ML20073B087

Forwards Addl Info Requested in Re NUREG-0737, Item II.B.3, Post-Accident Sampling Sys. Auxiliary Sys Not Required to Maintain Functionality of post-accident Sampling Sys

ML20073B087
Person / Time
Site:	Farley
Issue date:	04/01/1983
From:	Clayton F ALABAMA POWER CO.
To:	Varga S Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NUDOCS 8304110941
Download: ML20073B087 (19)

Text

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' Alabama Power Company 600 North 18th Street Post Office Box 2641 -

Birmingham. Alabama 35291 Telephone 205 783-6081 F. L Clayton, Jr. .

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' the southerr)CAX:frc system Ap ril 1, '1983 Docket Nos. 50-348 50-364 Di rector, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attenti on: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 NUREG-0737 Item II.B.3 Post Accident Sampling System Gentlemen:

Attached is the additional information requested by your letter of July 22, 1982 in accordance with the schedule committed to in Alabama Power Company letter dated August 18, 1982.

If you have any questions, please advise.

You rs very tru ,

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F. L. Clayton, J r.

FLCJ r/GGY: mjh-D37 Attachment cc: Mr. R. A. Thomas-Mr. G. F. Trowbridge Mr. J. P. O'Reilly Mr. - E. A. Reeves '()]f ()

Mr. W. H. Bradf ord 83041gog1 0 348 PDR A PDR ' g.

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ATTACHMENT Alabama Power Company (APCo) Responses to NRC Questions / Criteria NRC Criteria (1):

The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time alloted for sampling and analysis should be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or less from the time a decision is made to take a sample.

NRC Clarification (1):

Provide information on sampling system (s) and analytical laboratories locations including a discussion of relative elencions, distances and methods for sample transport. Responses to this item should also include a '

discussion of sample recirculation, sample handling and analytical times to demonstrate that the three-hour time limit will be met (see (6) below relative to radiation exposure). Also describe provisions for sampling during loss of off-site power (i.e. designate an alternative backup power source, not necessarily the vital (Class 1E) bus, that can be energized in suf ficient time to meet the three-hour sampling and analysis time limit).

APCo Response (1):

The Sample Room, Counting Room and Radiochemistry Lab for the Farley Nuclear Plant (FNP) Post Accident Sampling System are located on the 139' elevation of the Auxiliary Building (Figure 1). A pressurized Reactor Coolant Sample (RCS) can be promptly obtained from the remote sample station located on the 139' elevation of the Auxiliary Building. The pressurized sample is circulated through a 40 cc stainless steel sample bomb which purges any resident water and ensures that the resultant sample is representative of the coolant system (Figure 2). The sample is degassed and an aliquot of the gas is drawn off by microsyringe and injected into a gas chromatograph for analysis. The liquid portion of the sample is injected into a 50 cc shielded sample vial located in the sample room (Figure 3).

The sample is then transported to the lab for analysis by means of a shielded sample transport cart which has been designed to minimize any radiation exposure.

In case of containment isolation prior to a sample request, approximately 30 minutes will be required to reestablish flow and obtain a representative sample at system pressure (system pressure [2200 psig])

provides the driving force for sample flow). If RHR has been initiated, an additional 15 minutes may be necessary due to the reduced system pressure.

Sample preparation will then require 30 minutes to one hour, dependent upon the availability of the unaf fected Unit's lab. Chemical and isotopic analyses will require approximately 30 minutes.

Containment atmosphere samples can be promptly obtained by means of a remotely activated inline air sampler parallel with RE-11/12 (Containment Leak Monitor) on the 121' elevation of the Auxiliary Building. The inline air sampler draws air through a 250 cc gas samgle bomb, a particulate filter and an iodine cartridge, from the 134'6" elevation of containment. The hydrogen, oxygen and noble gas samples are collected by microsyringe through

Attachment Page 2-a septum located.on the 250 cc gas sampl'e bomb. The particulate and iodine.

filters are collected by removing the sample holder by means of a' self-sealing quick disconnect fitting (Figure 4). - Collection and analyses of r containment. atmosphere samples require less than one hour f rom the time the sample is requested.

The Unit 1- and Unit 2 reactor .oolant sampling systems and the Unit 1-containment atmosphere sampling ' system are powered f rom Class 1E load i centers. The Unit 2 containment atmosphere sampling panel is powered from onsite AC load center and could be manually connected to a Class 1E power source during a loss of off-site power event. Alabama Power Company has

determined that the Unit 1 and Unit 2 sampling systems can be energized, following a loss of off-site power, in sufficient time to meet the

three-hour sampling and analysis time limit.

