ML19220B336
| ML19220B336 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/03/1975 |
| From: | Tedesco R Office of Nuclear Reactor Regulation |
| To: | Moore V Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7904250647 | |
| Download: ML19220B336 (21) | |
Text
,
9 UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHIN G TON, c.
C.
20555 6
.~
CC7 3 1375 Docket Ncr. 50-320 V. A. Moore, Assistant Director for Light Uater Reactors, Group 2, RL SAFETY EVALUATICN RE?CRT I';?UT 70R THREE liILE ISLA'ID, U'i1T 2 Plan Name: Three :lile Island Nuclear Statica, Unit 2 Licensing Stage: OL Decket Number: 50-320 liilestene Number:
24-01 Responsible 3 ranch:
L'.7. 2-2
? oject Leader:
H. Silver Description of Raspcnse: Radioactive Waste Management Section for SER Recuested Ccepletica Date: Septe:ber 5, 1975 Enclosed is the radioactive vaste canagement section for use in the Safety Evaluation Report for Three !!ile Island, Unit 2.
Be applicant has preposed to un state-of-the-art technology for the liquid, gaseous and solid rad-taste treatment systems and the systers are designed to acceptable codes and standards. Ue find thasa systers
~
acceptable.
Our analysis of radioactive releases due to liquid tank f ailures cu: side shaus the provisions incorporated in the applicant's design contain:ent to citigate the ef fects of tank failures involving contaninated liquids to be acceptable.
The determina:ica whether the proposed liquid and gaseous radwaste systens the dose design objectives of Appendi:c I to 10 C7" Part 50 uill be nee:
evaluated in a supple en: to the SEn.
m,
t
' y..s,
< w,
s y-
_c Ecbert L. Tedesco, Assistent Direc:cr for Containnent Safety Division of Technical Review
.-)
Enclosure:
As stated
[ f' i,.
.d cc:
(See ne:<t pcge)
CtuTIC9
$ i.R $
. - t e r, d c.m e
9
)
4 9
s x
2 W+sf
s - a#
'/30425 0(,
'/b~108
e
_ V. A. :4oore cc:
R. Boyd
'J.
IcDonald S. Hanauer R. Heinenan H. Den:ca S. Varga D. Eisenhut D.
luller K. K:2iel H. Silver J. '.lorris J. Glynn J. Kastner J. Collins
'4. Burke 5.,
- u... act P. Stoddar:
109
<a
q 11.0 Radioactive Waste Manace en:
11.1 Su==arv Descrictica Three Mile Island, Unit 2, will have independent radioac:1ve syrtens designed to provide for centro 11ed waste =anagement handling and treatment of reactor coolant liquid and gasecus The miscellaneous liquid waste and solid waste systems, vastes.
~
The reac:or located at Unit 1, will be shared by Units 1 and 2.
liquid was:e chain (RCL;;C)'will process wastes f rem the coolan:
The miscellaneous waste chain CC:C) will shim bleed systen.
and flecr process radioactive liquid wastes frc= equipecn:
drains, decenta=1natica and laboratory wastes and laundry and The gasecus vaste sys:cm will provide holdup shower rastes.
lived noble gases stripped frc, the primary capccity to decay short of ventilatica exhaus:s through HEFA coolant and treatment filters and charecal adscrbers as necessary to reduce relcases of radioactive materials to "as leu as practicable" levels in acccrdance with 10 CF?,?at: 20 and 10 CFR Par: 50.34a.
The will provide for the solidifica: ion, packaging solid waste sys:c and s:crage of radioactiva vastes generated durin'g staticn offsite to a licensed facility for operation prior to shipment bu ri a.'..
In our evaluatica of the liquid and gaseous raduaste systems, vc have considered: 1) the cap.fsility of the cyste=s for keeping low as practicabic",
the levels of radioactivity in cffluents "as r -
$ g) u
_2-based en expected radwaste inputs over the life of :he plant,
- 2) the capability cf the systens to =aintain, releases below :he li=1ts in 10 CF? Part 20, Appendix 3, Table II, Colu=ns 1 and 2, during periods of fissica product leakage at design levels from the fuel, 3) the capability of the systems to =eet the processing demands of the statica duutag anticipated operaticnal occurrences,
- 4) the quality grcup and seismic design classifica:1cn applied to the syste= design, 5) the design features tha: wi1I b~e incorpo-rated to con:rol the releases of radioactive esteilars in ac ord-ance with General Ocsign Criterion 60 of.ippendix X to 10 CyR Part 50, and 6) the potential for gaseous release due to hydrogen explosiens in the gasecus radwaste systa=.
