, and a high efficiency filter prior to entering the charcoal adsorbers. 1 The charcoal adsorbers provide further delay of the xenon and krypton { isotopes in the offgas. Two parallel trains of adsorbers are used to i minimize back pressure. Heat is removed from the vault housing the adsorbers j to maintain the charcoal beds at an approximate constant temperature of i 77 F. The offgas effluent from the adsorbers is passed through a:mther high efficiency filter prior to discharge to the offgas stack. T ..
. ./ -.
u No ' dilution air is added to 'the offgas streem during steady state ; operation. 'The air present during operation is from air in-leakage into ~~
- ~ ~
m g the main condensers which operate at sub-atmospheric pressure. Oil free ;~i
- air is bled into the system during startup of the system. Its flow rate is 56.7 pounds per hour, which is stopped after the recombiner comes up _
- ._ to temperature. Air is supplied during recombiner startup in order to 9, prevent wetting of the recombiner catalyst and subsequent deterioration ,y
'of the hydrogen recombiner performance, j
.L 1'
~ ~ In the event of failure of a non-redundant Augmented Offgas System .9 component, it is intended to make provisions to bypass the Augmented ]
Offgas System and operate the station using the presently installed 30 .2 minute offgas holdup system until maintenance of the Augmented Offgas y-System could be completed. ~
- 6. Inspecticn and Testing C The Augmented Offgas System is used on a routine basis and does not 4- require specific testing to assure operability. Calibration and maintenance
, of monitoring equipment is performed on a routine basis. -
The particulate filters are tested after installation using a i diocty1phthalate (DOP) smoke test or equivalent. During operation, they 1 vill be periodically tested by laboratory analyses of inlet and outlet 1
*:millipore filter samples. ~
, . i
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;p' 3-
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_ - . . - _= . - - - . - - -- . - - -. - - _ _ . - ..
, .. i 39 6 '
Experience with boiling water reactors has shewn that the calibration of the offgas and effluent monitors changes with isotopic content. Isotopic content can change depending on the presence or absence of fuel clam 4 ng leaks in the reactor and the nature of the leaks. Because of this, the monitors are calibrated against grab camples periodically and at arq time ' there appears to be a significant change in offgas release rate. I i
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40. V. (b) Miscellaneous Gaseous Effluents h9 The previous section describes in detail the Augnented Offgas System '\] which is designed to process radioactive gaseous effluents prior to their release to the envirorcent. This section describes additional miscellaneous sources of potential low level radicactive airborne contaminants in the station which could be released to the environment. These are:
- 1. Release from the turbine gland seal holdup system.
- 2. Venting of the primary contain:: ant following periods of primary coolant leakage within,the containment.
- 3. Station ventilation exhausts.
- 4. Vents from liquid waste storage lanks, aerated resin regeneration tanks and open equipment and floor drain sumps, 5 Vents from radiochem hoods.
- 6. Periodic testing of the HPCIS.
- 7. Station startups utilicing mechanical vacuum pump operation. .
(b) 1. Turbine Gland Seal Holdup System Iow level gaseous radioactivity is released from the str. tion main _ p~y - . stack as the result of the operation of the turbine gland seal steam m f.V fsystem. This section describes the design and operation of the gland seal holdup system whose function is to reduce offsite radiation exposures 1' as the result of the operation of this system. .
- 1. Design Basis , ,]
...a The gland seal holdup system is designed to: :
' .:4
- a. Limit the discharge of radioactive gases from the main turbine gland sealing system and the mechanical vacuum pump (startup). H
- -6
- b. Provide processing for and discharge of the main turbine gland sealing system and mechanical vacuum pump effluent;s such that operation and availability of the station are not limited. .,
- 2. System Function
. The gland seal holdup' system collects and processes (by delay) the non-condensible exhaust from the main turbine glarid seal ;
condenser. During startup operation the discharge'of the condenser , mechanical vacuum pump is routed through the' gland acal holdup j ' system. After processing, the effluent of the gland seal holdup 1 system is routed to the main station stack where it is continuously [) Nj' monitored by the main stack radiation monitoring system (FSAR Section 7.12.3) before discharge to the environment. j ) 1 1 o g 3 4 _. 7
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r 41. pq 3 System Description v] T'h e gland seal holdup system shares with the Augmented Offgas Syste= the main stack, dilution fans and the stack radiation monitoring system. During normal operation the amount of radioactive activation and fission gases associated with the gland seal holdup system is extremely small. The radioactivity that is collected and
. processed by the gland seal holdup system is proportional to the amount of main steam utilised in the main turbine gland sealing system. This amount of steam is less than 0.1% of the full pcwer rated steam flow. In addition to the small amount of radioactivity processed, there is a correspondingly small amount of radiolytic hydrogen and oxygen which are well below the explosive limits.
The gland seal offgas subsystem (FSAR Section 9.4.4.3) is designed to provide a 175 minute holdup delay time for the radioactive gases before discharge to the main stack. This design is consistent with maintaining discharges within allowable limits due to the extremely small a=ount of radioactivity , associated
. with this system. -
During startup operations the condenser mechanical vacuum pump is used to assist the steam jet air ejectors in achieving condenser vacuum. The discharge of the mechanical vacuum pump is processed h' prior to releas,e by the gland seal holdup system. The holdup normally provided by the gland seal holdup system is reduced during startup due to higher air through-put when the mechanical .
- vacuum pump is operating, however, because the radioactive gases n' in the main condenser are only a s=all fraction of the design -M evolution rate during startup, the effect on radioactive effluents j released to the environment is negligible.
3
- 4. Equipment Description 1.4 The gland seal system provides a 175 minute holdup prior to release of gland seal non-condensibles to the main stack. The -
holdup line is a 16 inch diameter underground pipe extending from the turbine building to the main station stack.
.l.
5 System Operation During normal operation of k;he gland seal holdup system, a 2200 lb/hr. ' saturated air-water vapor mixture containing trace amounts of hydrogen, oxygen, and radioactive gases is exhausted from the turbine generator gland seal condenser and enters the . 16 inch diameter holdup line. After being delayed for a period of , approximately 1.75 minutcc the cffluent is routed to the main
/N stack where it is mixed with the Augmented Offgas System effluent J
( ) v and the discharge of the stack dilution fans before release to the environ =ent. , . .;b 4
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v) 6. Instrumentation 1+2. The gland seal holdup system is a passive system operating at atmospheric pressure requiring no particular control or instrumentation. Monitoring of the gland seal holdup system effluent is provided by the stack radiation monitoring system (FSAR Section 7.12 3) which monitors the combined effluents of the gland seal holdup syster and Augmented Offgas System. 7 _ Safety Evaluation The amount of radioactivity associated with the turbine sealing system is negligible. The extremely low levels of radioactivity released from the gland seal holdup system make direct radiation monitoring impractical, therefore, the total stack effluent is continuously monitored by the stack radiation monitoring system. Excessive release of radioactivity from this system is not considered credible due to its passive design and the small amount of main steam utilized in the sealing process. l
- 8. n spection and Testing I_
The gland seal holdup system is continuously operated during station operation and does not require specific testing to insure
%- operability. , ,
(b) 2. Primary system leakage inside the primary containment could
" occur as a result of recirculation pump seal leakage, valve flange leakage and valve stem packing leakage. The latter two are also sources of 1 potential leakage outside the primary containment. The magnitudes of M these leaks are minimized to the extent possible by regular periodic inspections and station maintenance procedures. "
An analysis was perfor=ed to estimate the site boundary exposures , resulting from primary containment purging assuming a five gpm unidentified steam leak for a period of time sufficient to reach equilibrium concentrations for all isotopes except Kr-85. Hoble gas activity concentrations in steam equivalent to a 25,000 microcurie /second offgas - rate after 30 minute decaywereused. The resultant whole body exposure
. per purge is esti=ated to be less than .001 mr assuming that purging commences after the reactor has been brought to hot standby. During station operation, drywell atmosphere vill be sampled for activity level prior to purging to assure that releases from this source vill be minimal.
(b) 3. The site boundary exposures resulting from steam leakage outside
, the primary containment have been estimated based upon releases equivalent to a continuous steem leak equivalent to 7 gpm of saturated liquid, from ;
the station ventilation exhaust. This has been selected based on experience VN in operating plants. The release to the environment may occur from the D 'd ) turbine building roof vant or the reactor building exhaust vent. Upper estimates of the magnitude of the whole body exposure resulcing from nobic gas releases frem the steam range from 0.1 to 0.1+ mr/yr. ,
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"' Assuming a leak rate of 7 gpm and a coolant concentration consistent with an offgas release rate of 25,c00 =ierocuries/second as measured at 30 minute decay and a condensation plateout factor of a results in an environmental release rate of 0.04 microcuries/second of I-131, with corresponding releases of I-132 to I-135 The value of 0.04 microcuries/
second release rate for I-131 can be ecmpared to measurements which have been made on operating BiR's which have shown release rates from the building ventilation systems of 2 X 10-3 microcurf.es/second to 4 X 10-2 , microcuries/second. The rate of release predictel results in a site boundary exposure rate of 0.6 millirem / year. (b) 4. Vents frcm liquid waste storage tanks, aerat'ed resin regeneration tanks and open equipment and floor drain sumps provide very little potential contamination. The radioactive noble gases are not present in solution except to the extent of their solubility in ambient temperature water. Particulate activity in the air space above stored liquid radwaste solutions is related to the gas liquid partition coefficient at the air water interface. The magnitude of this coefficient coupled with the filtration of a majority of the vents through high efficiency particulate
] (EPA) filters minimizes the potential for particulate releases from liquid v waste storage tanks and open su=ps. The collection of ionic halogens on station demineralizer resins creates an additional potential source of noble gases through the decay of halogens to their nobic ga;; daughters.
