ML20040G071

Responds to NRC Re Violations Noted in IE Emergency Preparedness Appraisal Rept 50-220/81-18. Corrective actions:re-evaluated Interim post-accident Sampling Equipment

ML20040G071
Person / Time
Site:	Nine Mile Point
Issue date:	10/20/1981
From:	Lempges T NIAGARA MOHAWK POWER CORP.
To:	Haynes R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
Shared Package
ML20040G055	List: ML20040G054 ML20040G063 ML20040G067 ML20040G071
References
NUDOCS 8202110213
Download: ML20040G071 (23)

Text

,

NIAGARA MOH AWK POWER CORPORATION NIAGARA ~

MOHAWK 300 ERIE BOULEVARO WEST SYRACUSE, N.Y 03202 JJ.^81.EIO18-o October 20, 1981 Mt. Ronald C. Haynes Director United States Nacicar Regulatory Commission Region I 631 Park Avalue King of Prussia, PA.

19406

Dear St. Haynes:

Your Letter of September 3,1981 identifLed comnibnents made to.

your staff as a result of an Emergency Preparedness Appraisat (Inspection Number 81-18) performed at the Nine Mile Point Nuclear S.tation.

The purpose of.this Letter is to transnit totitten reports of evaluations requested by I.tems 3 and 4 of your Lettet.

Attachmott I ptovides our re-evaluation of istterim post-accidertt sampling equipmeist and ptocedures to determine maximum concattrations tJiat could be handled and analyzed under accidertt conditions.

Attachmott II provides a re-evaluation of our ability to rapidly and accurately deteet and measure airborne radio-iodine concattrations under field conditions ist the presence of radiation levels due.to noble gas es.

It is our understanding that.the attached reports and their transmittal satisf.ies the comnitmatt made.to your staff on August 27, 1981.

Vert) truly yolvts,

L

. ? t L-71 Thomas E. Lempges Vice Presidatt Nuclear Generation PV/mtm Attachments 8202110213 820125 PDR ADOCK O*000220 0

-PDR

l ATTACHMENT I EVALUATIONS OF THE NINE MILE POINT NUCLEAR STATION i

l POST-ACCIDENT SAMPLING SYSTEMS I.

INTRODUCTION A prompt and safe means of obtaining post-accident samples is neces-sary to provide plant personnel a basis for estimating the magnitude of the accident and for determining protective action recommendations to State and Local Authorities.

This study was undertaken because of NRC Emergency Preparedness Appraisal (Inspection Number 81-18) performed at the Nine Mile Point Nuclear Station. As a result of this appraisal, the NMPNS was requested to determine the maximum concentrations that could be handled and analy:cd under accident conditions.

The study is broken down into two parts.

Part A evaluates the NMPNS capability to sample reactor water, drywell air and plant effluents per the requirements of NUREG 0737.

Part B discusses the NMPNS capability to analyze the sampics collected per Part A.

II.

REFERENCES 1.

"TMI Lessons Learned Task Force Report (Short Term)," NUREG-0578, U.S. Nuclear Regulatory Commission, July 18, 1979.

2.

" Clarification of TMI Action Plan Requirements", NUREG-0737, U.S.

Nuclear Regulatory Commission, Oc ober 31,'1980.

3.

Denton, H:

" Discussion of Lessons Learned Short Term Requirements,"

U.S. Nuclear Regulatory Commission, October 30, 1979.

4.

" Radiation Sources," G.E. Document No. 22A2703R, Rev. 5, MPL No. A62-4100, 1978.

5.

Rockwell, Theodore, (Ed.):

" Reactor Shielding Design Manual," 1st Ed.,

United States Atomic Energy Commission.

6.

Hazard Summary Report for Nine Mile Point Nuclear Station - Unit #1, Appendix E.

7.

Bowers, R.

(Ed.):

"Nucicar Power Station Shiciding Manual, Volume I Gamma Shiciding," Buffalo: Niagara Mohawk Power Corporation,1965.

8.

" Final Safety and Analysis Report," Nine Mile Point Nuclear Station, U.S Atomic Energy Commission Docket 50-220 Exhibit D-2,1967.

9.

Lederer, et. al., " Table of the 1sotopes," 6th Edition, John Wiley G Sons, Inc., 1967.

