ML18026A370

Forwards Revised Response to NUREG-0737,Item II.B.2,to Address post-accident Sampling & Analysis Areas & to Verify That Doses from Containment Were Used in Analysis & Zones. Response Completes Action on SER Outstanding Issue 80

ML18026A370
Person / Time
Site:	Susquehanna
Issue date:	06/11/1981
From:	Curtis N PENNSYLVANIA POWER & LIGHT CO.
To:	Schwencer A Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM PLA-810, NUDOCS 8106150245
Download: ML18026A370 (38)

Text

REGULATORY INFOR'AVIATION DISTRIBUTION SYSTEvi (RIOS)

'CCESSION NBR:8106150245 OOC,DATE: 81/06/11 NOTARIZED: NO DOCKET ¹ FACIL:50-387 Susquehanna Steam Electric Stationi Unit lE Pennsylva 0500 50 388 Susquehanna Steam Electric Station> Unit 2i Pennsylva AUTH, NAME AUTHOR AFFILIATION CURTIS < xI ~ '8 ~ Pennsylvania Power 8 Light Co ~

REC IP ~ VAME RECIPIENT AFFILIATION SCH'HENCEREA ~ Licensing Br anch 2

SUBJECT:

Forwards revised response to NUREG 0737r Item II BE 2Eto post accident sampling 8 analysis areas 'ddress to ver i fy tnat doses from containment were used in analysi s 8 Completes action on SER Outstandinq Issue 80. zones'IZE:

DISTRIBUTION CODE: 80018 COPIES RECEIVED:LTR J ENCL J kg TITLE: PSAR/FSAR AMOTS and Rel ated Cor resoondence NOTES; Send I8E 3 copi es FSAR 8, al 1 amends. 1 cy'.BlrrR LRG P'4I(L ~ RIB) 05000387 Send I8E 3 copies FSAR 8 all amends' cy.'BAR"LRG PM(CRIB) 05000388 RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LT1R ENCL IO CODE/NAME LTTR ENCL ACTION: A/0 L ICENSNG 1 0 LIC BR ¹2 BC 1 0 LIC BR ¹2 LA 1 0 STARKERS 04 1 1 INTERNAL: ACCID EVAL RR26 CHFM ENG SR 11 1

1 1

1 AUX SYS BR CONT SYS BR '927 1 1

1 1

CORE PELF BP. 10 E~cRG PREP 1

1 1

0 EFF TP SYS BR12 ELIRG PRP OEV '5 1 1

1 1

EMRG PRP L IC 36 3 3 ,

EQUIP QUAL BR13 3 3 FE'<A-REP r)IV 39 1 1 GEOSC IENCES 28 2 2 HUI FACT FNG 40 1 1 HYO/GEO BR 30 2 2 I 8,C SYS BR '16 1 1 I8E 06 3 3 LIC GUIO BR 33 1 1 LIC QUAL BR 32 1 1

'UIATL FNG BR 17 1 1 ~ECH E "IG BR 18 1 1

'c1P A 1 0 NRC PDR 02 1 OELD 1 0 OP LIC BR 34 1 1 PO"c'rER SYS BR 19 1 1 PROC/TST REV 20 1 1

! QA BR 21 1 1 RAD ASSESS BR22 1 1 REAC SYS BR 23 1 1 01 1 1 SIT Ah! AL BR 24 1 1 CT ENG BR25 1 1 "XTERNAL: ACRS 41 16 16 LPOR 03 1 1 NSIC 05 1 1 JUN16 $ 81 S7 TOTAL RURSER OF COPIES RIERUEULOTTP PA ENCL

TWO NORTH NINTH STREET, ALLENTOWN, PA. 18101 PHONEr {215) 770-5151 NORMAN W. CURTIS Vice President-Engineering 5 Construction-Nuclear 770.538I June 11, 1981 Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555 SUSQUEHANNA STEAM ELECTRIG STATION SER OUTSTANDING ISSUE ¹80 ER 100450 FILE 841-2 PLA-810

Dear Mr. Schwencer:

Attached is a copy of the revised response to NUREG-0737, II.B.2 which has been. revised to address the post accident sampling and analysis areas and revised to verify that doses from containment were used in the analysis and the zones as indicated on the radiation zone maps.

This completes our action on SER Outstanding Issue ¹80.

Very truly yours, KcQ~~+

N. W. Curtis Vice President-Engineering and Construction-Nuclear cc: R. M. Stark goo/

j/(

PENNSYLVANIA POWER IL LIGHT COPhPANY P g O 6 y g OQQQ ~

~ ~

drywell equipment drain tank. {Subsection 5 1 and Piqure 5.1-3a) .

2 Normally open reactor head vent line 2 DBA-112 which discharges to main steam line "A". (Subsection 5 1 and Figure 5.1-3a) .

3. Hain steam driven RCIC and HPCX system turbines, operable from the control room which exhaust to

- suppression pool. (Subsections 5.3, 6 3 and Pigures 5 4-9a, 6.3-1a) .

Although the power operated relief valves fully satisfy the intent of the NRC requirement these other means also provide protection against accumulation of. non-condensables in the RPV The design of the RCS and RPV vent systems is in agreement with the generic capabilities proposed by the BMB Owners Group, with the exception of isolation condensers. SSES is no equipped, with isolation condensers. The BQR Owners'roup position is summarized in NEDO-24782 Operation of the equipment described above during abnormal operatinq conditions is controlled by the Emergency Operating Procedures While these procedures do not specifically address venting of non-condensable gases, they do address proper utilization of equipment to recover from undesirable conditions presented by the presence of non-condensables or by other circumstances The RCS and BPV vent systems are part of the original SSES design basis. A pipe break in either of these systems would be the same as a .small mainsteam line break. A complete mainsteam line break is within the desiqn basis (see Subsections 6.2.1.1.3.3.2 and 6 3.3). Smaller size breaks have been shown to be of lesser severity (see Subsections 6.2.1.1.3 .3 5 and 6.3 3 7 3)

Therefore, no new supporting analysis is necessary in response to NUREG 0737. Tn addition, no new 10CPR50. 46 con formance calculations or containment combustible gas concentration calculations are necessary Non-condensable gas releases due to a vent line break would be no more severe than the releases associated with a mainsteam line break. Hainsteam line break analyses included continuous venting of non-condensable gases with high hydrogen concentrations These analyses demonstrate conforma nce to 10CPR50- 46.