~ ~ ~ ~
in our evaluation of the solid radkaste treatment system, we have considered: 1) the systen design objectives in terns of expected typea, volumes and activities of waste processed for offsite shipment, 2) easte packaging and ccnformance to applicable Federal packaging regulaticas, and provisions for centrolling potentia 11;.
radicactive airborne dus:s during baling cperation, and 3) pro-visions for onsite 3:orage prior to shipping.
In cur evaluation of the process and effluent radiological monitoring and sampling systers, we have considered the system's capability: 1) to.cnitor all normal and pctential pathways for release of radicactive materials to the environment, 2) to
^
cont rol the release of radioactive naterials to the environ =ent,
-s.(
@ and and 3) to rend:cr the performance of process equipment detect radicactive caterial leakage between sys ces.
We have determined the quanti:ies of radioactive esterials that will be released in the liquid and gasecus efflucnts and the quantity of =aterial that will be shipped offsite as solid waste for burial during normal opera:icn. The principal radienuclides associater with liquid and gasecus ef fluents and solid wastes -
are listed in the draft Envirennental Sta:crent for Uni: 2hr-and In making these de:crdinaticas we censidered waste flcus, activities and equipment perfor ance consistent with expected ner:al plant cperation, including anticipated operaricnal eccur-that are censistent wi:h an assu=2d 20 years of normal
- rences, plant operation. The liquid and gasecus scurce terns were cal-culated using the ?'.:R-GALE ccda described in "2:af t Re gula to ry Guide 1.23, Calculatica cf Releases of Radicactive Materials in Liquid and Casecus Ef fluents fro: Pressurized ~.ater Reactors
(?WRs)", Doc'-M 'lo. E 0-2, dated February 20, 1974 The principal para:eters used in these calculations, along fith : heir bases, are given in Appendi:c 3 :: Draf: Regulatory Guide 1.22.
Based on cur evaluation, as described belcu, we find the proposed liquid, gaseous and solid radwaste systens and associated prccess and ef fluent radiological enitoring and sa pling syste=s to be acceptable.
(0**
.~
_4-The capability of the liquid and gaseous' radioactive waste treat-the dosa design objectives of Appendix I to
=ent systers to meet 10 CFR Part 50 will be evaluated in a supplement to :he SER.
11.2 Systen Descriotion and Evaluation 11.2..
Licuid Radioactive Wasta Treatment Systems of The liquid radicactive vaste treatment syste= will consist process equip ent and instrucentation necesaary to collect, process, =enitor and recycle or dispose of radioactive liquii - _ _
All potential radioactive liquids generated in the wastes.
plant are collected and processed through the reactor coolant liquid xaste chain (RCLUC) or the =iscellaneous waste chain (MAC) prior to recycle or discharge to the blowdown systen of the rechanical draf t cooling tower.
The liquid radioactive waste will be processed on a batch basis to perni: cptimum control o f releases. Prior to being released, sa ples will be analyzed to deter:ine the types _and amounts of radioac:ivity present. 3ased on the results of the analyses, the waste will be retained for further processing or released under controlled conditions to the blowdown sys:c of :he techanical draf cooling tower for dischtrge to the Susquehanna River.
Aerated drain wastes collected in the reactor aerated drain systen such as containment and auxiliary building drains, laboratory drains, and deminerali cr regeneration solutions, will be segre-gated based on their origin and activity and processed through W 113 Turbine building floor drain wastes from the potentially the MNC.
conta=inated makeup water treatment and condensate polishing Turbine systes areas will be directed to the MNC for treatment.
building liquid wastes from nor: ally nonradioactive areas will be collected separately from other turbine building wastes and will be discharged wi:hout treatment into the yard drainage systen unless routine sa pling indicates that the radioactivity 'evel in the turbine building surp exceeds a predeter=ined level, in which case the waste will be processed through the MNC.