,10.

Nuclear Containment Systems, 'nc., Report on In-Place Testing of Nuclear Air Cleaning Systems for Nine Mile Point Nuclear Station Unit #1, 7/22/80.

)

-~

II.

REFERENCES (Cont'd.) ~

11.

'Nucicar Energy ' Services _ Inc., "Shiciding Design Review of Nine Mile Point Nuclear Station Unit #1, Document No. 81A0636, Rev. 1, 1980.

12.

Stack Sampling, NMP-1, NMPNS Chemistry and Radiation Protection Procedure N1-SP-7, Rev. 2, October 1980.

13.

Reactor Water Sampling. Suspected High Activity, NMPNS Chemistry and Radiation Protection Procedure N1-PSP-10, Rev. O, March 1980.

14.

High Activity Drywell Atmosphere Sampling and Analysis, NMPNS Chemistry and Radiation Protection Procedure N1-PSP-11, Rev. O, March 1980.

I IS.

Interim Procedure for High Range Stack Noble Gas Release Rate Monitoring, NMPNS Chemistry and Radiation Protection Procedure N1-PSP-12, Rev. O, October, 1980.

III. METHODOLOGY i

A.

Part A - Post-Accident Sampling Evaluation In order to appropriately' determine NMPNS post-accident sampling capabilities, it was necessary to determine the total doses to be received by individual; obtaining and transporting the samples. As described in NUREG-0578 and 0737 (See References 1 6 2), individuals during post-accident l

sampling areflimited to 3 Rem whole body and 18.7S Rem to the extremities.

The assumptions and/or references used during this evaluation are listed below segregated into two areas: Source Term Calculations and Dose / Dose Rate Calculations.

1.

Source Term Assumptions

-(1) For this evaluation, the General Electric Isotopic inventory for UO2 at 30 minutes after shutdown was used (See Reference 4).

(2) The release fracti.ons assumed in this report are based on the NUREG-0578 and 0737 (See References 1 and 2) release fractions for l

i a Loss of Coolant Accident. This release fraction consisted of 100% Noble Gases, 50% Halogens and 1% core solids being released from the core inventory to the reactor plants liquid system.

(3) The inventory released was distributed into.seven energy groups to obtain the mov/sec release rate.

These seven groups consisted of 0.8, 1. 3, 1. 7, 2. 5, 4. 0, 5. 0 and 6. 4 mov. respectively.

t (4) Since emergency core cooling systca was assumed functioning, the inventory released was equally distributed throughout the l

Reactor and Torus Water Volume.

i e

'(5) of the inventory released, it 'was assumed that 100% of the noble l

gases and 25% of. the halogens were released to the drywell atmosphere i

(See references 1 and 2). This drywell air inventory was equally dis-

{.

tributed throughout the free spaces of the drywell and torus.

t i

(6) Source terms related to stack sampling considered three stack release scenarios:

a) LOCA with Icakage postul ated to the Reactor Building 30 minutes after the accident; Emergency Ventilation System running and maintaining a negative pressure in Reactor Building; Emergency Ventilation System operating at an Iodine removal efficiency of 99.99%

(see Reference 10); No other exhaust fans available for dilution, b) Same as (a), but at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident has occurred.

c) Same as (a), but with a drywell purge through the Emergency Ventilation System at 30 minutes following the accident; Iodine removal efficiency same as (a).

(7) The predominate sources assumed in the Stack and Marinelli sampling flask were mainly noble gases.

Iodines were a factor of 4

10 less than noble gases due to the emergency ventilation system removal efficiency.

(8) The predominate source assumed in the glass fiber and charcoal cartridge used for stack sample are Iodines.

Their collection efficiency per manufacturer is 99%.

2.

Dose / Dose Rate Calculations Assumptions (1) Equations used to perform dose calculations were obtained from T. Rockwell's Shiciding Design Manual (see Reference 5).

(2) All piping at sample locations except Containment Spray lines and Emergency Condenser lines were considered as line sources.

(3) Containment Spray lines and Emergency Condenser lines were con-sidered to be cylindrical sources due to their large diameters (10 and 12 inch) and because of their proximity to the dose point location.

(4)

Reactor Water and Drywell air sample vials were considered to be point sources.