X.1.20.1 Statement. of Requirement Pith the assumption of a postaccident release of radioactivity equivalent to that described in Regulatory Guides 1 3 and 1.4 (i e , the equivalent of 505 of the core radioiodine, 3.00% of the core noble qas inventory, and 1% of the core solids are contained in the primary coolant), each licensee shall perform a radiation

0 0 and shieldinq-design review of the spaces around systems that

~

. may, as a result of an accident, contain highly radioactive materials The design review should identify the location of vital areas and equipment, such as the control room radwaste control stations, emerqency power supplies, motor con trol centers, and instrument areas, in which personnel occupancy may be unduly limited or safety equipment may be unduly d,egraded by the radiation fields during postaccident operations of these systems Each licensee shall provide for adequate access to vital areas and. protection of safety equipment by design changes, increased.

permanent or temporary shielding, or postaccident procedural controls The design review shall determine which types of corrective actions are needed for vital areas thorughout the facility X.l.20.1. 1-- Documentation Regai~ed for- Vital Area Access Po vital area access, operatinq license applicants need to provide a summary of the shielding desiqn review, a description of the review'esults, and a description of the modif ications made or to be made to implement the result of the review. Also to be provided by the licensee:

(1) Source terms used including time after shutdown that was assumed for source terms in systems.

(2) Systems assumed to contain high levels of activity in a post-accident situation and jusitification for excluding any of those qiven in the "Clarification>> of NUREG 0737.

(3) Areas assumed vital for post-accident operations including justification for exclusion of any of those given in the

>>Clarification" of NUREG 0737.,

(0) Projected doses to individuals for necessary occupancy times in vital areas and a dose rate map for potentially occupied.

areas.

X. 1. 20. 1 . 2 =

Documentation Reguiged for Equipment Qualification II 8.2 states, "Provide the information requested by the Commission tiemorandum and Order on equipment qualification (CLI-80-21).>> This memo andum with regard to equipment qualification, requests information on environmental qualif ication of saf ety related electrical equipment X 1- 20. 2. ~ Inte~oreta t ion-X.l.20.2.1 = Source Terms-

l The source term for recirculated depressurized coolant need not be assumed to contain noble qases, therefore the RHR shutdovn cooling system vhich may initiate at low reactor pressure only will. be assumed to contain 'solely halogens and particulates The HPCI and LPCI systems do not.recirculate reactor coolant but rather, suppression pool water. They void of noble qases vill also be essentially Leakage from systems outside of containment need not b considered as potential sources. Also, containment and equipment leakage (from systems outside containment) need not be considered.

as potential airborne sources vithin the reactor building Zt follovs that airborne sources and any other uncontained sources in the reactor building do not need be considered. in this shielding review.

X l,2Q,2 2 - Post-Accident.Systems The standby gas treatment system, or equivalent, is given as a system vhich may contain high levels of radioactivity after an accident. Airborne activity from leakage of equipment outside containment has been clearly established as being outside the review requirements Dryvell leakage must then provide the activity processed by the 'toSGTS. This review vill assume the drywell does indeed leak the reactor building to provide a source vithin the SGTS. Hovever, this airborne sour e vill not be evaluated any further in the reviev.

X 1-20 2 3 Eauioment Qualification Provide a description of the environmental qualification program and results for safety related electrical equipment both inside and outside of containment.

radiation qualification of It is our understandinq that non-electrical safety related equipment need not be reported 3

X.l 20 3 Statement of Response The required post-accident study is divided into tvo parts: one

.dealinq vith a summary of the. shielding design review plus vital area access, another dealing vith equipment qualification summary of the shielding design reviev, results, and'ethodology used to determine radiation doses is presented belov The results of the equipment qualification progran are scheduled to be submitted in April 1981 in revision 2 of the SSES Environmental Qualification Report for Class 1E Equipment The results of the shieldinq reviev of contained sources are that all vital areas are accessible post-accident and no shielding modifications are necessary to comply to BUREG 0737

X 1 20 3-1 ~

In trod uctgon If an accident is postulated in which .large amounts of activity are released from the reactor core, then pathways exist vhich can transfer this activity to various areas of the reactor building These larqe radiation source'terms present a hazard rega-ding potentially hiqh doses to personnel Zn order to deal vith this problem it has become necessary to quantify these source terms trace their presence and determine their effects on the efficient performance of post-accident recovery operations To this end the plant shielding of the Susquehanna Steam Electric Station, Units 1 and 2 ~ has been revieved for post-accident adequacy This summary presents the analytical bases by'hich the review vas carried out. Systems required or postulated to process primary reactor coolant outside the containment during post-accident conditions vere selected f or evaluation Large radiation sources beyond the original selected systems Radiation levels in adjacent plant areas due to contained sources in piping and equipment of these systems vere then estimated to yield the desired information. Also. included herein is a discussion of radiation exposure guidelines for plant personnel, identification of areas vital to post-accident operations and availability of access to these areas.

As a byproduct of this reviev, several radiation zone maps and associated curves have been produced The maps vill allow operations personnel to identify potential high exposure vital areas of the plant should an accident occur vhich contaninates the system considered in this study. The curves vill allov them to estimate radiation levels in these areas at various times followinq an accident.

X 1 20 3 2 . - Desian Review Bases X 1 20. 3 2 1 Systems Selected for Shield jug Revie~

A reviev vas made to determine which systems could be required. to oper'ate and/or be expected to contain hiqhly radioactive materials folloving a postulated accident where substantial core damage has occurred The documentation governing the approach to the shielding review is NUREG;0737 review of containment isolation provisions vas conducted in accordance vith item IZ;E.O 2. This was done to assure isolation of non-essential systems penetrating the containment boundary.

Thus, systems other than those identified as having a specified function follovinq an accident are assumed not to contain post-accident activity and do not need to be considered in the shielding reviev X-1,20 3-2-1 g Core Spray~ HPCI~ RCIC and RHR gLPCl modeL

~

The Core Spray. RHR

~ (LPCX mode),

HPCX (vater

~

side) and RCXC (water side) systems would contain suppression pool vater being injected into the reactor coolant system. Althouqh the HPCZ and RCXC systems. could also drav from the condensate storage tank, suppression pool water is assumed to be their only source of water for injection. The steam sides of the HPCX and RCXC systems vould operate on reactor steam and vould not be required the reactor is depressurized. However, as a first estimate it

'hen for equipment qualification, is assumed that these systems should a,iso be available until one year post-accident 4

X 1.$ 0.$ ;g.l. p- RHR /Shutdown Cooling Node) ~

The RHR system recirculates reactor vaste vhen it operates in the shutdovn coolinq mode operation in this mode requires that the reactor be in depressurized condition Depressuzization is expected to remove substantially all of the noble gases released into the reactor coolant vhethez it be by direct venting to the dryvell or by quenching reactor steam in the suppression pool consideration is. following a postulated serious 'nother accident, the HPCI, RCIC, RHR (LPCX dode) and/or Core Spray systems would inject a substantial amount of vater into the reactor coolant system This shielding review vill assume that there are no noble gases in the reactor vater in the RHR system from the shutdovn cooling mode However, since the exact amount of'ilution of the reactor water is difficult to determine, no dilution in addition to the reactor coolant volume is assumed.