The MNC will consist of holdup tanks, an evaporator, a de=ineralicer, a filter, and waste test tanks.
Detergent (lcundry and decontamina-tion) wastes will be collected in a holdup tank and sampled and analyced. The applicant has indicated that if the activity level
-b is less than 10 uC1/cc the waste will be discharged without treatrent through the plant sani:ary vasta system.
If the activit-,
level exceeds 10' uCi/cc the was c will be processed through the MNC.
The RClWC eill process a portion of the Chemical and Volume Control Systen (CVCS) flow (shin bleed), along with nonaerated drain wastes, such as equiptent drains, collected in the reactor plant drain syste=, for bor.ca renoval and cleanup prio to discharge to the environ =ent.
The ECLUC will consist of holdup tanks, an evaporator, denineraliccrs, filters and test tanks.
t t
r..
O 6-along The principal conpenents making up each of these sys:ces, are listed in Table 11.2-1.
with their principal design parameters, The design capacities of the RCLUC and MWC evaporators are Cur calculaticas indicata 22,000 gpd and 13,000 gpd, respectively.
the average expected waste flous to the RCLWC to be 1,300.gpd that 2.
The applicant and to the shared MUC to be 6,400 gpd frc= Unit proposes to utilice intercennections between the RCLUC and :.-lC The difference between to provide reserve capacity for th'e "WC, the expected f1cus a.nd design capacity and the provisions for interconnections between systens will provide adequate reserve We consider the systen capacity for p:ccessing surge ficws.
capacity and systen design to be adequate for =ecting the de= ands of the statica during anticipated cperational occurrences.
- eis ic Category I The liquid radwaste systens will be located in a The seis=ic and quality group designaticas of the ecuip-structure.
nent which are ec==ensurate with the guidelines of 3 ranch Tech.-ic2l Pcsitica ETS3 11-1 are listed in Table 11.2-1.
The syste= will also (continued on next page) 4 It)
@ be designed to control the release of ra'dioactive raterials due to overficws f rom tanks outside contain=ent by providing level instrumentation which will alare in the control rcce, and by means of curbs and drains which will collect liquid spillage and retain it for processing. We consider these provisions to be capable of preventing the uncontrolled release _q1. radioactive __
=aterials to the environment.
We have determined that during normal operatic" '-
---n-eed liquid radwas:c treatment systens will be capable of reducing the release of radioactive esterials in liquid effluen:s to approximately 0.26 Ci/ year, excluding tricia: and dissolved gases, and 350 Ci/ year for tritic=.
11.2.2 Radicactive Caseous icste Trea: ent 5 esters The radioactive gaseous vaste treatman: systems (RGWS) will be designed to process gaseous vastes based en the origin of the wastes in the plan: and the enpacted activity levels.
Ane RGUS will consist of the radvas:e gas syster (RGS) and :he buildin; ventilation exhaust systens that control the release of radioactive effluents to the environ = ant.
The principal componen:s of :he RGS, along with their design parameters, are listed in Table 11.2-1.
M S ?{ --
M I t3 a6.L U
O
-S-The RCS will mis t of a process gas sys:em, a process vent header system and a decay tank storage sys:e=.
The process gas systen will collect and process the hydrogenated gases stripped frca the pri=ary coolant letdown strea=, the reactor building vent header, and the vent gas syste=.
The gases will be ec pressed and stored in the decay tank storage system.
Operating with two 40 sef= cc= pressors (one in ccatinuous use 3
and the other as backup) and two 1938 ft, 150 psig, gas decay tanks, the process gas systen will have adequate capacity to allow operation during periods of equipcent downti=e.
of the gas entering the RGS during nor:al operations Most will be cover gas displaced frc2 the borca recycle holdup tanks as they fill with liquid. This gas will censist ydrogen. To prevent crygen build-
- -d b
primarily of nitroga-up in the syste=, the vent header will be designed to operate at a slightly posi:ive pressure.
In additica, gas coaitor samples will be continually and aunctaticall:. drawn fran the tanks dis-charging to the vent header and from the decay tanks being filled.
A nonitor alarm will be activ;ted if the hydro ~"r content of any sample exceeds three percent by 'rolunc.
In this =anner, the
~.tial for explosive hydrogen /o:cygen mintures will be =itigated.