(5) The stack was considered a cylindrical source 30 feet high, 21.5 feet in diameter and containing a 1.25 feet concrete wall thickness.

(6) The dose point considered for the stack was located 5 feet from l

the floor and 1 foot from the stack wall.

(7) The contributions to the total dose received from the stack below the floor was negligible due to the additional attenuation provided by l

the concrete floor.

In addition, the dose contribution from the stack j

above the 30 foot level being considered was also negligibic due to l

its angular orientation to the dose point location.

(8) The Marinelli sample flask was considered to be a cylindrical source.

l t

(9) The stack glass fiber filter and charcoal cartridge were assumed to be point sources.

B.

Part B - Post-Accident Sample Analysis Evaluation Each applicable site procedure was scrutinized to determine whether indicated processes could be accomplished under the limitations imposed by the Part A sample concentrations and dose rates.

In addition, a sup-plemental laboratory report was conducted to determine the maximum activi-ties permitting isotopic analys'3 (MAPIA). The laboratory report is con-tained as Attachment 1 to Part B and provides the basis for the results recommendation made for Part B.

IV.

RESULTS/ CONCLUSIONS A.

Part A - Post-Accident Sampling Evaluation Table 1 summarizes concentrations calculated under the three stack release assumptions, as well as the reactor water and drywell air initial activities at 30 minutes. These are the maximum concentrations expected under a LOCA condition.

Tabic 2 summarizes Drywell Air Sampling doses, for a sample drawn at 30 minutes. The conclusion to be drawn from Table 2 is that the drywell sample can be drawn, even at TMI-postulated conditons, if justified. Two teams can be used - one to set up the equipment and one to draw the sample.

This allotment of tasks would lead to approximately 2 Rem exposure to each individual. Alternatively, the current procedures may be considered ade-quate for obtaining a drywell sample at 50's of TMI postulated activity.

Procedure revisions identified in this review include:

1)

Provisions for communication should be made with this procedure.

2)

Provisions for the utilization of SR and 50R dosimeters should also be made to the procedure.

3)

Cross reference to appropriate Emergency Plan Implementing Procedure should be included.

f Table 3 summarizes doses and dose rates to be encountered at the Reactor Water Sample Sink, 30 minutes after shutdown.

Tabic 4 summarizes dose rates which could be encountered in the Emergency Condenser Isolation Valve Room in the process of opening manual sample line isolation valves.

The results of Tables 3 and 4 indicate the following:

1)

The portion of the sampling procedure performed at the Reactor Water Sample Sink on El. 261' 0" presents no exposure problem.

2)

The portion of the sampling procedure involving entry into the Emergency Condenser Isolation Valve Room presents a large ex-posure problem, especially since there exist also two emergency L

~

condenser steam lines which will contain accident sources.

Essentially, 'the dose rate from all the lines could be con-

.sidered to be' three times that of the containment spgay line.

3

-lfowever, if the accident releases were 5x10 to 1x10 times.

less' (based on the same release fraction), i.e., between 20 to 40 bCi/ml, then the overall exposures in the ECIV room would be within the above-stated limits.

3)

Remote operation of the ECIV room and Drywell Isolation Valves would alleviate potential-exposure problems related to these

samples.

Remote operation of the ECIV room and Drywell Isola-

. tion Valves is currently scheduled to be completed by 3/82, dependent on receiving remaining electrical parts.

1 4)

Provi-ion for the use of SR and 50R dosimeters, communications and asbestos _ gloves for operation of the valves in the ECIV should be made within the procedure.

5)

' Procedure needs to cross-reference appropriate Emergency Plan Implementing procedures.

Table 5 summarizes stack sampling dose rates encountered when stack sampling scenarios 1 and 2 are postulated.

4 Table 6 summarizes stack sampling capability during drywell purge.

The results of Table 5 indicate that there should be no exposure problem for stack sampling and monitoring during a LOCA if no drywell purge is required-If a drywell purge is required at 30 minutes, Table 6 indicates that na samples will be drawn during this period.

During these circumstances, the interim high range stack monitor procedure will be util-ized for gross estimates of release rates.