X.1.20.3 2.1.3 RHR ]Suppression Pool Cooling Node)

The RHR system in this mode circulates and removes heat from suppression pool vater to prevent pool boiling. This assures availability of suppression pool vater as a source for cooling the reactor and, increases the efficiency of a given cooling operation vith this source.

Under post-accident conditions, vater pumped from the suppression pool through the BHR heat exchanger may be diverted to spray header system loops located hiqh in the dryvell and. above the suppression pool This mode of operation provides the ability to reduce dryvell pressure by condensing atmospheric steam 'awhile cooling the suppression pool vater No credit is taken f or spray removal of iodines X 1.20.3.2 1 5 CRD Hydraulic System ~

The operation of the CBD system vas revieved to determine scram discharge headers vill contain highly radioactive vater if the followinq a postulated accident. Prior to a scram the CRD X.l-25

housings contain

~

condensate water delivered by

~the CRD pumps-

%hen a scram occurs some of this condensate water from the CRD is discharged to the scram discharqe header. After the scram, some condensate and reactor water flows to the scram discharge header which fills in a matter of a few seconds Since the vents and drains in the scram discharge headers are isolated by the scram, all discharge flow then stops is not reasonable to assume, that significant core damage occurs Since it in the first few seconds followinq a scram, the scram discharge header will initially contain only a mixtur of condensate and pre-accident reactor water following this postulated accident After the reactor scram, the scram discharqe and instrument volumes will contain about 700 gallons of pre-accident water isolated by initial scram a single drain valve leak tested to 20 cc/hr closed the drain valve, then this leakage is If the insignificant compared to the scram discharge volume and insignificant as a post-accident concern. If the drain valve faiks to close, operator action is required to reset the scram and close the soft-seated scram discharge valve If is not taken or fails to close the valve, then post-'accident this action sources can enter the liquid radwaste system by leaking past the CRD seals. The CRD withdraw line does not directly communicate with the reactor coolant.

In light of the anticipated small leak rates and the lack of single failure criteria consideration requirenents, the scram discharge drain valve was assumed to remain closed and any leakage was disregarded.

X 1 20 3.2.1 6 RHCU Svstem For a ma d'or accident with resulting core damage, the RQCU system would automatically isolate on a low reactor coolant level signal.

and would contain no hiqhly radioactive materials beyond the second isolation valve Since the cleaning capacity for this system is small, accident recovery and it it would be impractical to use it for TNI type is excluded from this shielding review X 1 20.3 g 1 7 Liquid Radwaste Sgstem-Equipment drains and compartment floor drains servicing ECCS systems are isolated from the reactor build'ng sump. All piping that may contain hiqh activity post-accident water is also isolated from the reactor building sump and radwaste systems CRD system isolation is discussed in Section X 1 20 3.3.1 5.

Since no significant amounts of post-accident activity can reach the liquid radwaste system, it is excluded from this shielding revie w.

X-l,20.3,2.1 8 NSIV Leakage Control System X 1-26

0 ~

Subsequent to a postulated accident, system operation may begin upon actuation of the manual switches in the control room. This system may only be activated upon a permissive reactor pressure siqnal (35 psiq) . The method used to depressurize the reactor to this level has a large effect on the amount of activity potentially available for passaqe through this system. For example, the HPCX system can deplete the reactor steam activity considerably vith only a few minutes operation Whichever depressurization method is chosen, the HSXV-LCS system remains as one that must be included in the shielding review.

X 1 20.-3 2 1 9 Sampling.Systems Sampling Systems required or desired for post-accident use include the existinq Reactor Sampling System, the Containment Atmosphere Honitorinq System, the Plant Vent Sampling Syste~ and.

the Post-Accident Sample System Each of these systems/stations may contain post-accident sources and is included in the shielding review.

X.l 20.3.2.1.10 ~ ~

Standby Gas Treatment System The Reactor Building Recirculation system is used after an accident. This disperses airborne activity throughout the reactor buildinq and refueLing floor. The SGTS system collects airborne activity, concentratinq halogens vithin the charcoal filters while releasing noble qases outside the secondary containment The charcoal filter is considered to be a source of contained activity and is included in this shieldinq revie~ The assumptions used in determining @his contained, source are=

1) .Drywell leakeaqe at 15 per day.

2) SGTS process rate of 1 reactor building/refueling floor volume per day

3) 99% charcoal .filter efficiency for halogens 0% charcoal filter efficiency for noble gases.

X 1 20,3.2 1.11 Containment Atmosphere JDrywell)

The free volume of the primary containment is assumed to initially contain Larqe amounts of post-accident activity, namely 100% of the core noble gases and 25% of the core halogens Shine throuqh the dryvell wall vas examined to determine the effects on reactor buildinq radiation levels. Results indicate the'six foot thick dryvell shield wall reduces shine to radiation Zone I levels. Shine through penetrations presents no additional hazard because pipinq is directed to penetration rooms vhere area dose rates vill be dominated by internal piping-X 1.20. 3.2 1. 12 Sugoression Pool QNetvell}

X. 1-27

1 The suppression pool is assumed to initially contain 50$ of the core halogens and 1% of the core particulates post-accident Shine" through the vetvell vali vas examined to determine the effects on radiation levels in the reactor building determined that the six foot thick vetvall shield vali reduces It vas vetvell shine to radiation Zone I levels in the reactor building X.l.20.3.2 2 . -Radioactive Source Release Fractions The follovinq release fractions vere used as a basis for determining the concentzations for the shielding reviev=

Source A: Containment Atmosphere: 100% noble gases 25$ halo gens Source B: Reactor Liquids: 1005 noble gases, 50% halogens 1% solids Source C: Suppression Pool Liquid:50$ halogens, 15 solids Source D: Reactor Steam: 100> noble gases, 25% halogens The above release fractions vere applied to the total curies available for the particular chemical species (i.e, noble gas, halogen, or solid) for an equilibrium fission product inventory foz Susquehanna as listed in Table X.l 20-1.