Ecwever,these gases are rot released promptly to the environment. The operation of the liquid radwaste system minimizes the release of gaseous daughter fis:: ion products by allowing the system components to act as gaseous delay tanks to effect the decay of the significant noble gas daughters. Only the occasional necessary air scrubbing and air sparging of certain radwaste tankage end normal tankage filling provides potential release mechanisms. The site boundary dose contribution from these sources is expected to be negligible. (b) ,5 Radiochemical hood vents provide a potential miscellaneous cource of release of airborno activity from the station. However, the sampling frequencies and volumes result in releases which are small fractions of the releases from other miscellaneous sources from the station. Further, the EPA filters installed in the exhaust ducting frem the radiochem hoods act to assure that no particulate radioactivity is released. (b) 6. The site boundary exposure due to testing of the high pressure coolant injection system (HMIS) for an assumed thirty hours per year has been ' evaluated. The HMIS turbine uses primary system steam for motive force of which 500 pounds per hour is used as HMIS turbine gland sealing steam and is condensed in the HPCIS gland seal condenser. The associated non-condensibles including trace amounts of noble gases are released during test operation to the environment through the reactor building exhaust vent or the standby gas treatment system which discharges to the main stack. The resultant site boundary whole body exposure is negligible, less than approximately 1% of that expected due to primary system leaks outside the primary containment. ,
. . 4 i N
1 44. wm di (b) 7 The magnitude of the sources and resultant site boundary exposures reulting from station startups utilizing mechanical vacuum
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pump operation are difficult to quantify because the number, nature, and duration of the preceding shutdowns are difficult to estimate. An order of magnitude estimate of the annual average exposure from ten startups per year was performed, assuming four hours of mechanical vacuum pump operation per startup, which indicated that maximum whole body exposures are less than approximately 0.05 mr/ year. These estimated
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1+5. VI. CONTROL OF LIQUID EFFLUENTS @O (a) Augmented Chemical Radwaste System (1) Desian Objective The augmented chemical radwaste system is designed to: (a) Collect radioactive and potentially radioactive liquid wastes of I high conductivity. (b) Provide processing of the chemical radwastes such that operation and availability.of the station are not limited. . (c) Maintain safe operating conditions and system integrity throughout all expected operating conditions. (d) Minimize radioactive and chemical releases to the environment to levels which are as low as practicable. !
- 2. System Bahis .
The augmented chemical radweste system is designed to collect 1 floor drainage, laboratory drains and chemical regenerant wastes from the condensate demineraliser system. The system will process, store for decay, and prepare liquid radwastes for discharge or for solidifi- ,' cation and off-site disposal. Radioactive liquid wastes, if released to the environment will be ' sampled and analyzed, then released on a controlled basis into the circulatir.3 water discharge canal.
~~
If released, the wastes will be continually monitored for activity during discharge. Solidified liquid ' wastes will be shipped in accordance with applicable regulations to an AgC licensed disposal site.
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- 3. System Description Chemical radwestes are liquid wastes generally high in conduc- '
- tivity with a varying amount of suspended solids.
F Floor drain vastes are collected in the sumps listed:
- 1. Drywell Floor Drain Sump
'2 Reactor Building Floor Drain Sump
- 3. Turbine Building Floor Drain Sump -
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- 4. Radweste Building Floor Drain Sump i lia, !
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The sump wastes are primarily minor equipment leakages, tank overflows, equipment drains, and floor drainage. When a sump has 6" filled to a, preset liquid level, the wastes are automatically pumped l j to the chemical waste receiver tank. Floor drain sump wastes may ' also be processed through the clean radweste system if the* wastes are ' relatively low in conductivity. _. ~ Condensate demineraliser regeneration' wastes and laboratory vastes are routed directly to the chemical waste receiver tank. The regeneration wastes consist of sulfuric acid and sodium hydroxide in
. dilute solution and are used to restore the ion exchange capacity of i the cation and anion resin in the domineralizers. The regenerant wastes are combined in chemical waste receivers to form sodium sulfate.
An ultrasonic resin cleaner (URC) will be used to remove suspended solids which collect on the resin. The URC cleans condensate deminera- ' liner resin beads by passing the beads through an ultrasonic energy ; field. No chemicals are used in the process. This minimizes the chemical regeneration frequency of the condensate domineralizers. After a chemical waste tank has been filled, the wastes will be pumped through the chemical waste filters, then processed through the radwaste concentrator. The concentrated wastes will be ptasped to the monitor ' tanks for storage and decay. The concentrator distillate will be ' pumped to the treated water holdup tanks and then returned to the condensate storage tank for recycling within the station.
" Depending on the activity level, the concentrated wastes after : '
-h' storage and decay may be pumped by the redweste metering pumps to the
- circulating water discharge canal or solidified with cement in the .~s o
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" The chemic'al waste receiver and annitor tanks (see Figure VI-1) 75 '
are atmospheric tanks with a capacity of 15,000 gallons and 20,000 'y gallons'each, respectively. The tanks are designed and constructed _, in accordance with API 650 and are internally coated with a phenotic 9 ' lining for corrosion protection. The receiver tanks have level and pH indicators and annuncistors which will be used in monitoring the " waste inventory. The monitor tanks also have level indicators and , annunciators and are insulated and heated. The tank heaters are the . electric immersion type and maintain the concentrated waste at 110*F. The heaters have been designed to allow quick replacement if necessary. + Both the chemical waste receiver and monitor tanks are located in shielded cells to maintain safe operating conditions and minimize radiation exposure to station personnel. ;
;9 \
, The chemical waste filters are pressure filters designed to section III, Class C of the ASMg Code. They re cartridge type s .)
filters rated at 10 microns. The cartridges can be interchanged to j a lower or higher micron rating as operating experience dictates. d The filters are located in a shielded cell from which cartridges may ' ,l
' be changed from an overhead platform by remote manual operation. 1' The eartridges will be replaced when a high pressure drop is f g experienced across the filter. spent filter cartridges will be placed 'i d into shielded containers,. then shipped cffsite for disposal. Two j filters and two shield casks will be provided to maintain maximum j
[ }, system availability. ,
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e 48. hV . ' The two chemical waste process and two monitor tank pumps are horizontal centrifugal pumps each rated at 100 spa. The pumps have 1007, capacity each and will provide maximum systam reliability. ; The radwaste concentrator (see Figure VI-2) is designed and con-structed in accordance with Section VIII of the ASME Code. It is capable of continuously processing 15 spm. The concentrator is , designed to achieve feed to distillate activity decontamination factors of approximately 10 5 to 106 . The concentrator is a spray film vapor ' compression concentrator that recycles the latent heat of vaporization of the distillate. The main components of the con-centrator are the concentrator shell and tube bundle, feed preheaters, vent condenser, concentrate cooler, makeup boiler, compressor, two distillate and two recirculation pumps. The makeup boiler is used for startup and for making up heat losses from the concentrator system. The co.ncentrator has redundant pumps for distillate and recirculation and a warehouse spara com-pressor. These components assure maximum system availability. .; After storage and decay of the concentrated wastes, and if the l
- activity levels permit, the concentrated wastes can be pumped to L'
- the circulating water discharge canal through the positive displace- _ ; ;
g ment radweste metering pumps. ' Two of the four pumps installed are rated at 0.1 to 0.5 spm while the remaining two are rated at 0.5 ; gr
- to 2.0 syn. . ._
..A radwatte solidification station is included in the design of the Augmented Chemical Radweste System to combine concentrated wastes 1 and cemet t into a homogeneous, solid product suitable for transpor- T{i .:
1, . tation to an AgC licensed disposal site. The major components of j the solidification station are the waste feed tank, waste feed pisap, Tf '
- t cement mixer, cement storage tank and cement feeder. The solidif t- O cation facility design has been based upon proven equipment to ,,
obtain maximum facility availability. See Figure VI-3 for a detailed i piping and instrumentation diagram of the radweste solidification _ ) system. See Figure VI-4 for the location of the system. , e, )
- j . The ultrasonic resin cleaner (URC) is designed to remove suspended I a , ,
solids from condensate domineraliser resins without requiring chemical l l regeneration. The major components of the URC are the cleaning column, , 1 , , flow adjustment panel, and control panel. Resin enters the cleaning ! column at a rate of 2 cubic feet per minute and falls through an ultra-sonic field.where the so11ds are removed. (See Figure VI-5 for the ;
- i See Figure VI-3 for a detailed piping and instru-
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./ A counter current flow' of water removes the solids and resin fines [
and transfers them to a holding tank. The waste water containing the i solids is then pumped to the clean radweste system. The cleaned resin j is then transferred back to the condensate demineralizer system for , 1 reuse. - ! i i j
- 4. $sfety Evaluation i e '
i The activity level of the chemical wastes processed through the ! chemical waste system is generally low. Releases to the environment l l j will be kept to a minimum by use of the ultrasonic resin cleaner to l reduce regeneration waste volumes. Solidification of the concentrated ( j chemical wastes for offsite disposal can be stilized as required. ! Backup equipment has been installed wherever practicable to , achieve maximum syste'm availability. Piping has been designed to allow greatest system ficxibility and system recycle capability. The chemical waste system is operated on a batch basis allowing station
; personnel maximum system control. i i 5. System operation '
I ,
. i (a) Normal Operation j -
j j ,During operation it is expected that the daily flow from ht e ficor i drain sumps will be 5,000 gallons. The drywell floor sump wastes will 1 h normally be transferred to the cican redweste system.