L

l PART A TABLE 1 ACTIVITY CONCENTRATION OF S/J41 L_E5 ITEM pCi/ml Reactor Water 5

at 30 minutes 2x10 Drywell Air 4

at 30 minutes 4x10 4

(1.4x10 Iodines)

Stack Gases, LOCA at 30 minutes, post-

~y accident 4.1x10 Charcoal 6 Glass Filter, LOCA at 30 minutes, post-

-5 accident 2.0x10 Stack Gases, LOCA at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, post-accident 4.1 Charcoal 6 Glass Filter, LOCA at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, post-

_4 accident 1.4x10 y

Stack Gases, Drywell Purge 4

at 30 minutes, post-accident 2.6x10 Charcoal 6 Glass Filter, Drywell Purge at 30 minutes, post-accident 1.2

Procedure:

N1-PSP-11 Location:

Turbine Bldg.

El. 291' 0" PART A TABLE 2 DRYWELL AIR SAMPLING AT H -0, MONITORING PANEL 2

ITEM DOSE RATE DURATION INTEGRATED DOSE R/hr MINUTES Rem Drywell Air Sample Lines 9.83(supply lines 4

0.655 (outside panel) only) 19.66(supply G 8

2.621 return - only) 15 ml Sample vial, unshicided 0.57 2

.019 15 ml Sample vial, shielded 0.037 10

.006 TOTAL DOSE 3.301 NOTES:

1.

Dose rates are calculated at 2' from the sample lines and from the sample vial.

2.

Eight minute sample time is not expected to be spent in the highest dose rate area. Actual dose rate will also include some contributions from the 11 -0 sample cabinet.

2 2

Procedure: N1-PSP-10 Location:

Reactor Bldg.

El. 261' 0" PART A TABLE 3 REACTOR WATER SAMPLING AT Tile SAMPLE SINK ITEM DOSE RATE DURATION INTEGRATED DOSE R/hr MINUTES Rem Sample Lines at 5' O.91 12 0.182 1 ml Sample Vial at S', unshicided 0.0385 4

0.0026 1 ml Sample Vial at 2', unshielded 0.0143 10 0.0024 TOTAL 0.19 5

NOTE: Reactor water activity at 30 minutes totals 2x10 pCi/ml.

Procedure: N1-PSP-10 Location:

Reactor Bldg.

El. 281' 0" PART A TABLE 4 REACTOR WATER SAMPLING AT Tile EMERGENCY CONDENSER ISOLATION VALVE ROOM ITEM DOSE RATE DURATION INTEGRATED DOSE R/hr MINUTES Rem Containment

• Spray Line at 2' 15,874 12 N/A Sample Line at 2' 70 12 N/A

• Dose rate also representative of Emergency Condenser steam lines.

l

Procedures: N1-PSP-12 N1-SP-7 Locations:

Turbine Bldg.

El. 261' 0" Screenhouse, El. 256' 0" PART A TABLE 5 STACK SAMPLING (LOCA)

ITEM DOSE RATE

• DURATION INTEGRATED DOSE mr/hr MINUTES mrem Stack Sample Lines at 2' O.41 20 0.137 Marinelli Beaker 4000 ml at 1.5' 1.56 10 0.260 Charcoal Cartridge and Glass Filter at 2' O.05 8

0.007 Stack at l' 1.57 25 0.654 TOTAL 1.058

• Dose rates correspond to sampling at 30 minutes.

For sampling at 24 hrs.,

multiply by 10.

Procedures: N1-PSP-12 N1-SP-7 Locations:

Turbine Bldg.

El. 261' 0" Screenhouse El. 256' 0" PART A TABLE 6 STACK SAMPLING (DRYWELL PURGE)

ITEM DOSE RATE DURATION INTEGRATED DOSE R/hr MINUTES Rem Stack Sample Lines at 2' 24.5 20 N/A Marinelli Beaker 4000 ml at 1.5' 93.4 10 N/A Charcoal Cartridge and Glass Filter at 2' 3.4 8

N/A Stack at l' 97.2 25 N/A

B.

Part B - Post-Accident Sample Analysis Evaluation 1.

Liquid Systems (i.e., Reactor Water, Torus)

Present methodology for sampling and diluting reactor coolant sampics during a LOCA can be found in site procedure N1-PSP-10.