The Regulatorv Guide 1.7 solids release fraction of 1~ vas used for Cs and Rb on this reviev., Further evaluations of the TAXI radioactivity releases may conclude that higher release fractions are appropria.e. However, until the release mechanisms and release fractions have been quantified, the existing regulatory guidance vill be folloved. No noble gases vere included in the suppression pool liquid (Source C) because Regulatozy Guide 1 7 has also set this precedent in modeling liquids in the pool (See Ref.. 3 and. 9) . Furthermore cursory analyses have indicated that the halogens dominate all shielding requirements and that contributions to the total dose rates from noble gases are neqliqible for the purposes of shielding design reviev X.l.20 3 2 3 -Source Term Quantification.

Section X 1 20.3.2.2 above outlines the assumptions used for release fractions for the shielding design reviev These release fractions are, however, only the first step in modeling the source terms for the activity concentrations in the systems under reviev. The important modelinq parameters, decay time and dilution volume obviously also affect any shielding analysis The follovinq sections outline the rationale for the selection of values for these key parameters.

X-1 20 3.2.3 1 Decay Time X 1-2S

Por the first stage 0of the shi'eldinq design review process minimal decav time credit was used with the above releases .

primary reason for this was to develop a set of accident radiation zone maps normalized to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> decay X.1 20. 3 2 3.g Dilution Volume The volume .used for dilution is important, affecting the calculations of dose rate in a linear fashion. The following dilution volumes were used with the release frac ions and decay times listed above to arrive at the final source terms for the shielding review:

Source A: Drywell and suppression pool free volumes Source B: Reactor coolant system normal liquid volume (based on reactor coolant density at the operating temperature and pressure) .

Source C: The volume of the reactor coolant system plus the suppression pool volume.

Source D: The reactor steam volume.

X.1 20 3.2 4 .System~Source Summary Core Spray System: Source C 0 High Pressure Coolant Injection System Liquid: Source C Steam: Source D (with credit for steam specific activity reduction due to turbine operation) 0 Reactor Core Isolation Cooling System Liquid: Source C Steam. Source D (with credit for steam specific activity reduction due to turbine operation)

Residual Heat Removal System LPCI Node: Source C Shutdown Cooling Node= Source B (with credit for noble gas release during vessel depressurization) .

Suppression Pool Coolinq and Containment Spray Nodes: Source C tlain Steam Isolation Valve-Leakage Control System Steam: Source D (with credit for steam specific

0 Activity reduction due to RCZC turbine operation)

Samplinq Systems Containment a'r sample: Source A Reactor coolant sample: Source B Plant vent sample: 15 per day Drywell leakage following the filtration by the Standby Gas Treatment System (see Section X.l 20 3 2 1.10 for discussion of SGTS source assumptions) 0 Standby Gas Treatment System Charcoal filter: 1$ per day drywell leakage (See Section X. l. 20 3. 2.1.10 for discussion of source assumptions) 0 Drywell: Source A o Met well. Source C For each of these systems, piping associated with the appropriate operatinq mode was identified on PAID drawings and traced throughout the plant to their final destination X. 1. 20. 3 2 5- Dose Integration Factors for personnel ~

Cummulative radiation exposure to persona 1 in vital areas (continuous occupancy) is determined based upon a maximum one year exposure period. The inteqrated doses are modified using Ref. 7 occupancy factors listed below Tl,me

/dyes) Occupanc~Fa ctors 0 to 1 1 0 1 to 0 0. 6 over 0 Exposures for areas not continuously occupied (frequent and infrequent occupancy) must be determined case by case that is, multiply the task duration by the area dose rate at the time of exposure X 1 20.3.3 - - Shielding Review Methodology X.l.20 3.3 g radiation Dose Calculation Model, The previous sections outlined the rationale and. assumptions for the selection of systems that would undergo a. shielding design

reviev as well as the formulation of the sources for those systems. The next step in the reviev process vas to use those sources alonq with standard point kernel shielding analytical techniques (Ref. 13 6 10) to estimate dose rates from those selected syste ms Scattered radiation (e q., shine over partial shield valls) vas considered but vas not siqnificant siace the net reduction in dose is several orders of magnitude and no vital area is separated from a hiqh activity source solely by a partial vali Radiation levels f or comPartments containing the systems under review were based on the maximum'ontact dose rate for any component in the compartment Radiation levels ia. areas not containing unshielded sources vere based. on maximum dose rates transmitted into areas through valls of these adjacent compartments. Checks vere also made for any piping or equipment that could directly contribute to corridor dose rates, i.e piping that may be running directly in the corridor or equipment/piping in a compartment that could shine directly into corridors vith ao attenuation through compartment valls There is no'ield routed small piping. (i.e., piping less than 2" ia diameter) for ECCS systems Dose rates are cummulative an d ar e summed over all systems in simultaneous operation in most cases. The exception is steam piping for the RCXC and HPCl systems Both are high pressure systems and cannot be operated simultaneously vith low pressure systems such as core spray This becomes a moot point, since these steam lines are routed in well shielded, compartments causing no appreciable personnel doses.

X 1 20 3 3.2 Post-Accident Radiation Zone HaPs one of the principal products of this reviev is the. series of accident radiation zone maps (Figures X.l 20-1 to.8) The zone boundaries used in the maps are defined in Table X 1.20-2. The zone maps present the calculated dose ra-es at one hour after the accident due to the sources described. in Section X 1. 20.3-2 0 ia various areas of the plant site The principal sources of radiation in each area are identified in Table X.l 20-0 The dose rates presented do not include contributions from normal

'operating sources which may be contained in the plant at the time of the accident since these contributions vill be minor outside of veil defined and shielded areas. They also do not include dose rate contributions due to potential airborne sources resultiag from equipment or dryvell leakage The zone maps were used to determine the accessibility of vital areas described in Section X 1 20.3.3.4.

X.1-3t

~ .

X',$ ,20,3 3,3 Pe psgggeg Radiation Exposure Guidelines In order that doses to occupied areas take on meaningful proportions, guidelines it is necessary to establish exposure goals or The qeneral design basis for these guidelines is 10CPR50, Appendix A, GDC 19. That material addresses control room habitability includiaq access and occupancy under worst case conditions Exposures are aot to exceed 5 rem vhole body, or its equivalent to any part of the body, for the duration of any postulated accident. GDC 19 is also used. to govern design bases for the maximum permissible dosage to personnel performing any task required post-accident These requirements translate rouqhly into the objectives to be met in the post-accident review as given below.