! The primary source of high dissolved solids waste results from the ! regeneration of the condensate dominera11:ers. The process flow dia-
- j. grams,' Figure VI-6, shows the expected volumes and activity levels through the chemical waste system during plant operation for an expected
- j. reactor water activity concentration which corresponds to a noble gas release ratu of 25,000 microcuries/second. Figure VI-6 is based upon a regeneration frequency of the condensate domineraliser system of one I i
vessel every five days. At this regeneration frequency a domineraliser l l will remain in service for a period of 30 days. During the 30 day run, the domineraliser will periodically have accumulated suspended solids
- removed to reduce the pressure drop across the domineralizer and use ,
- more efficiently the ion exchange espacity of the domineralizer resin. l j .. The resin cleaning will be done by the ultrasonic resin cleaner (11RC).
- The llRC will greatly reduce the volume of regenerant wastes and total '
! - curies of radioactivity to be processed through the chemical radweste
', system by lengthening the interval between regenerations allowing '
i longer accumulation periods on the domineralisers and subsequent saturation of a greater proportion of isotopes, thereby, minimizing the total curies processed. Figure VI-6 presents process flow para-
~ msters in the augmented chemical radwaste system assuming that all
> chemical radweste is solidified for offsite disposal. This mode of i
' operation was chosen for presentation in Figure VI 6 to display the i
maximum msount of information on process parameters. No intent to limit the augmented ehemical radweste system to this mode of opera-lj\ tion is implied. )[ c-
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Table VI-1 gives estimates of the expected annual curie releases QV , of radioactive isotopes frem the chemical radwaste system for retcases in excess 6f one microcurie per year at fuel conditions corresponding to an offgas release rate of 25,000 microcuries/second. The radio-nuclides F-18, Mn-56, Ni-65, Zn-69m, Na-24, W-187, Cr-51. Br-83, Br-84, Br-85, Tc-99m, Tc-101, I-132, 1-133, 1-134, I-135, S r-91, Sr-92, Zr-95, Zr-97, Nb-95, Ru-103, Te-129m, Te-132, Cs-136, Cs-138, Ba-139 Ba-141, Ba-142, Cc-141, Ce-143, Cc-144, Pr-143, Nd-147, Np-239, Cm-242, Ba-143, La-143, La-142, Cs-141, La-141, Cs-140, La-140, Cs-139, Cs-135, Sr-95, Y-95, Nb-95m, Rb-94, Sr-94, Y-94, Sr-93, Y-93, Y-92, Rb-91, Y-91m, Rb-89, Y-89m, Rb-88, Rb-87, Rb-93, Zr-93, Nb-93m, Pu-241, Rb-92, Am-241, Np-237, U-237, Pu-238, U-234, Pu-239, U-235. Th-231, Pa-231, Pm-147, Sm-147, Pr-144, Nd-144, Te-129,1-129, Rh-106, Rh-103m, Tc-99, Nb-97m, - Nb-97, Ag-110, Re-187, ?n-69 were also considered but are omitted from the listing in Table VI-1. These radionuclides may be present, but if present they will be of negligibic radiological significance in com-parison with the isotopes listed. The isotopes Ba-140 and Mo-99 are listed in response to specific questions raised. They are also of negligible radiological significance in comparison with the other iso-tapes listed in Table VI-1. Figure VI-7 shows activity levels assuming a 90 day run on the condensate demineralizers. This long run is considered to be a limit-ing case for operation of the condensate demineralizers without chemical regeneration. Figure VI-7 assumen augmented chemical radwaste system activity levels for a reactor water activity concentration which Q corresponds to a noble gas release rate of 100,000 microcuries/second. As such, Figure VI-7 represents a design basis set of parameters rather than a normally expected level of system performance. When a chemical waste receiver tank has been filled, the wastes will be neutralized with sodium hydroxide or sulfuric acid and then buffered with sodium bicarbonate to a pH of about 8.2 The wastes will then be treated with sodium sulfite for oxygen scavenging. (Neutralization and oxygen removal prevent acid and stress corrosion in the radwaste concentrator). At a pH of 8.2 ferric hydroxide and other metal hydroxides (both radioactive and non radioactive) may be precipitated in the receiver tank. When the Icvel of the precipitated wastes reaches a preset level they will be pumped directly to the solidification station and prepared for offsite shipment. It is expected that the precipitated waste will constitute a small fraction of the total liquid chemical waste processed. A floating suction has been provided in the receiver tanks to
< prevent carryover of the precipitated wastes. The neutralized and oxygen scavenged liquid wastes will be purtped by the chemical waste process pumps through a chemical filter to remove suspended solids.
I r'
,r TABLE VI-1 EXPECTED RADIOACTIVE LIQUID RELEASES FROM CIDtICAL RAIMASTE SYSTEM .
(At fuel conditions corresponding to an Offgas Releare Rate of 25,000 microcuries/second.) IS010PE CURIES / YEAR 1 Co-60 1.5 X 10 Co-58 1.6 X 10 1 - Mn-54 7 3 X 10-1 . Ag-110m . 9 9 X 10-1 , Fe-59 'T.9 X 10-2 0 Y-90 1 7 X 10 Zn-65 1.6 X 10-2 0 Sr-90 1.7 X 10 Cs-134 8 9 X 10-1 5 Co-137 1.8 X 100 0 Sr-89 2.4 X 10 Y-91 2.4 X 10-1 I-131 1 5 X 10-3 Ru-106 1.2 X 10-2 P-32 4.3 X 10-5 Ba-140 2 7 X 10-3 Mo-99 1.3 X 10-14 In addition to the activity releases listed in thin Table, a total of 12 curien per year of tritium in expected to be rolcased in the 1f quid ' radvaste system effluent. c0 - e -
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58. After filtration the wastes will be transferred to the radwaste concentrator. The feed to the concentr.ator is first passed through the feed preheaters and then sprayed over the concentrator tube bundle where vaporization takes place. The vapor is co= pressed and recycled to the inside of the tube bundle where it condenses. The condensed distillate is then pumped through the feed preheaters and into the treated water holdup tanks. The concentrated wastes drop from the tube bundle and are pumped through a concentrate cooler into the monitor tanks. The wastes are concentrated to approximately 25 percent by weight sodium
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sulfate. The concentrated wastes will be heldup in the monitor tanks for decay for the longest practicable period. If it is determined by sampling and analysis that the concentratc'd wastes have an activity concentration within the permissible release range, the wastes can be pumped by the radwaste =ctering pumps at a controlled rate through the radwaste discharge header and into the circulating water discharge canal. The waste will be continuously monitored for activity as it
. passes into the discharge header.
The concentrated wastes can also be pumped by the monitor tank pumps to the radwaste solidification system. The system mixes (on a batch basis) separate feeds of concentrated wastes with dry cement in a continuous mixer. The mixture can be placed into a disposable 'TN container where solidification of the cement takes place. The solid
$M waste can then be shipped offsite to an AEC licensed disposal site.
(b) Malfunction and Fcilure JSde Analysis During periods of main condenser sea water in-leakage, the volumes and decign basis activity levels in the chemical radwaste system will be as shown in Figure VI-8. This process flow diagram is based upon a regeneration frequency of two vessels per day. At this regeneration rate a demineralizer will remain in service for three days and the demineralizer resina will be backwashed instead of being ultrasonically cleaned. The backwash water will be transferred to the clean radwaste system. It is expected that significant sea water in-leakage into the condenser will be an infrequent occurrence. Figure VI-8 is included in order to chow a limiting mode of system operation at a sea coast site. This case is also the "cencentrator limited" mode of operation where regeneration frequency is keyed to the ability of the radwaste concentrator to process condensate demineralizer regeneration vastes. Sea water in-leakage'at higher rates would require station shutdown and repair of main condenser sea water in-leakage. Figure VI-8 assumes augmented chemical radaaste system activity 1cvels for a reactor water activity concentration which corresponds to a noble gas release rate - of 100,000 microcuries/second. As such, Figure VI-8 represents a design basis set of parameters rather than a normally expected icvel of system performance. The chemical waste system will function in the normal mode during
-- sea water in-leakage into the condenner, but with an increase in waste volume and a slight activity increase in the system.
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60. l 1 1 Table VI-1A provides an Equipment Halfunction and Failure Mode Analysis summary. The various components in the chemical radwaste system are instru-
; mented to indicate the status of vital process functions. Annunciators are provided on important system controls to alert station personnel to any abnormal process deviations and allow adequate time for corrective action. ,
- The following is a list of instruments on a system component basis:
1.- Chemical waste receiver tanks
- a. Level indicator .
- b. Level annunciation (high and low)
- c. Low-low and high-high level' switch (pump trip) .