Using the one-step dilution process found in the procedure, a 1 m1 reactor water sample containing a.2 Ci/ml (the activity predicted during a LOCA assuming conditiogs specified in NUREGs 0578 and 0737) can be diluted by a factor of 10 within exposure limitations of 10 CFR 20.

A 1 ml aliquot of this dilution has an activity well below the maximum activity permitting isotopic analysis (blAPIA) of 1.8 mci for small sources at 50 cm from the GeLi detector (see Attachment 1 to Part B for supplemental laboratory report).

Despite the present capability to isotopically analyze reactor water sampics during a LOCA, the following procedural amendments to N1-PSP-10 are now under investigation.

1)

Dilution of the reactor water at the sampling location.

2)

Adjust allowable sample configuration based on the values calcu-lated in this report.

3)

Calibration of the GeLi detector at greater source distances (i.e., >50 cm shelf) thereby enabling direct analysis of higher activity sampics.

These procedural amendments, together with present capabilities for sampling and analysis, should provide a sufficient degree of safety during a LOCA event until the reactor water portion of the GE post-accident sampling system is operable in h! arch 1982.

2.

Gaseous Systems (i.e., Drywell Atmosphere and Stack Gas)

The interim procedure for sampling and analysis of the drywell atmos-phere during a LOCA event can be found in N1-PSP-11.

As written, the procedure does not allow for the capability to isotopically analyze drywell atmospheres with activities of.04 Ci/ml (the activity pre-dicted during a LOCA per Part A of this report.) Gaseous dilution capa-bilities on the order of 300x are needed to meet the maximum activities permitting isotopic analysis (h!APIA) value for 50 cm distance from the detector (see Attachment 1 to Part B for supplemental laboratory report).

The isotopic analysis of stack samples during a LOCA should be possibic using site procedure N1-SP-7 with minor amendments, provided a drywell purge does not occur. Isotopic analysis of iodines and particulates at concen-trations of 2.0x 10-5 pCi/ml (the activity predicted during a LOCA per Part A of this report) can be accomplished by limiting sample volumes passed through co11cetion media. Sample volumes up to 90,000 liters (cor-responds to a 31 hr. co11cetion period) could be counted at a distance of 50 cm from the GeLi detector without exceeding 51APIA values (see Attach-1 to Part B for supplemental laboratory report).

ment

2.

Gaseous Systems (Cont'd)

Isotopic analysis of stack noble gases at a WCA concentration of 0.41pCi/ml is possibic using a Marinelli flask called for in procedure N1-SP-7.

Ilowever, because this specific geometry has not been eval-uated at 50 cm, there will be an error associated with this analysis (i.e., 2" diameter geometry vs 7.5" diameter at 50 cm.). The specific geometry is now under investigation in an attempt to cvaluate and re-duce the error associated with this analysis.

As evident in Part A of this report, dose rates near the stack during a LOCA drywell purge are prohibitive.

For this reason, isotopic analysis of stack gases under these conditions is not possible.

llowever, the installation of the SAI Stack Gas Analyzer, currently scheduled for completion by January 1983, will gliow for dilution and analysis of 3

noble gas concentrations up to 10 UCi/ml.

In an effort to further enhance isotopic analysis of gaseous systems during a LOCA, the following measures are now under consideration or planned for implementation:

1)

The installation of the GE Post-Accident Sampling System to enable us to sample and dilute the drywell atmosphere.

2)

Current sampling techniques are adequate for short term sampling (i.e., 30 minutes after accident). Ilowever, use of a smaller volume flask will be investigated for long tern sampics at later periods of an accident (i.e., 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or greater).

3)

The development of gaseous dilution techniques.

4)

Calibration of the Ge!.i detector at greater source distances and with different geometries.

PART B ATTACllMENT l' SUPPLEMENTAL LABORATORY REPORT ESTIMATED M3X1 MUM ACTIVITIES PERMITTING ISOTOPIC ANALYSIS.(MAPIA)

Methodology The radioactivity.from a 15 m1 sample of off-gas was counted for 10 minutes with a GeLi detector - MCA combination at a distance of 10 cm from the crystal.

Isotopic Analysis of the data was accomplished with a Hewlett-Packard computer equipped with APT peak search software and interfaced with a MCA. Percent dead time of the detector - MCA combination was observed prior to counting at sample distances of 0 cm, 3 cm and 10 cm.