Radiation Exposure Guidelines Occupanc y Dose Rate Ob j ect iv es Dose Objective Continuous 15'R/hr 5 Rem for duration Prequeat 100 m B/h r 5 Rem for all activities Infrequent 500 mR/hr 5,Rem per activity Accessvay 5 R/hr Included in above doses X 1.20 3.3.Q - Vital Area Identification aad Access X.3. 20.3.3.0.1- Vital Area Clarification-Vital areas are those "which will or may require occupancy to permit an operato- to aid in the mitigation of or recovery from an accident" Reference (15) further defines recovery from an accident as, "when the plant is in a safe and. stable condition>>

"This may either be hot or cold shutdovn, depending on the situation The 10 CPR 73.2 definition of vital, area shall not apply here.

Por the purposes of this study, the evaluation to determine necessar7 vital areas considers all of those listed in Reference (2) . Upon examination several plant areas were determined aot to be vital Instrument panels vere excluded because essential equipment control aad alignment has been established in the control room and requires no local actions. The radwaste control room is exclud.ed because 1) no 'local actions are required to prevent spread of postaccident sources into the liquid radvaste system. -2) gaseous radvaste processing is aot required and: 3) activity sources early in the post-accident transient are much too high to be effectively processed through the liquid and, eventually solid radvaste systems Also: excluded are the post-LOCA hydrogen control system and the containment isolation reset control area (vhich are operator actuated from the main control room) . Lastly the emer'qeacy pover supply (i e, diesel generators) was excluded since system initiatioa comes from the control room and requires no local actions X.1-32

The resulting list of areas considered vital for post-accident operations at Susquehanna appears in Table X 1-20-3.. Note that security facilities aze included as vital areas vith regards to maintaininq plant security.

X g 20 3.3.0.g Vigal Area Access Those operator actions required post-accident vere revieved. to assure that first priority safety actions can be achieved in the postulated zadiation fields. This review assures that access is available and required operator actions can be achieved Ingress and eq ess area dose rates to those vital areas identified in Table 3 vere examined to ensure compatibility with the areas beinq accessed Plant effluent monitorinq stations are located at five (5) plant vents: tvo(2) for the Reactor Building, two(2) for the Turbine Building, and one (1) for the Standby Gas Treatment System The Reactor Building monitors are automatically isolated'post-LOCA, and vill contain no post-accident activity. The SGTS effluent sample station vill contain post-accident activity in sample cartridges: one {1) volumetric and one(l) charcoal filter The samples are locally shielded and present no access problems in the area of the station However, transportation and. handling of the filter cartridges vill require local shieldinq.

The Turbine. Building Plant Vent Sample Station (PVSS) may also contain post-accident activity Doses, if any, vill be a lover maqnitude than that of the SGTS effluent filters because of environmental dispersion and re-entzy to the Turbine Building ventilation system. In the vorse case, the Turbine Building PVSS doses vill be much lover than those of the SGTS. In the best case, control room personnel may shut dovn the Turbine Building HVAC system (vhich is non-safety related) . In this case, the Turbine Building PVSS may be void of post-accident a tivity.

X 1.20. 3. 0 ~

Results X 1.20 3 0 1- . Radioactive Decay Effects Results of the radiation level evaluation for .the shielding design reviev aze presented in Fiqures X.l-20-1 to 8 Table X.1.20-0 identifies the sources contributinq to dose rates in each of the plant areas shown on those figures., This table can be used in conjunction with the decay curves (Figures X 1.20-9 and 10) to estimate radiation levels at times other than one hour. The procedure for times less than one day, is to multiply the radiation level {i.e, radiation zone limit) by the decay factor given in Fiqure X 1. 20-9 For times greater than one day, it is necessary to multiply by the decay factor in Figure X 1.20-9 at 2Q hours and by the decay factor in Figure X 1 20-10 at the

~

desired decay time This procedure is conservative for areas in

vhich the sources are shielded because it does not rigorously take into account the softeninq of the energy spectrum an consequent increase in attenuation for longer decay times A decay curve for source D, reactor steam, is not included because the depletion effects due to.steam usage by HpCI or RCIC removes much of this source shortly after the accident. In addition HPCI and RCIC piping containinq source D is run in shielded cubicles and does not .contribute signif icantly outside those cubicles X 1.20.3.4 2 - Integzated Personnel Exposures-Personnel inteqrated exposures in continuously occlxpied areas

'ere calculated based on 100Ã occupancy for the first day, 60%

occupancy fzom day one throuqh four and 40% occupancy for the duration (1 year) . These calculations shoved. that personnel exposures vould be vithin the design objective of 5 Rem Exposures in Zones I II and Il? of the control structure are 0 24, 1.6 and 3.1 Rem, respectively These doses do no include the shielding effects of interior valls, equipment, etc th'erefore they represent the maximum dose to contzol building personnel due to contained sources. Personnel doses to the Horth Gate House (ASCC) and Security Control Center from contained sources vere found to be insiqnificant (i.e, O.l Rem) . These areas are a minimum of 300 feet from the zeactor building whose valls are a minimum of 2. 5 feet of concrete.

PoG Ag~'L GFM'v A ~YELLS 5, 'g~)c7'A eh Personnel doses at the Post-Accident Sample Station,hand Plant Vent Sample Station are calculated based on an estimated. task duration at specified times post-accident for a one person task force (Ref er to Table X. l. 20-3)

X 1.20,3 4 g -Reactor Building Accessihilitg-The results show that the reactor build'ng vill be generally inaccessible for several days after the accident due to contained radiation sources High radiation levels can be expected at Elevation 645'-0" (r iqure X.l. 20-2) regardless of vhich system {s) is (are) in operation. Radiation levels at, Elevation 719'-0" (Piqu e X 1 20-5) and above are expected to generally be vithin Zone IV'limits if the core spray and RHR containment spray systems have not been operated folloving the accident This is because these are the only unshielded post-accident system sources at these elevations Other system sources are contained.

in shielded cubicles.

Exceptions to these qeneral Zone IY levels are areas in the vicinity of reactor coolant and containment atmosphere sampling lines vhich are routed to the reactor building sample station at Elevation 779'-0". The dose'ate 10 feet from the reactor coolant.samplinq line one hour after the postulated accident may exceed 100 R/hr

~,

in the Reactor Building is accessible post-accident.

X 1 20.3.4.4 - Control Building Accessibility.

~

Results for contained radiation sources show that the vital area Results for contained radiation sources show that vital areas'n the control structure are accessible post-accident X.l $0 3.5 . References.