- d. pH f ndication and annunciation (high-low)
- e. pH switch (high-low for pump trip) ^
- 2. Chemical waste process pumps ggs a. Discharge pres'sure indicator
- 6) Chemical waste filters
- 3. -
- a. Differential pressure indicator
- b. Differential pressure switch with annunciators (high-low) 4 Radwaste concentrator ,
- a. Feed flow indicating recorder .
- b. Feed temperature indicating recorder
- c. Concentrator pressure switch with annunciator (high-low)
- d. Concentrator pressure indicator .
- e. Concentrator temperature indicating recorder with annunciator (high-low)
- f. Concentrator foam detection annunciator . . -
~
- g. Compressor suction and discharge pressure indicator
- h. Distillate hotwell level indicator 3
- 4. Distillate pump discharge pressure indicator i
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' TABLE VI-1A E
- 5. (b) MALFUNCTION AND FAIIURE MODE ANALYSIS T DESIGN PRECAUTIONS MAIFUNCTION CORRECTIQ l l_ MUIIMEN'i' MAIFUNCTION CONSMUENCES l Iower tank level il. Chemical Waste a. Tank over filled Tank overflows Tank overfbw is piped to the E l- Receiver Tank radwnste building sump. Floor and process incoming
! drain sumps are tri med on vastes -into the I high-high level in tar.k. The other receiver tank.
[ level instrumentation will c > a alarm befora the tank over-f ' flows. L. I' v - Chemical addition
- - b. pH too' acidic or Ibssible radwaste pH indicator with annunciator concentrator chemical waste process pump as required.
{ -
+
basic corrosion trip on high or low pH,- vastes cannot be processed j until pH is acceptabic.
~ ,
l c. Tank corrosion Pin hole leak in Tanks are lined with a Incate leak and I due to possible tank-slight leaktge corrosion resistant lining repair tank. E = chemical attack. onto floor, spillage will be contained f, within the tank cell. I
- d. Ioss of air for Chemical waste not Tanks have mixing educator Station switches 2
l sparging homogeneously mixed. which can be used in place to second low E i pressure air blower. of the sparger y a E e. Floating suction No fluid will be The tank piping has been Drain tank and
- j. becomes fixed in pumped when level designed to enable the free the floating 1
of tink is below chemical waste process pump suction.
- }
I place, suction, to take suction from the j tank drain. The floating suction has a large moment , l - about its pivot point to keep l itself free. 6
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a s y . . j -- MUIIMENT MALFUNCTION CONSEQUENCES DESIGN PRECAUTIONS . MAIFUNCTION CORRECTION 1 .
- 2. Chemical waste a. Pump failure Cannot process wastes The pump piping has been Replace pump and use
. Process Pumps from receiver tank arranged so either pump 'other pump during may take auction from the interim.
~ '
either tank. t .l[ +
- b. . Pump cavitation Possible pump motor The pumps will'be tripped Replace motor and
'[ due to Inck of failure '
out of service on low-low use other pump during
; suction head. level of the tank. The the interim.
% operator will be alerted h , f , '
~ ~
before the pump trit s by the C}2 low level annunciators. 9 c.' Pump running at Possible pump mo' tor he pump motor has been Find cause of high pumping head h I; - shutoff head for failure designed with a 1.15 service a prolonged factor for overload and a requirements and D period. minimum recirculation flow . correct. Use other is, maintained by an orifice pump during the
.d. , \ l y; , f. , 4 loop around the pump. interim.
1.* zu ' 3.' Chemical waste a'. Exhausted filter. High pumping head. Se differential pressure Switch to spare (Nfilttra '
~ cartridge and flow through indicating switch will annunciate, filter unit and
- g? . . decreases. , replace exhausted
% cartridge.
pz t , W Radioactive 'a. Feed preheater Intermixing of feed High conductivity distillate Iocate and repair !!; waste internal leakage. and distillate if, out of the feed preheater leak. [tconcentrator sufficient feed is will be recycled back to the l .. leaking the system concentrator. Annunciator Q_ will trip, alerts operator. yn
\ b. Vent condenser Feed enters condensing The feed will drain back to Incate and repair -
internal leakage. section of condenser. distillate hotwell. The leak. lJn concentrator can still be
,pg operated with this malfunction.
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( a NA (w) .' l !l'MUIIMENT MALFUNCTION CONSMUENCES DESIGN PRECAUTIONS MAIEUNCTION COElRECTION
- 34. 1 Radioactive c. Concentrate Cooling water leaks into The pressure' of the cooling .Iocate and repair
} vaste cooler internal the concentrates. water is higher than that of leak.
j concentrator leakage. the concentrate and therefore i leakage will be into the l system. The concentrator can i be kept in operation during l' this malfunction. - 1 l d. Concentrator shell Distillate leaks.into The pressure of the distillate locate and repair tube bundle leakage. the concentrator shell is higher than the shell and leak. [. side. leakage will be into the shell.
~
The concentrator will operate
- t. with distillate leakage with
{' , only a slight decrease in heat efficiency. , 'ji '
- e. Compressor failure. System trip. The compressor is a centrifugal Replace compressor
., type designed for this service. with spare.
C A varehouse spare compressor
, has been supplied to reduce
' . g', system unavailability. Flanged rather than welded corrections have been employed, jf f. Distillate pump failure.
Momentary distillate backlog in concentra-Two distillate pumps have been provided, Switch,to second distillate pump and
!. tor. System trip. restart operation.
l, 3_ g. Recirculation pump Momentary concentrates Two recirculation pumps have Switch to second failure. backlog in concentrator, been provided. recirculation pu'up
, System trip. and restart operation.
- h. Concentrator high System trip. The concentrator has a high Reduce level in level level switch which will concentrator by pump-j; ,
shutdown the system. ing out concentrates y and restarting 1-system. ja &. . x Q
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' MUIIMENT MAIEUNCTION CONSEQUENCES DESIGN IEECAUTIONS . MALFUNCTION CORRECTION
- 4. Radioactive i. High or bw System trip. The concentrator has two Reduce or increase concentrator pressure' switches for high heat load input into pressure and low pressure. theccor.ruttrator tube bundle.
'j. High or low pH System trip. A pH meter is installed in Add chemicals as the recirculation line. requi.ed.
- 5. M)nitor tanks a. Tank overfilled T a k overflows. The tank 13 vel instrumentation Reduce level in vill alarm before the tank has tank by discharging
. , reached the overflow point. wastes to circulating water discharge canal or by processing wastes through the solidification system.
' b. Temperature of co Possible solidification The tanks are insulated to Process wastes to concentrated wastes of waste in the tank. reduce heat loss and are heated solidification
, low heater failure. by readi.ly replaceable immersion system or transfer
- ~ ' heaters. vastes to one of the two other tanks.
- c. Tank level low due to Possible burn out of Level instrumentation has been Add makeup water pump out. the heaters, insta. Lied to prevent the waste to keep tank level level from going below the above the heaters. s heater level. The pumps will
, trip on low-low level, a 6; Monitor tank
- a. Pump failure. Cannot process wastes The pump piping has been Replace pump and use
, - pumps from the monitor tank. arranged so either pump can other pump during the take suction from any of three interim. , monitor tanks.
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l RUIINENT MAIEUNCTION CONSMUENCES DESIGN PRECAUTIONS MALFUNCTION CORRECTIch c I i
; 6. 2 nitor tank b. Pump cavitation due Possible pump motor The p mps will be tripped on Replace motor and use j pumps, to lack of suction failure.
low-low level on the tank. The other pump during the g head, operator win be alerted before interim. e q
- the pump trips by the low level g annunciator.
- c. Pump running at shut- Possible pump motor The pumps will be operated with Correct cause of high off head for a prolonged failure. a recirculation flow back to the pump back pressure.
period. monitor tank. I'7.Radwaste' . a.'High level in the Backup of liquid to : The ' tank is provided with level Shutoff monitor tank
; ; solidification waste feed tank. the monitor tank indicat on and alarm. An pump and process vastes 6
; Station. pumps. operator will be observing the through the solidifica-I waste tank fill operation thereby tion str. tion to lower !
"' providing an additional safeguard. level in the tank.
O > b. Cement'or waste Cannot solidify '
'Jhe system win not operate if Repair cement or waste j; ;2 , feeder failure. wastes. either the cercent or vaste feeder as required.
feeder 'is inoperable. [' 1 ,
'Y c. Cement mixer driver No intermixing or The mixer has a remote manual Repair mixer driver.
J .g failure. flow of waste and operator which can be used to s cement. empty the mixer contents. The , mixer driver can be readily
- removed for repair.
f. , d. Cement mixer, Wastes remain in feed The colidification trystem Return vastes to if cement feeder or tank causing high piping is designed to allow monitor tanks and proceed with repair h . waste feeder radiation to prevent return of the wastes to the
, failure. repair operations. . monitor tanks. operation. ;
,?, -
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- t 66.
I Distillate conductivity indicating recorder with annunciater g f j. (high) 1
~
- k. Distillate temperature out of feed preheater and annunciator (high) l Concentrates density indicator 1.
i I m. Recirculation pump discharge pressure
\ -
- n. Concentrates flow indicating recorder
- o. Concentrates outlet temperature indicator with annunciator (high-low)
! 5. Monitor tanks -
- a. Level indicator
- b. Level annunciater (high-low)
- c. Temperature annunciator (low-low) 6 Monitor tank pumps ,
- a. . Discharge pressure indicator C.
g _. l-
- 7. Radwaste solidification station
[' Cement tank level with annunciator (high-low)
- a. ___
'b. Waste feed tank ' level with annunciator (high) . .