NRC Criteria (2):

} The licensee shall establish an onsite radiological and chemical j analysis capability to provide, within the 3-hour time frame established

above, quantification of the following

} a) certain radionuclides in the reactor coolant and containment j

~

atmopshere that may be indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and non-volatile isotopes);

j b) hydrogen levels in the containment atmosphere; c) dissolved gases (e.g. 2H ), chloride (time allotted for analysis subject to discussion below), and boron concentration of liquids.

d) Alternatively, have inline monitoring capabilities to perform all

or part of the above analyses.

I

a NRC Clarification (2):

2(a) A discussion of the counting equ'ipment capabilities is needed, including provisions to handle samples and reduce background radiation to minimize personnel radiation exposures (ALARA). Also, a procedure is required for relating radionuclide concentrations to core damage. The-procedure should include:

1) Monitoring for short and long lived volatile and non-volatile radionuclides such as 133Xe,1311,137Cs,134Cs, 85Kr,140Ba, and :

88Kr (See Vol. II, Part 2, pp. 524-527 of Rogovin Report.for further information).

2) Provisions to estimate the extent of core damage based on radionuclide concentrations and taking .into consideration other physical parameters such as corr temperature data and sample location.

2(b) Show a capability to obtain a grab' sample, transport and analyze.

for hydrogen.

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L j 2(c) Discuss the capabilities to sample and analyze forlthe. accident

sample species listed here and:in Regulatory Guide 1.97, Rev.l 2.. .

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2(d):Prov'ide:a discussion of the reliability and maintenance ;infor .

mation to demonstrate that the selected on-line; instrument is appropriate-

[ forLthis application. .(See '(8) and ,(10)-below relative to back-up grab

- sample capability and instrument range and accuracy).

i l APCo Response (2): ,

i i p - 2(a) Each unit at Farley Nuclear Plant has a radiochemistry laboratory

! (Figure 1) with radiological and chemical analysis capabilities (that -- -

i supports three hour quantification of analysis) to dilute and aliquot

samples that are extremely radioactive (i.e. up to 10. C1/gm). Wet chemical' i analysis and separations are peformed in each laboratory under shielded F conditions to keep personnel exposure as

low as r.easonably achievable.  ;

(ALARA). In addition, each laboratory is equipped for gas chromatography.

j The FNP Emergency Operations Facility (E0F), located one~-fourth mile southwest. of the plant, has a radiochemistry lab equipped with a charcoal ,

l filter hood, a hold-up tank and drumming facilities for chemical and radio-j active waste material. Each unit has a counting room' equippe'd with solid 4 state detectors coupled to multichannel analyzers, gross alpha and beta- 't gamma gas proportional counters, liquid scintillator and adequate' computer j access to reduce data and provide detailed printed reports of analyses.

Detectors are shielded and compensated for reduction of Compton Scattering.

l The E0F is equipped with a counting room with solid state detectors, a I multichannel analyzer, gross alpha and beta-gamma gas flow proportional, counters and computer access to provide analytical-analysis which-' equals that of the plant counting rooms.

4 Radionuclides in the reactor coolant and containment atmosphere which s may be indicators of the degree of core damage (e.g., noble. gases, iodines

, and cesiums, and short and long lived volatile and nonvolatile isotopes) can.  ;

, be determined by diluting an aliquot of the RCS sample or containment;1 '_ ,

1 i

atmosphere to a point where analysis can be performed with less than 10%  ;

dead time on a multichannel analyzer. The dilution and isotopic analysis .

can be -done in the radiochemistry lab of the unaf fected unit if it _is accessible or in the E0F radiochemistry lab _if the unaffected unit is ,

inaccessible. Whenever an iodine filter cartridge or particulate filter -is

too active to analyze, a sample of less volume can be collected or~ the .

sample can.be divided'into smaller portions and analyzed acccedingly. ; Lead '

, brick- shielding and; lead glass' window's/ provide :shieldingiduring ' sample e preparation and chemical . analysis to keep personnel' exposure ALARA.-

Procedures which relate radionuclide concentrations to core: damage are =

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contained in the Farley' Nuclear Plantiprocedures .which consider physical .

, parameters:such -as core' temperature _and (sample . location.- The Farley Nuclear :

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F Plant procedures have been reviewed and. accepted by the .NRC as documented in-

! Supplement No.= 5 to NUREG-0117. _ Additionally, Alabama Power. Company is =

L working with the Westinghouse:0wners Group to determine;if improved -

procedures can be developed to assess the extent' of core damage., ,

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.Page-4 2(b) Hydrogen levels in the containment atmosphere are normally measured by inline hydrogen analyzers. Remote madout from the hydrogen analyzers .is provided in the control ' room.. The procedures to obtain a grab sample, transport and analyze for hydrogen are the same as described in the APCo response to NRC Criteria (1).