The Maximum Acti-vities Permitting Isotopic Analysis at various distances from the detector were found by application of the inverse square law to the above data.

?he applicability of the inverse squarc law was verified by (1) ratioing detector efficiencies found at source distances of 3 cm and 50 cm and (2) showing that the efficiency ratios (efficiency at 3 cm/ Efficiency at 50 cm) either exceeded or approximated efficiency ratios estimated by the inverse square law.

Results Table #1 shows the isotopic analysis (decay uncorrected) of the 15 ml off-gas sample. The total activity of the sample was 6.6pCi.

Also shown is the percent dead time data as read on the MCA before counting commenced. Typically, sources resulting in < 20% dead time are considered identifiable, although some instrument gain adjustment. may be necessary.

Table #3 and Graph #1 shows the -results of the efficiency experiments. The ratio of the efficiencies (i.e., Efficiency at 3 cm/ efficiency at 50 cm) exceeded or approximated the efficiency ratio predicted by the inverse square law.

Conclusions By assuming that GeLi detector efficiencies change with the inverse of the square of the distance from the detector, a conservative estimate of the maximum activity permitting isotopic analysis (MAPIA) can be made.

For eg, knowing that the 6.6pCi off-gas sample resulted in 18% detgetgr dead time at 3 cm, it can be estimated that at 50 cm distance, a (50 /a ) x 6.6pci =

1.8 pCi small source would result in less than 18% dead time. Usin~g similar methodology, the following MAPIA's for small sources can be estimated:

SLR-TABLE 1

_d(cm)

MAPIA (millicuries) 50 1.80 30

.66 10

.07 3

6.6E-3

SLR-TABLE 2

% Relative Error Isotope Activity pCi/ml Energy * (Kev},

Xe-133

.007 81.08 9.1 Kr-88

.012 196.40 4.6 Xe-135

.022 250.26 1.4 Xc-138

.229 258.76 0.6 Kr-87

.019 403.14 2.1 Xe-135m

.152 526.84 1.1

( =.441 pCi/ml x 15 ml = 6.6 pCi Off-Gas Sample Distance From The Detector

% Dead Time 0 cm 45%

3 cm 18%

10 cm 5%

• For isotopes with multiple peaks, only the peak giving the lowest % relative error was considered.

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SLR-TABLE 3 Peak Energy Detector Efficiency

• Detector Efficiency *

(E)/ (A)

(kev) at 50 cm (A) (cts /v) at 3 cm (B) (cts /y) 88 4.78 E-5 2.00 E-2 628 122 4.60 E-5 2.4 E-2 520 662 1.81 E-5 5.1 E-3 283 1173 1.28 E-5 3.0 E-3 231

's 1836 2.80 E-5.-

1.9 E-3 (679,/

c 1..

• GeLi #2 calibration data compiled 2/10/81 and 2/17/81 using Standard Radio-nuclide source 121A3-09, 2" glass fiber filter.

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V

ATTACllMENT II Detection and Measurement of Airborne Iodine Under Field Conditions During Emergency Situations I.

INTRODUCTION A rapid means of detecting airborne iodine activity during an emergency is necessary to expedite the identification of the plume centerline and the recommendation of protective actions to State and Local authorities.

Valuabic time would be consumed if sampics taken out in the field had to be transported to a counting facility to determine airborne iodine activity.

During June of 1981, the NMPNS performed an evaluation of charcoal cartridges (SAI CP-100) face loaded with I-131 by Analytics, Inc. of Atlanta, Georgia.

This evaluation proved to be incon-clusive because it used I-131 as the only nuclide being counted.

As a result of the NRC Emergency Preparedness Appraisal (Inspection Number 81-18), NMPNS re-evaluated this earlier work using a mixed source of iodines in determining the iodine detection efficiency for environmental field samples.

The iodine detection efficiency determined as a result of this evaluation will be used in the field to provide ar expeditious means of evaluating an airborne release of iodines during an emergency.

II.

MET 110DOLOGY To determine an iodine detection efficiency, iodines were chemically separated for NMPNS Reactor Water and then surface loaded on SA1 CP-100 Radio-iodine Charcoal Cartridges.