- c. Waste feed flow indicator --?
. .x
- d. Disposable container level indicator with annunciator (high)
- 6. Inspection and Testing The chemical radwaste system is continuously operated during station
- operation thereby demonstrating system operability without special in-spections or testing. Routine analyses for activity levcis will be done t .. on the concentrated wastes. Periodic testing will be done on influent
~ wastes in the chemical vaste receiver tank. Area radiation monitoring
~
~ ~ -
'_ will also be done' routinely. . .
A _. ~
- 1. - - . 1:
*f _
~
,~ > _ _._;
.. f
- l. p , .
i', g. -
* * * ; <. . f . 'k
- 5 7, [4
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,g -
; , 3,l . . >m , . gy
, m x..ex : as xe ~~ :
.r~ ~.. ~
. , . ; - ; ~;;;; g , ,;;7
,.3 y 3. p.c~3 y .
.- qq __
67 Ox VI. CONTROL OF LIQUII) EFFLUEIES ( (b) Miscellaneous Wastes ! 1. Design Basis l l i l The miscellaneous waste system is designed to: I (a) . Collect low level radioactive liquid wastes that have potential'ly high detergent levels. (b) Provide processin6 of the miscellaneous wastes such that operation , , and availability of the station are not limited. l (c) Minimize radioactive liquid effluent releases to levels which are l as low as practicable.
- 2. System Function -
. The miscellaneous waste system collects equipment washdown and decontamination solution vastes, radiochemistry laboratory detergent solution wastes, and personnel decontamination vastes, The miscella-neous waste system processes and' filters these liquid wastes before
, discharge through the radwaste discharge header into the circulating CD"-
""*"" *** """"* *"* '"* **"**"""***""*""">**""*""*27=** -
before release and continually monitored during release. ~ ~ I 3. System Description _ When one half of the miscellaneous waste drain tank has filled, ' f the liquid wastes are sampled and analyzed for activity and filtered " '. before release. If necessary, miscellaneous wastes of high radio- e activity concentration and low detergent levels may be transferred to
~
r 9 the chemical waste receiver tank for further processing. I 4 Equipment I)escriution '
. The miscellaneous waste drain tank is an atmospheric tank with a capacity of 1,000 gallons. The tank is designed and fabricated in accordance with API 650. The tank is carbon steel and divided into two sections each with a capacity of 500 gallons. The tank is located within a shielded cell.. The miscellaneous waste filter is a pressure type cartridge filter designed to Section VIII of the ASME Code. The filter cartridges are rated at 10 microns, however, they may be inter-changed with a lower or higher micron rated cartridge.
s t ( [ . s , lD ~ [ a g . p t. 3=r," [ ,
/ _z;
' ,, -fG
, . y . .g L 7 - ?!x we , . .r~~ r = - - .. m m .u m w
I , . i, : I I 68. e \ ' h 5 system Overation (a) Normal. Operation i j During normal operation it is expected that the monthly volune
]
of miscellaneous wastes will be approximately 1000 gallons. The j* ' sources of miscellaneous wastes will be kept to a mini =um through j . administrative control. When one section of the misocllaneous waste tank is filled, the wastes are sampled and analyzed for radioactivity. The expected gross radgoactivityconcentrationinthemiscellaneouswastedraintankis 10" microcuries/co. Table VI-2 gives estimates of the expected annual curie release of radioactive isotopes from the miscellaneous waste sys-i tem at fuel conditions corresponding to an offgas release rate of 25,000 j microcuries/second. The estimate is based upon a 5 gpm flow of miscel-laneous waste into the circulating water discharge canal for a period of.200 minutes every 30 days. Thetotglexpectedannualreleasesare estimated to be approximately 4 0 X 10 microcuries per year. The vastes are pumped through a filter and discharged at a controlled rate through the liquid radwaste discharge header into the circulating water discharge canal. The miscellaneous vaste is continuously monitored
. for activity as it passes through the radcaste discharge header.,
~ ~ ~ ~
, (b) Malfbnction and Failure Mode Analysis When higher than normal activity levels are experienced in.the . ,
miscellaneous wastes system, the wastes can be processed through the chemical radwaste system or sent directly to the radwaste solidifi-cation system for packaging and eventual offsite disposal. The occur-rence of high activity is not expected. .
~
- 6. Instrumentation The following is a list of miscellaneous waste system instrumentation:
l
. Miscellaneous Waste Drain Tank
- a. ~ Level indicators. ,
- b. Levelswitches(highandlow,pumptriponlow-low).
- c. Levelannunciators(highandlow).
- 2. Miscellaneous Waste Drain Tank Pump.
- a. Discharge pressure indicator. .
- b. Flow indicator.
f
- 3. Miscellaneous Waste Filter L a., Differential pressure switch with annunciator.
- f y p -
g e- - w ...&. .-
, ;m d.-
4 9 t
~
69 TABLE VI-2 EXPECTED RADIOACTIVE LIQUID RELEASES FROM THE MISCELIANEoUS RAIMASTE SYSTDI (CURIES / TEAR) (At. fuel conditions corresponding to an offgas release rate of 25,000 microcuries/second. MISCELIAtEoUS RAIMASTE ISOTOPE ' SYSTD1 Co-60 2 9 x 10-4 co-58 2.2 x 10-3 - Mn-54 2.2 X 10-5 _
, Ag-110m 3.3 X 10-5 Fe-59 / 3.0 x 10-5 .
'Y-90 3.1 X 10#5 g .
' Zn-65 5 5 x 10-7 Sr - 3.1 X 10-5 _ , ,
cs-134 2.0 x 10-5
. a I
Cs-137 3.1 X 10-5 - 'd Sr-89 2 7 x 10-4 Y-91 . 2.0 x 10-5 I-131 8.1 x 10 4
-Ru-106 3.2 X 10-7 P 32 2.8 x lo-6 Ba-140 2.3'X 10-4
~
No-99 1 7 x 10-6 , 1 J Ru-103 _1l.5 X lo-6 Approximate , Total 4.0 X 10-3 C p-- h g [ f - -~l r s. '
- v. -
y , l ' e',
< * , s s
* * ' ' b +
,3 3 e, _; .
~
7 . - .. w .
. q? ...- ..
.q,.
. ~ . u ,. j, 'j); /.:? . ,'
l G,)i) ',~ ^ ~ _ 4l , &, b
< ~i
, . ~
s m 70. , i
- 7. safety evaluation l The activit concentration of the miscellaneous wastes (10-6 to 10-4 microcuries o)islow. Releases to the environment can be kept to a minimum by administrative control of vaste generation. The mis-cellaneous waste system is operated on a batch basis allowing station personnel maximum system control.
f
- 8.
- Inspection and Testing The miscellaneous waste system is operated during normal station operation thereb/' de=onstrating system operability without special in-spections or ' testing. Routine analysis for activity levels will be done on the miscellaneous wastes prior to release.
O
. I
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- 49 ;. w 3
h . =:. , s&w - - . 1 --m-=-.__ _ :. We's_T7'r%v. - - - --e m n:?v
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'N
/ 71.
U VII. ENVIRONMENTAL EFFECTS OF RADIOACTIVE EFFLUENTS (a) Site Boundary and Large Population Group Exposures from Caseous Effluents
- 1. Exposures Resulting from Effluent from the Augmented Offgas ystem l ,
The design basis for the Pilgrim Nuclear Power Station Augmented Offgas System is a noble gas release rate of 100,000 microcuries/second after a 30 minute holdup. It is noted that the present Technical Specifications ibnit the offgas release rate after 30 minute holdup to 100,000 microcuries/second averaged over the preceding three calendar months. This restriction has the effect of limiting the annual average offgas release rate (after 30-minute holdup time) to levels on the order 1 of 50,000 microcuries/second. For the purpose of estimating the expected site boundary and large population group exposure rate, a long-term average release rate of 25,000 microcuries/second af ter 30-minute holdup tiac is uaed in this section as the input source term _ to the Augmented Of fgas System. Table VII-1 lists the noble gas ac-tivity concentrations leaving the existing 30 minute holdup and the noble gas activity concentration leaving the charcoal absorbers in the ' l Augmented Offgas System. , f.. Previous calculations have indicated that. the NNE and the SW sec-tors represent the highest dose sectors due to meteorological conditions, h) site boundary distances and terrain height. 3 b For completeness, the site boundary gamma dose rates for both sectors were calculated. However, the site boundary in the NNE sector : -v is the shoreline. No people live there, so no one will receive the "C} indicated dose. The dose rates were weighted by the frequency of . ;-; occurrence of wind speed, direction, and atmospheric stability. The .fb meteorological data were collected from May 1968 through April 1969 '?j , from a 220 foot instrumented tower at Plymouth. qg I A tabulation of these doses is included in Table VII-2. In all fi cases, dose is calculated for a fixed point (air dose) with no con- "'! sideration of human occupancy or shielding. However, such reduction '
~% l factors are real and should be considered in any estbnate of "true" g' effect. The SW sector dose is calculated for an effectiva stack 1' height of 61 meters at a site boundary distance of .53 kilometers.