U.S. Nuclear Regulatory Commission, >>TMX-2 Lessons Learned Task Porce Status Report and Short-Term Recommendations>>

USNRC Report NUREG-0578, July 1979,. Recommendation 2.3. 6b.

2) U.S. Nuclear Requlatory Commission, <<NRC Action Plan Developed as a Result of the TMI-2 Accident,>> USNRC-0660 Vols. 1 and 2 ~ May 1980,Section IX. B. 2

3) Letter from D. G Eisenhut (NRC) to All Licensees of Operating Plants and Applicants for Operating Licenses and, Holders of Construction Permits,

Subject:

Preliminary Clarification of TMI Action Plan Requirements, dated September 5, 1980.

U.S. Nuclear Regulatory Commission, >>Clarification of TMX Action Plan Requirements," USHRC Report NUREG-0737, November, 1980, Item II.B.2 U.S. Nuclear Requlatory Commission, IE Bulletin No.79-013,

>>Environmental Qualification of Class IZ Equipment>>, January 14, 1980

6) U S Nuclear Regulatory Commission, <<Inte im Staff Position on Environmental Qualification Report NUREG-0588, December 1979-

7) USNRC Standard Review Plan 6.0, >>Habitability Systems",

Revision 1

8) USNRC Regulatory Guide 1.3, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Rater Reactors", Revision 2 June 1970

9) USNRC Regulatory Guide 1.7, >>Control of Combustible Gas Concentrations in Containment Eollowing a Loss-of-Coolant Accident,>> Revision 2, November 1978.

10) USHRC Regulatory Guide 1.89, "Qualification of Class XE Equipment for. Nuclear Power Plants,>> November 1970 Code of Pederal Regulations, lOCPR Part 50, Appendix A GDC 19,. Revised as of January 1, 1980.

12) C.

0 ~

Hichael Lederer, et al., Table of Isotopes~ Lawrence Radiation Laboratory, University of California, Harch 1968 3.3) D. S. Duncan and A B Spear, GRACE I An 'IBM 704-709 Program Design. for Computing Ga ma Ray Attenuation and

~

Heating in Reactor Shields~ Atomics International, (June 1959) .

14) D. S Duncan and A. B. Spear, GRACE for Computing Gamma Bgg Attenuation II An IBH 709 Pro~ram and Heating .in Cylindrical and Spherical Geometries~ Atomics International~

November 1959

15) Hemorandum of Telephone Conversation, S. Ford of LIS to H..

Ande son of NRC's Lessons Learned Task Force,

Subject:

THI Requirements at SHNPP, April 9, 1980

16) USHRC Regional Heeting Hinutes, Region I, Subject= THI Reviev Requirements at SHRPP, April 9, 1980

17) USNRC Reqional Heetinq Hinutes, Region IV and V,

Subject:

THI Review Requirements, 9/26/79 X. 1-36

X.1.20"1 INITIAL CORE ISOTOPIC IiiENTORY( )

~Isoto e Curies ~isoto e Curies Isotere Curies I--131 8.66+7 Y---93 1.82+8 TE-129 2 '8+7.

I"-132 1.29+8 Y---94 1.61+8 TE131M 1.31+7 1--133 1 ~ 99+8 Y---95 1.84+8 TE-131 . 7.74+7 I--134 2.32+8 ZR""95 1.84+8 TE-132 1.29+S 1--135 1.82+8 ZR--97 2.86+8 TE133M 1.40+8 I--136 9.22+7 NB-95M ~

3. 81+6 TE-133 8.93+7 BR--83 1.52+7 NB--95 1.91+6 TE-134 2.05+8 BR--84 2.74+7 NB-97M 1.78+8 CS-137 1.13+7 BR--85 3.84+7 NB 97 1.85+8 CS-138 1 ~ 90+8 KR-83M 1.55+7 MO"-99 1.84+8 CS-139 1.93+8 KR-85M 3.87+7 MO-101 1.49+8 CS-140 1.76+8 W--85 1.31+6 MO-102 1.19+8 CS-142 9 '2+7 KR""87 7.44+7 MO-105 2.05+7 KR 88 1.04+8 TC-99M 1.63+8 BA137M BA-139 1.. 75+8 1.87+8 KR--89 1.37+8 TC-1.01 1.49+8 BA-140 1.87+8 3Z133M 5.06+6 TC-102 1.23+8 BA-141 1.87+8 XE-3.33 1.98+8 TC-105 2.65+7 BA-142 1.71+8 XE135M 5.36+7 RU-103 8.93+7 LA-140 1.87+8 3Z-135 1.87+8 RU-105 2.68+7 LA-141 1.. 90+8 M-137 1.79+8 RU-106 9.84 7 LA-143 1.74+8 XZ-138 1.76+8 RU-107 5.65+6 LA-142 1.74+8 SE 81 4.17+6 RH103M 8.93+7 CE-141. 1.90+8 S" -83M 8.63+6 RH105M 5'.62+6 CE-143 1.75+8

.SE- 83 6.55+6 RH-105 2.68+7 CE-144 1.45+8 SB 84 2.92+7 RH-106 1.16+7 CE-145 l. 15+8 RB-"SS 1.07+8 RH-107 5.65+6 CE-146 8.81+7 RB--89 1.42+8 SN-127 3.27+6 PR-143 1. 75+8 RB--90 1.72+8 SN-126 1.75+1 PR-144 1.49+8 RB"-91 1.62+8 SN-128 1.. 10+7 PR-145 1.15+8 RB--92 1.31+8 SN-130 5.95+7 PR-1.46 9.14+7 SR--89 1.42+8 SB-127 3.87+6 ND-147 6.70+7 SR--90 1.14+7 SB-128 1.64+7 ND-149 3.24+7 SR--91 1.73+8 SB-129 2.20+7 ND-151 1.19+7 SR-"92 1.58+8 SB-3.30 5.95+7 PM-147 3.45+7 SR 93 1.67+8 SB-131 8.03+7 PM-149 3.24+7 SR 94 1.28+8 SB-132 9,97+7 PM-151 1.25+7 Y---90 1.72+8 SB-1.33 1.01+8 SM-151 2.70+5 Y-"91M 1.01+8 TE127M 1.04+6 SM-153 4.70+6 Y---91 1 ~ 70+8 TE-127 3.87+6 Y--"92 1.76+8 TE129M 1.04+7 (1) Based on 1000 reactor operating Cays at 3440 MWt. Refereace GE Internal Report Documeat,. "Summary of Fission Yield for U-235,'-238, and PU-239," published by Meek and Rider, June, 1977.