Prior to loading, the Reactor Water was analyzed on the Station GeLi to verify that only nuclides of iodine were present.

The cartridges were loaded with 1, 2, 3, and 4 ml of organic solution containing the mixed iodine activity. The cartridges were allowed to dry overnight and then counted on the station GeLi the next day to determine the deposited activity on each cartridge.

Subsequent to the GeLi analysis, each cartridge was counted using an Eberline RM-14 countrate meter and a llP-210 GM probe.

Each cartridge was held approximately 1/2 inch from the GM probe and counted for a total time of 1 minute.

Background for the detcrninations was performed using a clean CP-100 charcoal car-tridge and counted in the same manner as the loaded cartridges.

Each cartridge was counted three (3) times to ensure reproduci-I bility and the data averaged for the detection efficiency deter-

mination, t

Page 2 Attachment II II. METIIODOLGY (continued) ~

The data retrieved.from the GeLi~and~RM-14/ IIP-210 analyses was-used to determine-total activity and count rate respectively-and inserted in the following equation to determine iodine-detection efficiency:

Total Count Rate (cpm) - Backgrour.1 Count. Rate (cpm) x 100 '

% officiency

=

Activity in dpc III. RESULTS The data collected is summarized in Table 1 and indicates that an efficiency of 5% would be more than conservative in estimating airborne iodine activity in the field immediately following a release.

Based.on this efficiency, Tabic 2 summarizes the Minimum Detectable Activities we would be able to' detect-in the field given a different set of variables.

In all cases MDA is well above the 1 x 10-7 uc/cc detection capability specified in NUREG-0654.

IV.

CONCLUSION During the initial days of an emergency, the shorter lived.

iodines _(I-132, 133,:134, and 135) will predominate over I-131.

Given this set of circumstances, the calculated 5% iodine detection efficiency should be more than adequate in evaluating the mixed

-iodine activity of a released plume. Subsequently, as the 1-131 becomes the predominate nuclide the detection efficiency will decrease. From previous evaluations, it could' appear that this efficiency would be a factor of 10 less than the mixed iodine efficiency.

In order to compensate for this difference in efficiency, current emergency implementing procedures assume _that all iodine activity measured in the field is due to I-131.

This over-compen-sation'during the initial moments of an emergency would ensure that-appropriate protective actions are recommended for the general public.

Subsequently, as the emergency condition continues, environmental samples would be expeditiously counted using the station GeLi to verify the isotopic mix in the sample.

With respect to noble gas interference, it is not believed that nobic gases will interfere with our field determination because:

1.

If noble gas interference is suspected (eg. high gamma exposure rate measurements) current emergency implementing procedures require the use of Silver Zeolite cartridge for the collection of iodine samples. Silver Zeolite has a reported Xenon retention cfficiency of less than 5 x 10-6 g,

Page 3 Attachment II IV.

CONCLUSION (continued) 2.

If noble gases are suspected in the counting area, current emergency implementing procedures require survey teams to retreat to a low background area (Count rate <100 cpm) for counting of the air samples.

3.

All field sampics are expeditiously returned to station for quantitative analysis on GeLi to verify field results.

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TABLE 2 Counting Statistics for use of RM-14/llP-210 to Detect Iodines on a charcoal / Silver Zeolite Cartridge in the field.

,8 A.

Given:

1.

Air Sample Volume - 15ft3 and 20ft3 2.

Cartridge retention efficiency for Iodines -

Silver Zeolite

- 95%

Charcoal (CP-100) - 99%

3.

Background - 100 cpm or less 4.

Detection Efficiency for Mixed Iodines - 5*.

5.

MDA =

MDC (6.28 x 101u dpm-cc) (ft3)(cff. of Det)(eff. of cartridge retention) uCi-ft3 B.

Data:

Count Air Vol Bkgd MDC MDA uCi/cc MDA uCi/cc Time (ft )

(cpm)

(4.664Il Silver Zeolite CP-100 3

1 min.

15 60 36 8.05 x 10-10 7,72 x 10-10 10 1 min.

20 60 36 6.03 x 10 9 5.79 x 10-10 1 min.

15 100 47 1.05 x 10-1.01 x 10-9 1 min.

20 100 47 7.88 x 10-10 7.56 x 10-10 t

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