We consider the syste= capacity and the sys:ca design to be
^
adequate for necting the decands of the station during nor=al t~ r -
3a.4 r, f n) g'
O
_ g_
operations and anticipated operational occurrences. The systen The seis=ic will be located in a seismic Category I structure.
which are and quality group designaticas of the equip ent, co==ensurate with :he guidelines of 3 ranch Technical Position ETS3-11-1, are listed in Table 11.1-1.
We find the system quality group, scismic design criteria, and the explosion design provisions to be acceptable.
The rain condenser vacuna pu p discharge will be monitored for noble gas radicacti-fity concentration and combined with the turbine gland seal condenser discharge and the auxiliary building ventilatico This ef fluent strees will normally be released untreated exhaust.
through the plan: vent.
In the event the radioactivity level exceeds a predetermined value the effluent strea will be processed o the through HE?A filters and charcoal adsorbers prior :o release environment. The syste: releases will depend en the ra:e of primarr to sacondary systen leaha~e and the primary ecolant activity.
In e
the even of excessive primary to secondary leakage, the affected genera:cr could be isolated befora radicactive catarial sten:
concentrations in cain condenser of f; s exceeds the limits in 10 C7R Part 20.
The discha'rge from the condenser vacuu: pungs during s:artup will be released untreated.
Ventilatica encausts frc the auxiliary and fuel handlin' ouil inza o
~
will acrnally be released untrea:cd; however, if the racioactivity r7d 44 t
8 O in the exhaust air is above a predeter=ined value or during activi:les such as fuel handling, it will be processed.through HEPA filters and charcoal adsorbers. Ventilation exhausts fro = poten:ially contaminated areas of the service building will be processed through HEPA filters prior to release. The contain=ent building atmosphere will be recirculated through filters and charcoal adsorbers prior to purging to the ventilatica exhaus t syste. The turbine buildin; ve~ntilation exhausts and pertions of the service building ventilatica exhaus:s will be released to the environment without treatment.
The plant ventilation systers will be designed to induce air flows frc potentially less radioactively contaminated areas to areas having a greater potential for radioactive con:a=ination.
We have deternined that the proposed gasecus radwaste treatment systets and plant ventilation syste: will be capable of reducing the release of radicactive materials in gascous effluents frc:
Unit 2 during normal operation including anticipated operational occurrences c approximately 23,000 C1/ year cf ncble gases and 0.13 Ci/ year of iodine-131.
11.2.3 Solid "2dicactive Masta Trentront Svs:ces The solid radioactive waste trea::ent syste= will be designed to and need collect and process wastes based on their ehysical for:
for solidificatica prior to packaging.
" Wet" solid wastes, con-sisting of spent denineralirer resins, evaporato
,::ces, filtcr 7G 119
O
. sludges, and chemical drain tank effluents, will be cc=bined with a solidificatica agent and catalys: to for= a solid =strix and sealed in 35 gallen steel dru=s or 50 cubic feet shipping containers.
Dry solid wastes, consisting of ventilatica air filters, conta=-
inated clothing and paper, and =1scellanecus items such as tools and glassware, viil be ec= pac:ed into 55-gallon steel drums.
Miscellaneous solid wastes, such as irradiated primary system ecmponents, will be handled on a case-by-case basis based on their size and activity. Enpected solid waste volu es and ac:ivities shipped offsite for Un?it 2 will be 600 drums /yr of " wet" solid waste centaining an average of 10 Ci/dru= and 450 drums /yr of
" dry" solid waste centaining less than 5 Ci total.
Shipping container filling operatiens will be centrolled re ctely frc= consoles located cutside the fill area. The presence of free water in the shipping containers.till be indicated by renote visual inspectica..' Additional solidificatica agents will be added to solidify free water, if present.
Saling of dry wastes vill be carried out in a shrouded encicsure and the air ficw will be directed through a HEPA filter and exhausted frc= the building vent.
The soliu radwaste sys cm will be located in the waste. disposal area of the Uni 1 aunaliary building. The seismic and cualit,-
^
group designaticas of the building end equipment are cc=mensurate e
v<f120
O s with the guidelines cf Branch Technical Positica EIS3 11-1 and are listed in Table 11.2-1.