(2) 8.66+7 = 8.66x10 X.1-36

NOTES:

(1) At one hour post-accident, the steam source (D) will dominate area radiation levels. Following reactor steam activity depletion, radiation levels will be due to contained source (c) .

Radiation levels are based on system/source proximity, however, in this case each system noted contains source (C) . Therefore, radiation levels may be determined as a function of time by referring to the same curve on figures X.l.20-9 and X.l.20-10 for source (C) .

't

~ ~

Ihlh hthtthhthatlhtttthttttlht llh I ~ II

<<r~t tg+m na<<; <<a<.~:

0 I I I

I~

'J.

I

.f I iI I

~a.l taaal IJ Ital I~

I r'Ilatllaa alai lla ~

sr<<

I aJ I- '~l:f 0 tll pilot;-'-l~ Li:.I 'I 0<<

0 s ;II Q

VV Q.

8' 8 0 5 I3 8 6 6 5 8 8 h

I~ l jI'

~h ~aaa<<I ~II tlaal I(~ I Ilatalaa alallht

~ ~

rgg-at c C PI ~ W << ~ Ql I l I

ar Qs I ~ I Qr I

44~

<<wow <<sat% <<<<s<<ra<<as I st <<a<<Swl<<ra<<~pals<<<<<<ee<< Il~, I ~ rl/Il 8 e Q" 0 e 8 <<3 hl<<lilll

~ <<<<s sattwltstvsst r<<

~<<ruwwstrwl Ict << ~ l <<<<

'I XDiXFidwTiI'ral ll atlatl 4 ia atf.aas 6'LR~a<<<<<<t,lltai(.

I~

'I Li I C ~

' 1 NINNNNNNNNNWIINNNI I I NI ~

Nll ~

Tvtv>> Qao At>>+

e 899 6 eO'p" 6 8 v>> ~

~iS 0 ~ N>>t ~ tV NNe

~ >>Ql~

1L11% ~ N f/Olg Nllt' ll 7NN'IC

~J~>>t ~ ~

I

~c>

~

NNIDN Ntttt AI IN

>>i ~

It it Ill Atttle

~ ltlC'IN:~ I AADlhTION LEVELS

" lIlliIN0 It II

f NININNlmINNNNmINNINNNNNNIN f Oll ~

" 0 CC P tt CC CC Ql-

~ --5 Nizam Pt 8 8 8 Q Qt . e 9 8 Q 8 8 8 Qo g>> Oi I

I \ i I t I I

I ~ tete I~

Fe<<N>>t AIL>> vteev>>otvl el eve

' cgeveN<<et'I

!s C g' V

I 1.. CJCC ~

'C II ~ I- ~ "I. I ~

N CI Ct <<<<<< I I.L4, .

0-l T ~ ~

• I ~ - I C u

~

$8 I

G" tt G 0

~ t r=> ow e

>>>>'I <<1 II >> 'C t'

CNC IJ ~

I

'I I Pe Ie'I ~

I Ctl I'I I I

, te1 C-J

~ 1 g.1 0 I-ICw C1 C=

C fel 9 CI<<4 ~ I~ ~ C vo I t'o'3 pef'

~, 'J .CL>>

~

J gf,

,~j

l. LLCNIIL

~ t e ~

~ low o N te>> tee%>>

~ ottl>>t>>ltle elcceoclltlE>>

&LB ~ Lh BLR IIAOIAIIOIILLVEL3 Ie't ttCt,Nt I NCI,N ~ I let t

IIII I lllllllllllllllllllllllltNllIlla(a(

I f !l>JO f f'l.

g~ f

)

L-.

a aaaa aaaa ~

i/i r~ ~rZ '-> 'I<< '<<FTHM' Lj ~ ~a ~

0 (a

~ ~ 'F aaa(g ~

~ -.:-.=..-:-..P'~ -.>

Q rf.::".. ar r Oa

~

k-'f(-P ='

I ~

~I

~a Oa-

&-- r. I a

r 0

~ . ( v I ~

h~a aa 0

(>> << I taaaa( ag I IIKaal4aarf

(~ +"IL Qo- L.

QA I

Q~

xi] ~

....,.<...I Qa(

u - sob. I Far>> <<. lag>> r ca'a. In'e'In~(a ~ aoa

~ ~

./ QI j'j'L".1I 1 fi j'Pj 0 .

8- f $ ~

= Pq&0<<aawa a(ffg I(I I A/I/II I

~ ~affl Sfc ~

oaaa<<aK<<aaKaaw

~ <<aa a Faa<<+(IA((I~If(~((I ~ <<aaa<<K<<aR Iaaa<<<<aa f(AD(Af(OI(lEVELS A P a[== I=.=P

~ ~

I I lltlltmmml mtmll I INIIII

~ II~

"r.- ,0 Q tL I"~

,,-!/,I ~ I

}8 I! li

+R' "-}

-s~4

~ riii'III'!

I 'I&~lit P'5l

>>Ilili >" ~~ <>

""""Ol

~ ~

~<<>>}CCS>>>>>>

1 (s

I 8>> I~ 8 O~- . ~sl I'Igf>>IL 4 4S L ~ a pi a 8 I 0

'Ij'

)

QC.

I I ( ~

C" 0 s,s ~ 5 j PgpJ/

)

, P.I, i gci I I ~

0 '....,a

<< li is Q I Ql a<<v

-Il

~ ~

J I

CZ I

Q>> p~

~ ~ CSICW I ~

>>.<VIIIJ>>

It I CSI r- Sg l ~

) Cj L) I}

~ m g% Qtt Qt gi Ol

\'I <<IS~<<>> IS>>~s~~>>S }~c>>

Q L>>

C 8

~-l-)

8 ~V 0'f4" 'I y tt ) ss C I'Is "I

l i .. jl'-=

l J

Qi

~ <<<< ~ Qi I

I ~

1 ~ ~ ii I 4>>II~

l.l.. I.. o 0 ~ -IJ.- .~ 0 -.

CLCIILL

~ >><< ~

~ lll>>4 CCIC>> NISN

~ >>IIS>><<a IIIt>> I IC>>C IICICIS Cps ILILI<< ~ ILIC~LIL~ flADIAIIOII LfVLLS I.l~lldltll >>t'I'I>>>>~S

&~I".itsy I:.- lt c-.c ~ .I, m - I

II I I I Q I I aee ~

8 I

gQ~~ll,'I:...J t~J r

Le I

Qe a:O.CP=

ee ~ ~ ~

E')'ii ', 8 r .'

rQ'":i<

I ) w eeeetc eeaceeeaeae 0 - ~a F~ C I'e .,lil I r (rl IIIIICQ

. ",Ql (>>1-W'l.  ! i >Q f ~ IIIISI(

~ ~ e are>> ree J~f n l -u -uI I ale 5( ~ IIC( ~

~ Ia el I I I II I I ge n Par OOIIIIQI I~i .

fr Tte I ~I Ve 1

~

~

A>> ~

~

r e>>eQ

~.