The shorter stack height was used to account for the elevated terrain , height at the site boundary in this sector. The NNE Sector dose is calculated for an effective stack height of 122 meters at a shoreline _ distance of .25 kilometers.
- n The NNE sector dose is slightly higher, however, the difference between the sectors is not appreciable. For 12.3 SCFM condenser air 3 in-leakage, the site boundary dose in the SW sector rate is 0.023 ;j area / year. Normal background radiation at the Pilgrim Nuclear Power v s Station is on the order of 100 mrem per year. ;
. .,?
i f; * ,
~
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49
'l * . .) { 0 my 4
e.f' y ' .
~ ., ,.
< ; ; yiy
^~
7s_,, y. h- - - m--- - m -
i s l
! N 72.
1 I 1
-TABLE VII-1 l l
EXPECTED ANNUAL AVERAGE NJBLE GAS ACTIVITY CONCErffRATIONS f l (Microcuries/Second)
. Discharge From Existing 30 Minute Discharge Trem i - Isotope Holdup Line Charcoal Adsorbers i
2 -2 Kr-83m 7 2 X 10 1.3 X 10 Kr-85m 1.3 X 10 1.4 X 10 Kr-85 2-3 2-5
-4 Kr-87 3.8 X 103 4.2 X 10 3
k Kr-88 4.5 X 10 3 0 X 10 1 0 l Kr-89 4.5 X 10 l 0 1,
.'Xe-132m ,3.8 X 10 L
-2 -
,e
' ~
.Xe-133m 7 0 X 10 1 6.5 X 10
" 3 1 -~
Xe-133' - 2.0 X 10
-J..:
3 1,1 x 1o 2
- 7 _ 7 _ Xe-135m _
1 7 X 10 . . - . . 3 Xe-135 5 5 X 10 0 ..(
- 4 2
Xe-137 1 7 X 10 0 ,. l -Xe-138 5 2 X 10 3 o s Approximate Approximate -.- l Total 25,000 Total 135 - i f 9 L m5 i -
. a l ,.
.. l; .
# * ..en
*f,
,6 n
',t!
- .er- e
.'a,
.(
r
- s J
- , ,. 5 , .a .. g ,s ~
T
< " h ~. ,
- ~j^
+ t , ;. -<.
.,(
~
5
.s l
- 1
,' J ,- % .. ' sg e.' J }, - , . . . , , .[*' '
2 ,
;>-; g: ;.g:p- . , ,;; ,
y - . .
^
Q- a , , ,, . ? ,is '.
- 3 ,
~;w ,;c ,n y ;; p - : . , , , ., - lr . i . ;. .ya .4 C'q;: y;-
-7:-
i ,. ; . , Q: .
- . s. . , a , . ,
---,.c-- . . w _.- m..-.-
s I ~ [ .s s - 73. TABLE VII-2 AVERAGE ANNUAL WHOLE BCDY CIOUD GAMM DOSE AT SITE BOUNDARY AT 25,000 MICR0 CURIES /SECOND ANNUAL AVERAGE NOBLE GAS ACTIVITY
~
INPUT RATE TO AUGENTED OFFGAS SYSTB4 Parameter NNE Sector _ SET Sector
. j 1.
Distance to Site , Boundary (km) . 0.25 0 53 l
- 2. Effective Release Hei6ht(m) 122 61 -
3 cloud Gamma Dose 1 (er/ year)* for ; G release rate of 135micrecuries/second 0.025 0.023
~ ~ '
.-4_. )
i
. - ~ ,
' ,;- ?.,i
^'
- Dose rates are to a detector with no credit allesed for shielding or occupancy. Typic.s1 human exposures vould be 1/3 to 1/10 of these values.
Inhalation. dose resul. ting from particulate release from the Augmented Offgas System is expected to be negligible as the result of the long -
-delay times provided and the multiple filtration steps.
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,. - _ , . _ . . . _ _ _ - . _ _ .. ._._,_____ [ _~ . __mg_,__,
i i
= *.
i , I f p l l 74. Based upon the expected site boundary exposure rate of 0.023 mrcm/ year, the exposure to the large population group (4 million people living within 50 miles of Pilgrim Nuclear Power Station) is 0.48 manrem/ year as the result of the effluent from the Augmented Offgas System. 4 I I
.g n .
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75 I 2 Exposures Resulting from Miscellaneous Gaseous Effluents [ (1) Turbine Gland Seal System and Gland Seal Holdup System Table VII-3 gives the expected annual average release rate from operation of this system. The glan,d seal steam system effluent is j routed directly to the stack after a holdup of 1.75 minutes. Previous calculations have indicated that the NNE and the SW sec-tors represent the highest dose sectors due~ to meteorological conditions,
, site boundary distances and terrain height.
i i For completeness, the site boundary gamma dose rates for both l sectors were calculated. However, the site boundary in the NNE sector } is the shoreline.' No people live there, so no one will receive the i indicated dose. The dose rates were weighted by the frequency of j occurrence of wind speed, direction, and atmospheric stability. The i meteorological data were collected from May 1968 through April 1969 _ l from a 220 foot instrumented tower at Plymouth. - i < f A tabulation of these doses is included in Table VII-4 In all cases, dose is calculated for a fixed point (air dose) with no con-
- sideration of human occupancy or shielding. However such reduction ._ !
factors are real and should be considered in any estimate of "true" , 4 effect. The SW sector dose is calculated for an effective stack f D,6 height of 61 meters at a site boundary in this sector. The NNE d j Sector dose is calculated for an effective stack height of 122 meters _. t at,a shoreline distance of .25 kilometers. - 33
'1 hie dose is ' highest in the NNE sector, 0.23 mrem per year. The '
7tj NNE sector site boundary is the shoreline. No pecple live at this up,
- shoreline, so no one will receive the indicated dose. In the SW
;fg:
j sector, the dose is 0.15 mrem per year. These dose rates make no .: , , allowance for shielding factors, occupancy factors, and other factors 1l l which will further reduce the dose rate. . 1 1 . 1: Miscellaneous sources (2) through (7) (See Section V(b)) , The total annual estimated exposure due to release of noble gas ! activity from these miscellaneous sources within the station is ex-pected to be not greater than 0.5 mrem / year. In addition to the noble l , 7 gas exposures, calculations have been performed to estimate annual ' average exposures resulting from the release of I-131 from the station - ventilation building exhausts. These calculations result in annual . i exposures due to release of I-131 of 0.6 mrem / year. j _ y i -
^<
}_ < ;
'?g
- J j
]
.'a i - -
1 i -
.s. ~j
, 4 i y 9 i- s $ -
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, ; 9, -
-; .w; ,9pn> - ;7 ;}; _ :'
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.- +
z:<.y J qq;;. ';+= g-j:f - f ;Q Gy'
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. - - - . . .. - .. -- ._.-'"LL ' s 2,: lL L :. T :D- L L - 2 .z - --.
~~
i 76. i s TABLE VII-3 ' A, \ . i EXPECTED ANNUAL AVERAGE NOBLE GAS ACTIVITY CONCENTRATIONS 4' , _.
] -
(Microcuries/second) 1 (At fuel conditions corresponding to an offgas i release rate of 25,000 microcuries/second . i after 30-minute holdup) 4 1 j Discharge from Gland Seal i . . . Isotope , Holdup System E-83M 8.85 x 10-1 m-85M 1 54 x 10 m-85 6.02 x 10-3 m-87 4.83 x 100 - 1 0 m-88 4 93 x 10
. E-89 2.19 x lol .
" o 7 78 x lo 2-90 m-91' 5.42 x 10-2 4.92 x 10-3 ;
XE-131M ..
. XE-133M 7 36 x 10-2 .:
a . 0
] 'XE-133 2.07 x 10 6.10 x 100
~
XE-135M D' XE-135 5.4'5 x 100 l XE-137 2.83 x lo 't l XE-238 2.09 x lo
~XE-139 1.18 x lol !
f . . XE-140 3 75 x 10-1 -
, ~
-Approximate ,
Total 117 i [. ,. { ~ , I f .- . ( - y-
- a ; .
1 ,
' 4
- _,~ =
g c,,
,'{*,
* # i <- +, ., d 4, p , ,, , , , _
# . 4 s
*'h'.,
,,3 ,. _
' ' '%. ,',; a4
~, - '- .,y w
,y'": .
.<n '. n ,
[', ,, yr * * ".
."t [ [ ' .. '
f Y,' t, e j9 s i' ,
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- W.
m TABLE VII-4 q AVERAGE ANNUAL WHOLE BODY CLOUD GAMMA DOSE AT S*.TE BOUNDARY AT l 25,000 MICROCURIES/SECOND ANNUAL AVERAGE NOBLE GAS ACTIVITY INPUT TO GLAND SEAL HOLDUP SYSTEM l . \ .
~
l NNE SW l Sector Sector Distance to Site Boundary, Km .25 .53 Effective Release Height, m 122 61 l s i Cloud Ga::sna Dose (mr/ year)'* for Release Rate of 117 Microcuries/Second 0.23 0.15
~
i
*The whole body doses are calculated for detector locations. No ~
~
consideration is given to occupancy or shielding factors. Typical reduction factors in the range of 3 to 10 should be applied to obtain more appropriate doses to an individual. -', Inhalation dose resulting from particulate release from the gland seal holdup system in the SW sector has been estimated to be 0.07 mrem / year. r 6 c
- 3. -
2 r , ,f g 8% ,
, + +
,' b; jj 3 , ,-
> 1 ; .p . .