Shielded storage f acilities (1,153 f t') for up to 200 drums of 1 Fuel solid radioactive wastes will be provided in the Unit 3aled wastes with icw radiation levels will Handling building.
be s:cred in the Uni: 1 Heat Exchanger Vault roof area (7,000 f t~).
Temporary storage areas for shipping containers will be provided here also if needed. Based cn cur estinate of e:cpected solid' waste volutes, we find the storage capacity adequate for nee:ing Wastes will the demands of the statica during normal op.eration.
20, be packaged in accordance with the requirenents of 10 CFR Part 10 CFR Part 71 and 40 CFR Parts 171-179, and shipped to a licensed burial site in accordance with NRC and 20T regulations.
11.3 Precess and Effluen: Radici:cical Moni:crine systens will be The process and effluenc radiclogical =cnitoring syste designed to provide information cencerning radioactivi:y 1cvels in systers throughout the plant, indicate radioactive leakage between systens, 2cnitor equip:ent performance and =cnitor and ccatrol radicat.:ivity levels in plant discharges to the environs.
Table 11.3-1 Liquid and gaseous strerns will be conitored.
Monitors
'adicates the proposed locations of continuous nonitors.
on effluent stre:ns will automatically terminate discharges.shcu_
5 L)*-ss~ 4
@ radistica levels exceed a predetermined value; these monitors are identified in Table 11.3-1.
Syste=s which are not a enable to centinuous =enitoring or for which detailed radioisotopic analyses are required will be period-ically sa pled and the sa ples analyzed. The sa=pling syste will provide representative primary and seccadary liquid and gasecus samples as required to eff ectively =caitor the operation of both units and to ccatrol radioactivit'j levels in plant discharges to the environs.
This systen will provide sa:ples fro = approximately 200 different primary and secondary systen points, which can be sampled locally sample sinks located in the auxiliary building and service or at building. Radiccetive sa:ples vill be either delayed to allow decay of short-lived radicisotopes and/cr shicided frc the sa ple sink to protect personnel handling samples.
If a critical sc pling line becenes incperable, there is at least ene alternate pcth which can be used to obtain a similar sample.
We have reviewed the locations and types of affluent and process cenitoring provided.
Based en the plant desi~n and en the con-tinuouc cenitoring locaticas and intermittent s =pling lccations, we have concluded th all normal and potential release pathways will be =cnitored.
We have also dorcr=ined that the sampling and 75-122
s conitoring provisions will be adequate for detecting radioactive
=aterial leakage to normally unconta inated systers and for
=enitoring plant processes which affect radioactivity released during normal operation, includin; anticipated operational On this basis we consider the =cnitoring and occurrences.
sa:pling provisions to meet the require:ents of General Design Criteria 13, 60 and 64 and the guidelines of Regulatory Guide 1.21.
11.4 Findines Our review of the radvaste systems included syste= capabilities to process the types and volumes of wastes expected during normal operations including anticipated operational occurrences in accordance with General Design Criterion 60, the design provisions incorporated in accordance with General Design Cr.'terf n 60 to preclude uncontrolled releases of radioactive raterial due to leakage or overflows, and the quality group classification and seistic design criteria in confor:ance sich the guidelines of 3ranc2 Technical Position ETS3 11-1.
We have reviewed the applicant's descriptiens, process flou diagrams, piping and instrunan-syste:
tation diagrams, and design criteria for the components of'the radwaste treatment systers and for those auxiliary supporting systems that are essential to the operation of the radwaste treat-ment systcas.
'Je bave perfor cd an independent calculation of the releases of radioactive raterials in liquid and gaseous effluents based on the calculational cetheds of Draft Regulatorv Guide 1.3'J.
rt --,
a-)
i n) 4.
Our review of the process and effluent radiological onitoring and sr=pling systens included the provisions proposed for sa=pling and :x,.it - -ing all statica ef fluents in accordance with General Design Criterion 64, for providing autcratic termination of effluent releases and assuring control over discharges in accordance with General Design Criterien 60 and Regulatory Guide 1.21, for sampling and ncnitoring plant vaste process streams for process con:rcl in accordance with General Design Criterion 63, and for conducting sanpling and analytical programs in accordance with the guidelines in Regula:ory Guide l'.21.