~

la

~

SCCISI Iat Ilellelae C( ~ I ~ el ICC(55 III'"

Clallll Ill% (I llllleltl Clalell 12 ~ I

~ I~

el(IS Ill Slllll lllf. I l(l. I ~ el eel

~

~

(CS(tte<<

~ aeW eewleeeer ISC (I. 121'" aaae rare llew Il (cree eeeeue

~ SC li.

ee51 ICCII(~ I lll 5leell Slellae eaelllll tt. 121'"

III'" AAO(A(ION lEVELS

~ ICCI ~ IQI

' 52 ~ Illa (l. ~ ee e)SVI ee 'ee'a

~ ee ~ I I al Iee .lee(le fl OII'" el

~ ~

Atilt NNNNNNNItr<<NNNN tllr I I 5 NII ~

I ~~

fJ, I

~4l grte A>>AN pgv f aa&v ~l 5 I rt p5>;;~

Pl 8 ya

.,5 'V"f "ll'I Ci:

lllca

~ll. Cl 0--

o~ 5(r I ' 'I<<'h rt 'l ~ ~<<a At'cr<<15'I ' tt>>CI <<<<a Ctr I I+J Jt 0

Qa--.

Qa--

0-I l ~ ~ '.l')'j. 4 A<<a' l'ka 0r ~ aria I

Cap

'Ct

~3)eh

~ Pl 0

a 8 8 III'550 8$ 8 I

I g5 l 'tt-(

A Pa I

j J.

! i SLCCCLLL I~

alt ~I ~. ~ . J l~ ~ a IIla rlvarttlItlINCrl tati<<

~ art I <<>>r I ~ C AAOIAlI051 LLVLLS II cl88 III I cl I 55>> If 55 AAI ~ Ct Ac r>> ~

c Wf It\I IM ~ I 'h h

~ ~

~ ~~el<<l<<INIINNINIIN f I I I ee<< ~

~ ~

I~~K/~y/

e) i

~~ I el ~ N

~ '~reer

~ r 2

I ~e er<<N ~

rehem 4 awe ~ We 5

0 ~ <<I ~~

~ eo fl 0 ~ w ~ ew Kl 12

~.e~e~

e 12 Ii IS III A"" Il 4 IO

<<le '0 re~ e e

<<e fo t

IIICHlfL l I ~ lie jlelreN'll>>

M eeO<<reer&el ecol<<IIM Il<<N<<l

~ ~

IOII LI:VCLS 'AOIAI IIII rleN

~ me e

r. I

0. I k

I ~

I i'd I 1.20 1

R.-'. TO 0: >0 -IL:.4=RGY - I SS

~

0%

<<TV (~) T <<IT e - ~f ~if C T OS:- FO: R "-0% SO'WC:-S e I A .-'. ~

I ~ ~:

I '

I I I I I I

~

I I

~

~ I I

P

~

I

~

I I I ~ I ~

I I I ~

I I I i ~ ~ i I ~ I i I I

.j:

I I i "j ~ I ~

I I j 100

~ ~ ~ j

~ -N

.;.1 I

~

I I ~

~

~

~l: N-- -e- ~

~

I

~ ~

~

s ~

~

vl:!

'!il.' 'I <<

~

I

~ s ~ i  ! ~

i ~

I I ~ ~

10-1 ~ ~

1

~

i ~ ~

1

~ ~

I l~

~~

- SOURC SODRCE SOGRC

~ ~ ~

'~

~ ~

I ~ ~ ~ ~ ~" ~ ~ ~ ~ ~

~

~

~

~ I I I ~ \ 'I I ~ I ' I ' ' I I I ~ I I I ' , I I I..l  : ~ ~

I I'

I I

I I I I .  !

' I I I i ~ i I

'I 0 2i j i t i. j I i  ! i j ~

j . j I 0 2 4 6 8 10 > 2 14 16 18 20 2" 24 T ee('n sj

1OO

~

g< ' ~ ~ I ~ X.l.ZO-IO e

TL= (e) TO .-.L =i=-.,CZ ~SS-oy Z.,-

Ol.c. DAY =OH, SO'RC=

~ ~

~ ~ ~

I ~ ~ I ~

I I ' ~

I"

~ ~ II ~

I

~

~

I I ~ I I I I ~

I I: IIII I I

~ ~ 'I I

~ II' I I - ~

~

~ I I 'II I I

I I

I' I I I I I II I II ' 'I III

~ ~ ~ ~ ~

Ih'I ' I'

~ ~ ~

10 ~ . ~ 1 ~ I ~ I ~ I ~ ~ ~ ~ ~ ~

I~ ~ ' I ~ I ~ ~ ~ I I ' I I I I ~ I ~ a I

~

I

~

~

~

I, ~ I I+I ~ ~ ~ ' ~ 'I ~ I ~ ' ~

~ ~ I I

~

I I I ~ I I ~ I ~ ~ ~

~

I

~

I

~

~

(

rQ r~

I ~ ~ ~

': I I I I

'll

~ ~

';I I I ll I~

~ I I I I

'I 1'v I II ~ I

/.,

~

~

~ ~ ~ ~ ~

I I

~ ~ ~ I ~ I I. I I,l I I I I I

!/: I /','1

~ I

~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~

I I

/'/ I I I I I

~ ~ ~

~ .

I I I f /g; ~ '.: I I ski ~ I / ~ I I ~ 1 ~ ~ ~ ~ ~ ~ ~ ~

/ ~ l ~ I '.I ";' i I I 1

SOURCE A

~ ~

SOURCZ 3 '

I I SODA C

~ ~ ~ I III. ~

~

I ' I I ' '

~

II III III

~ I ~ ~ I ' ~ I ~ ~ ~ ~

~

I ~ I~ ~I

~

~ ~

~ ~

~

I I I I I I I I I I

~

'I

~ I I;'ll IIII

~

~

I ~

~

~

I

~ I I I I I I I ~ I'lI ~ ' ~ I I I ~ III ~ ' ~ I ~

1O1 10+2 10+3 10'~e (h s)