;z e ., . f tj+
. a ,
i , 78.
~
I 1
'~"N VII. (b) liquid Effluent Erposures to Individuals through the Aquatic
/ Food Chain Exposure rates to man resulting from the release of radioactive material in liquid effluents after processing through the Augmented l Chemical Radwaste System have been estimated. The analysis assumes no solidification and packaging of liquid wastes for offsite disposal.
This analysis supersedes a similar analysis presented in FSAR Amendment 20. These estimates considered the most significant exposure pathways to man. Gastro intestinal tract exposure resulting from consumption of seafood caught in the vicinity of the Pil6rim Nuclear Power Station is the most significant contributor to exposure to man. This exposure is estimated to occur at a rate equal to 0.36% of the 10 CHt 20 limit. For completeness, the exposure resulting from sub=ersion in the station j' discharge canal was estimated an.1 is reported in this section. 1 - Table VII-5 su=marizes the exposure rates in millirem / year to the f -'
" average individual" as defined in the note to Table VII-5. These exposure-rates assume a conservative dilution factor of 50 for fish, molluscs and crustacea. Exposure rates resulting from the consumption of carraEeenan, an extract frem the marine algae Chondrus Crispus or " Irish Moss", are also presented. Exposure rates as' a percentaEe of thelimits in 10 CFR 20 are also presented in Table VII-5 fd.
WV Analysis Methods and Assu=ptions - g , [ In this analysis, the following modes by which man may be exposed
~
J to the released radionuclides have been considered- - ~' w n Direct external exposure from the radionuclides in the liquid.
-- : 1. 3 effluent. This could happen to people engaged in extensive .
recreational activity in or on the water. The greatest _
~y e exposure would result from submersion in the water as in the
^
I case of swimming. The dose would depend on the time of [ _ . submersion, the amount of radionuclides present and the type , '
- .of radiation emitted.
E .- 2.' Internal exposure as a result of eating seafood. The '
~
' radionuclides in the water may concentrate in ,various marine
, organisms present in Cape Cod Bay. The importance of this
" mode of exposure would be determined by the specific y radionuclides and organisms involved and the quantity consumed.
The results of the' calculations are shown in Table VII-5. The
. radionuclides considered in this analysis are listed in~ column 1 of Table '
1, 1 VII-6. Their importance is based on the relation of MPow to discharge concentration and reconcentration in marine organisms. [ -1 W , ) '* a
+ - - >
b P 4 E- : . l %- i f- ,.
~
l , n
~~'
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, . rg f e ,
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,(' ,'
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._ _ . . . _ _ . . . . ~ . _ . _ . . _ _
- 8 I
79. ! a,\ l 1 TABLE VII-5 l '
- EXP0'SURE RATES TO THE " AVERAGE INDIVIDUAL"(1) l RESULTING FROM THE RELEASE OF RADIOACTIVE -
MATERIAL IN LIQUID EFFLUENTS (MREM / YEAR) l
. (At fuel conditions corresponding to an offgas release rate of 25,000 microcuries/second after 30-minute holdup)
Total Gastro-Intestinal Body Tract
~
, Consu=ption of Seafood (% of 10 CFR 20) 0.67 (0.134%) 5.4 (0 36%)
Consu=ption of Carrageenan (% of 10 CFR 20) 0.004 (0.00027%) 8 Submersion 7" '
~0.0019(0.00038%) 0.016 (0.0036%) ,
' ~
.- D ..t
.. i:
Ti .4 A ~ (1) The " average individual" (a) Eats 50' grams o5' seafood caught in the vicinity of the Pilgrim Nuclear Power Station each day. (b) Intakes 2 grams / day of carrageenan in the form of 1 quart of _ fchocolate milk and 1 pint of ice cream
-' _ (c) Submergesin.thedischargecanalfor100 hours / year
-j
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,; ll[ r Q hl"' l, ,Gl*
Y, .f! ' * ~4; _ ?Ys . ? O m? ~ f "? ,: . ncem-~a, - n m m mm mzerm m ~w w wm-n-- m na& M A wm
I ( j, 80. j l O The maxi =um permissible concentration for each radionuclide in I water (MKw) for soluble materials which are in general the limiting . b case, is listed in column 2 of Tabic VII-6. . These values are from Table h II, Appendix 3 of 10 CFR 20. The MECw was calculated on the basis of the limiting marine species used as a food, i.e. , fish, crustacea or 3 molluscs as shown in Appendix A to this section of Amend =ent 31. . fe . An additioniL calculation of the (IECw)a was perforced for algae as the limiting organism. While of no direct food value, Irish Moss, a j marine algae, is used as a source of an extract called carr;tgeenan. 9 Carrageenan is used to take stabilicing or thickening agents for products 3 such as chocolate milk and ice cream. The methods and assu=ptions used { for dose ce.lculation are su=rarized in Appendix A and B. The ratios of the expected concentrations to the IE w, the MPCCw and (MFCCw)a are related ! to dose, but not directly additive. This is because the MPCw values relate j to different critical organs in man. 4 Based on these calculations the most significant exposure is to the l - gastro 1.ntestinal (GI) tract through the consumption of seafood. The l limiting radionuclide for this mode of exposure is Co-60. Other significant j isotopes are Co-58 and Mn-5+ 1 which tcgether contribute about 70% of the Co-60 dose. The exposure to the four million people living within 50 miles of 4 e. Pilgrim Nuclear Power Station resulting from the consumption of seafood has been estimated to be approximately I rem per year. This estimate is based upon the total body dose presented in Table VII-5 and statistics on lobster and flounder landed in the Plymouth area in 1970. e b I a 4 I 7q $ ]v) g., - ,
- ,' 1
~ '
\ ' * :_c- .i. ,
~ '
e .,
. 4
81. TABLE VII-6 LIQUID EFFLUEUT UUCLIDE DISCHARGES AND DOSE REIATED PARAMETERS CANAL CONCENTRATION MICw CANAL ISOTOPE (MICROCURIE / MILLILITER) (MICROCURIE / MILLILITER) CONCENTRATION /I% Co 60 2.29 E-OS* 5.0 E-05 4.57 E-04 Co 58 2 55 E-08 1.,0 E-04 2.55 E-04 Mn 54 1.14 E-09 1.0 E-04 1.14 E-05 ' Ag HO" 1 55 E-09 3.0 E-05 5.16 E-05. Fe 59 1.24 E-10 6.0 E-05 2.08 E-06 Y 90 2 71 E-09 2.0 E-05 1.36 E 1 zn 65 2 51 E-11 1.0 E-04 2.51 E-07.
, Sr 90 2 71 E-09 -
3.0 E-07 9 03 E-03 Cs 134 1.39 E-09 9 0 E-06 - .1.55 E-04
~h -
Cs 137l 2.85 E-09 2.0 E-05 1.43 E-04
. v.
S: 89 3.80 E-09 - 3.0 E-06 1.27 E-03 i m;* Y 91 3 72 E-10 .3.0 E-05 1.24 E-05
^
I 131 . 2.33 E-12 3 0'E-07 7 78 E-06. ; ;;
. au 106 1.81 E- H 1.0 E-05 1.81 E,,06 F 32 6.78 E-14 2.0 E-05 3 39 E-09__
Ba'140 4.22 E-12 3.0 E-05 1.41 E-7 Mo 99 57 E-13 2.0 E-04 2. 8 E-9 s _ :s -
~
t f , o s i s_ A1
^
- . ' f* E - n = X 10-3 * -
t I
=s '
v v v-9 * , 4 ,
*O .-
- A* 4 , , ,J . t
. g
...,;'. . , , ,J,-- ' * = ;, ,c ..-,.] j
..,s.
=-
l k .'[
' '% yr -' .g 4 fp* ,
,f #
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3 4- ? ,, e . 7 , /6. .$ e .h
.,.,>w. y:: tg i
l i '. , - , y . - l.c,. . - V, _ .n,
' u.,.
[ j< , ;, , +. t.>.
. . cnw tr d ,i ~ . :8+. { . .n,u ' :.c,.;( , ,, '., . . .; ;- L .
. s,.,,
M.l,
*]; _, ? W y {f q , t 1*-Tj il - k,+;
l1 0- , y , [. ; *.- y .', p 7 7
- f,' , J ,
4.' 7 , ,
. ,-c
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.. e . ; . y ' ntg f, q f . , ,g ,
. , . - y. . ..i . , 9 7t y'd .,. 4
._ _ . _ . . . . _ . . ~ . _ - . _ . _ . _ . - _ . _ . . _ _ . . _ ~ . . _ _ .___ _. _-_
I-m TABLE VII Continued 82. A 1 CONCENTRATION
'. p t FACTOR (C.F.)
FISH, MOLLUSCS, CONCENTRATION FACTOR (C.F. ) MICC (fhe) MFCC ' P ISOTOPE OR CRUSTACEA (a) (AIGAE) (MICROCURIE / MILLILITER) (MICROCURIE /MILLILITEF , Co 60 10,000 c* 100 2.2 E-07 2.2 E-05 - Co 58 10,000 c 100 4.4 E-07 4.4 E-05 U Mn 54 50,000 10,000 8.8 E-08 m 4.4 E-07 Ag 110" 5,000 1,000 cm 2.6 E-07 '1.3 E-06 Fe 59 20,000 , 6,000 1 3 E-07 4.4 E-o7 Y 90 100 30o 8.8 E-06 3.0 E-06 za 65 50,000 , 1,000 8.8 E-08, 4.4 E-06 Sr 90 1 20 1.3 E-05 fc , 6.6 E-07.