The review included piping and instrument diagrams and process flow diagrams for the liquid, and solid radwaste systers and ventilation systems, and
- gasecus, the locatica of =cnitoring points relative to effluent release points en the site ploc diagram.
Based cn the foregoing evaluation, we conclude that the above aspects of the radwaste treatment and process and effluent meniter-ing sys cts are acceptable. The basis for accep;ance has been cen-formance of the dpplicant 's designs, design criteria, and desi;n to bases for the *adicactive waste treatnent and monitoring sys:c as sell as the applicable rega11tions and guides referenced above, to staff technical positicua and industry standards.
f n)ma, l
@ The capability of the liquid and gaseous radioactive vaste treat-systems to meet the dose design objectives of Appendix I cent to 10 C7R Part 50 vill 'ce evaluated in a supple:ent to the SER.
f (d-..a Ad
13.7.3 Radioactize Scills Cutside Containment Postulated Radioactivity Released Due to Licuid Tank Failures The consequences of cc=penent failures which will result in conta=inated liquid releases to the environs were evaluated for cc=penents centaining liquid radioactive =aterials located outside reactor contain=ent. The scope of the review included the calculation of radionuclide inventories in station ec po-nents at design basis fission product levels, the citigating ef f ects of the plant design, and the ef f ect of site geology and hydrology.
The tank that: will centain the highest total quantity of activity is the reacto'r coolant bleed holdup tank.
This tank will have a volume of 75,000 gallcas and is assu=ed to be c0% full with a liquid activity ccacentratica of approxinately 46 uCi/n1 (based on 1% operating power fission product release to the pri=ary cdolant). The fics of ground water will ecve radionuclides in We evaluated the liquid the ditectica of the Susquehanna River.
time for radwaste leakage to the river to be approxinatel;.
transit 6 at the We estimate a dilution facter of 7.3 x 10 1.6 years.
nearest water supply intake, located five miles dounstream.
We calculated that a rupture of the reactor coolant bleed holdup tank will give a concentratica of 4.4 x 10-8
~
uCi/21 at the naarest water supply intak2.
This talue is a small fraction of the limits l
~
e
-2_
of 10 CFR Part 20, Appendix 3, Table II, Colunn 2 for unrestricted areas.
Based on the foregoing evaluatien, we conclude that the postulated failure should not result in radionuclide concentrations in excess of 10 CFR Part 20 li=its at the nearest potable water supply.
Therefore, it is not necessary for the applicant to incorporate additional provisions in his design to nitigate the ef fects of co=ponent f ailures involving cent -4"ated liquids.
( L) *-
k sw A.
@~
e Table 11.2-1 DESIGN PA?A'E ERS OF ?RI5C!?AL CC"?CNENTS CONSIDEEED IN LIQUID, CASECUS, A'O SCLID R.G.ASTE E'.'ALUATIONS C"C '"'"'" ' S Capacity Quality Radioac:ive Licuid Waste Svstem No.
Each Grouc Miscellaneous Waste Ealdup Tank 1*
19,000 gal C
Miscellaneous Waste Neutrali:er Tank 2**
7,500 gal C
2**
9,200 gal D
Waste Test Tank 2*
2,100 gal C
Laundry 'Jaste Drain Tank Miscellaneous Waste Evaporator 1**
12.5 gp=
C Evaporator Ccndensate De:ineralizer.
2**
20 gpa D
Radioactive Gaseous Waste Svstc=
Process Gas Cc: pressor (RGS) 2*
40 scf=
C
'Jaste Gas Decay Tank (RGS) 2*
'1938 ft3 C
HEPA Filter (RGS) 1*
100 cfa D
Charcoal Adscrber (RCS) 1*
2-in bed D
depth Radioactive Solid Waste Svsta=
Spent Resia Tank 2*
340 gal C
Reclained 3cric Acid Tank 1*
7500 gal C
Ccncentrated Waste Storage Tank 1**
7500 gal C
Eoren Eccover-> S esten /Eeactor Ccolant Liquid *:.'as t e Chain)
Reactor Ccolan: 31eed.loidup Taak 3*
75,0C0 gal C
Reactor Ccolant Evaporator 1*
15 gpm C
Evaporator Ccndensate Cecineralizer 2*
20 gpa D
- Uni: 2
- Shared 'with Unit 1 r : -..g,
$ t) asf