~
Cs 134 50 10 ' 7 9 E-06
~ .y
- c. 4.0 E-05 ,
i Cs 137 50 c 10 y. e - ,., 1.8 E-05 8.8 E-05
- g. ~ '
.- Sr 89 ~ 20~
-1.3 E-04 6.6 E-06_
.s
-. . .-- - .1. fem . . ' ~
.., . = .
n($ Y 91 . looc , 300 , 1.3 E-05 ~4.5 E-06
^
- ~.. ~ . ~ . .
I 131 looy
- .~ ..
'10,000 ~ ~1.3 E-07 1 3 E-09 .; .. .
- Ru 106
- n;T, ..-1,000 100.. 'm c _ . . . . _ _ . 4.4 E-06 4.NE-07
^- , _ ~
. m. ,
P 32 10,000 fe, 100,000 8.8 E-08
' 8.8 E-09_ ,
Ba 140 J3 -100 - 4.4 E-04 fem - 1.3 E-05 .
-Mo 99 100 c,
- 100 ~8.8 E-05 8.8 E-05 l,. - _ _ . . . .
\. . - . . l:. _
- Subscripts f, c, m, a are used to denote fish, crustacea, molluses and algae, respective:
. , ,, 4 t
l .-a>
.s
.' ~
e
, s ..p 4 .s + . .
. e j
- 1. .
4 [ . .. s * .a* i .. s . i t g ,. fm. '* ,
- 4
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, a s a. ,a l . .
n ! + [\ .. x
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s .-
, a l -
, , = -
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i
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.w
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4 e n s ~ s s s _ s s
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;"W:)} .
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. .' , a - . . y c.- -oa; .
' f,.,-Jf.I[ ' f - u_,.,.
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a m :) L wg _'r_ 't' % . ;; q.. , t y (L ,..Q,y M..y c % u.;9 @ ? ,2 ,y.QlM y ; '. y; i,. " " W w q. ;;5'. x .a
, , a r,
) 83. t i TABLE VII Continued ISOT0FE , CANAL CONCEh"IRATIONAiICC CANAL CONCENTRATIONATCC
. (fme) (a)
Co 60 1.0 E-01 1.0 E-03 Co 58 5.8 E-02 5.8 E-04 , Mn 54 1.3 E-02 2.6 E-03 m 1.2 E-03 Ag im . 6.0 E-03 Fe 59 9 5 E-04 . 2.8 E-04 Y 90 3 1 E-04 9 0 E-04 Zn 65 2.8 E-04 5 7 E-06 Sr 90 2.1 E-04 4.1 E-03 Cs 134 1.8 E-04 7 7 E-06 Cs 137 1.6 E-04 * <' - 3.2 E-05 Sr 89 2 9 E-05 5.8 E-04 - h Y 91 2 9 E-05 8.2 E-05 J. I 131 1.8 E-05 1.8 E-03 . Ru 106 4.1 E-06 4.1 E-05 " . P 32 . 7 7 E-07 7 7 E-06 '4 Ba 140 9 5 E-09 _ 3 2 E-7 . l; No 99 6. 5 E-09 _
- 6. 5 E-09 ;.
't I
- l. - .
S &1 t
.f
. a.
Y "
~ '
. - a
# * " - -e
[' ,
,7 .y 9Ib ' ' #
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? c 1'
;fi j y jr, -,vs,'- . T .,
4
./ +
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,. , , x
- 7. , ,
, , s , , . s , 'r . #p e' , .
, . , . .,,I .
i ( .. . .,,+n . , c. :.~ . a . , .- , . n ~. v
l', *. l 84. A APPENDIX A The MECw is defined by the relation: MPCOs = 2E00 X IHCw 50 X C.F. where 2200 = the average quantity of water an ad it drinks each day in - cubic centimeters MICw = the maximum permissible concentration of the radionuclide in water, which would give the limiting exposure to the critical organ 50 = the average quantity of seafood consumed per day, in ' grams, Reference (1) - C.F. = the concentration factor or the ratio of the concentration of-the radionuclide in an aquatic organism to the concentration l of the radionuclide in water (microcuries/ gram per microcurie / ! cubic centimeter), Reference (2). The dose to the total body and the GI tract are then calculated using i equations (1) and (2) of this Appendix with the MICCv replacing the MICw.
=The largest concentration factor in the categories of fish, molluscs or ~^
l y### crustacea was used for each isotope. f;;
~ '
Dose calculations through the marine seafood chain are similar to , ' ~ those used for calculations of dose due to intake of drinking water. The a calculations for drinking water are based on the ratio of the concentration ' . of the radionuclide in water to the MFCw. The MPCw is dependent on the _m organ of reference, i. e. , total body, GI tract, thyroid, etc. In most , cases, the 10 CFR 20 MPCw is one-tenth of the ICRP MFCw for the critigal ~ ?) organ, i.e.', the limiting case of exposure. ICRP values are used in the calculations in the following manner: n Equation (1) Dose C
=5000mremX{
i=1 tb n Equation'(2) Dose C
= 15000 yr mrem X {
G.I. i,1 ( W G. I.
- A similar calediation was performed using the concentration factor for algae. An extract from Irish Moss called carrageenan is used in a
' number of products. It was assu=ed that the ~ extract contains the same
- ' distribution and concentration of radionuclides as the algae.
, ' The(MPCCw)a 0 is then , q
~
e
~(MFCCw)a = 2200 X MICw
- -2 X C.F.
where 2 = the average daily consumption in grams of carrageenan in -
. . chocolate milk and ice cream.
.. q 7
- The effective holdup period lfor' decay.of Irish moss prior to con ., ' '
[, 4
!D
#9 r eumption of _ the. carrageenan was conservatively assumed to be 30 days. ~ ' ' .
.a 4 _
.. q a m, y 3 ,y
t 85
, __ APPDIDIX B
, ,y
, ) The submersion dose calculation conservatively estinates the dose
] to a swin=er by assuming it is the sa=e as the dose to the water at a
} point below the surface. The point at which the dose is calculated is assumed to be several beta particle ranges and several gar:ma-ray mean-free-paths from the surface or bottom of the body of water. Under these conditions, electronic equilibria vill exist and the energy absorbed per unit cass of water is equal to the energy emitted per unit mass of water. This further assures that the radionuclide is uniformly distributed in the water body.
The beta dose is then calculated as: Dos =K n m Ci XE p Xf f 13 g3 Where K = 1.87 X 107 mrad / year _ (uC1) (Mev) ml dis Ci = the concentration of the i tn isotope ij = the j th average energy of each beta c=itted by the ith isotope (h f ij = the fractional yield of the j th energy beta from the ith isotopt Mev. - Similarly, the ga==a dose is . y. n m Y T Dose =K Ci X E Xf ; I where the symbols have the same meaning as applied to garra radiation.
; 4 I The beta dose calculated in this fashion is conservative in that the I range of the beta particles' beyond the skin is limited. The garna dose should be closely approximated by this method of calculation. ,
REFEREICES l
- 1. Weaver, C. L. , "A Proposed Radioactivity Concentration Guide for l Shellfish". Radiological Health Data and Reports, p. 491, September 1967
. 2. Freke, A. M. , "A Model for the Approximate Calculation of Safe Rates of Discharge of Radioactive Wastes into Marine Environments".
m Health Physics, 13, 743-758 (1967). (
% ,/
i h 1'
, r 'I j y{ '
s j a
g s - .l , ,' ., 86. a N (') ., VIII. Proposed Changes to Technical Specifications and Means for Determining Compliance l The first paragraph of Specification 4.8.A of the Technical Specifi- ) cation (General) attached to the proposed operating license would be
&. amended as follows:
f y i' Operating procedures shall be developed and used, and equipment j which han been installed to maintain control over radioactive Q materials in gaseous and liquid effluents produced during normal ] reactor operations, including expected operational occurrences,
, shall be maintained and except for the radwaste solidification and packaging system used, to keep levels of radioactiva material in effluents released to unrestricted areas as, low as practicable."
4 The radwaste solidification and packaging system shall be used
- whenever it is determined that without its use the exposure to the average individual (defined as in Table VII-5 in Section VII of FSAR Amendment
- 31) would exceed 5 millirems to the whole body, gonads or bone marrow or 15 millirems to the bone or thyroid.*
- Exposure to the whole body would be assessed as exposure to the gonads or red bone marrow. (Inserts underlined.)
E D The bases to Specifications 3.8 A and 4.8.A (General) would be U j added to as follows: l The radwaste solidification and packagine system would be operated { on an "as needed" basis to the extent that transportation and licensed AEC disposal sites are available. e l 4 Compliance with the Technical Specifications will be determined by the degree to which the applicant has used his best efforta, through maintenance, repair and replacement, to assure that the equipment that has been installed for the purpose of reducing radioactive material in liquid and gaseous effluents is available for use. 4 f . A - 1 4 h y (p3) { v I h
's
# )'
_- ---.-- m s_a ,
, e}}