2(c) Dissolved gases from the RCS are collected and analyzed as previously discussed in. response to criteria one and two. This analysis i includes percent of hydrogen in the containment atmosphere, cc/kg hydrogen in the RCS and pCi/cc of short and long lived volatile radionuclides.

Liquid samples of non-volatile radionuclides are analyzed as described in response to criteria two.

2(d) Tne only inline instruments in the FNP post accident sampling system are containment hydrogen analyzers. Each unit has two analyzers that read out in the corresponding control room which are functionally tested and calibrated quarterly on a range of zero to one percent and zero to ten percent full scale.

NRC Criteria (3):

Reactor coolant and containment atmosphere sampling during post accident conditions shall not mquire an isolated auxiliary system [e.g.,

the letdown system, reactor water cleanup system (RWCUS)] to be placed in operation in order to use the sampling system.

letC Clarification (3):

System schematics and discussions should clearly demonstrate that post accident sampling, including recirculation, from each sample source is possible without use of an isolated auxiliary system. It should be verified that valves which are not accessible after an accident are environmentally qualified for the conditions in which they must operate.

APCo Response (3):

No auxiliary system is required to maintain functionality of the ~FNP post accident sampling system. Sample cooling is provided by the component cooling water (CCW) system. Design changes which allow use of the CCW l without resetting the safety injection signal are scheduled to be implemented during the second refueling outage for Unit- 2 (scheduled for the.

l fourth quarter of 1983), and the fifth refueling outage for Unit 1 l

. (scheduled for the first quarter of 1984). Controls for all valves necessary .to obtain such samples are accessible during post accident conditions.

NRC Criteria (4):-

l Pressurized reactor coolant samples am not required if the licensee can quantify the amount of dissolved gases.with unpressurized reactor l coolant samples.- The measurement of either total dissolved gases -or I

hydrogen gas in reactor coolant samples is considered adequate. Measu ri ng the oxygen concentration is recommended, but. is not mandatory. '

Attachment i Page 5 NRC Clarification (4):

Discuss the method whereby total dissolved gas'or hydrogen and oxygen can be measured and related to reactor coolant system concentrations.

Additionally, if chlorides exceed 0.15 ppm, verification that dissolved ozygen is less than 0.1 ppm is necessary. Verification that dissolved oxygen is less than 0.1 ppm by measurement of a dissolved hydrogen residual of greater than or equal to 10 cc/kg is acceptable for up to 30 days after the accident. Within 30 days, consistent with minimizing personnel radiation exposure (ALARA), direct monitoring for dissolved oxygen is recommended.

APCo Response (4):

Both hydrogen and oxygen are degassed and analyzed as described in APCo response to Criteria (1) above. As previously discussed in APCo letter dated January 14, 1981 to the NRC, the onsite liquid analysis program can provide results within a factor of two error. Additional information concerning chloride and oxygen is contained in APCo response to Criteria (10).

NRC Criteria (5):

The time for a chloride analysis to be performed is dependent upon two f actors: (a) if the plant's coolant water is seawater, or brackish water and (b) if there is only a single barrier between primary containment systems and the cooling water. Under both of the above conditions the licensee shall provide for a chloride analysis within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sample being

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taken. For all other cases, the licensee shall provide for the analysis to be completed within 4 days. The chloride analysis does not have to be done onsite.

NRC Clarification (5):

BWR's on sea or brackish water sites, and plants which use sea or brackish water in essential heat exchangers (e.g. shutdown cooling) that i have only single barrier protection between the reactor coolant are required to analyze chloride within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. All other plants have 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to perform a chloride analysis. Samples diluted by up to a factor of one

, thousand are acceptable as initial scoping analysis for chloride, provided .

l (1) the results are reported as ppm C1 -(the licensee should establish this value: the number in the blank should be no greater than 10.0 ppm Cl) in the reactor coolant system and (2) that dissolved oxygen can be verified at less than 0.1 ppm, consistent with the guidelines above in clarification (4). Additionally, if chloride analysis is performed on a

-diluted sample, an undiluted sample need also be taken and retained for analysis within 30 days, consistent with ALARA.

APCo Response (5):

Farley Nuclear Plant's ' service. water comes f rom the Chattahoochee River which is a f resh water source. The service water is used to cool the

Attachment Page 6 Component Cooling Water (CCW) which is chromated to prevent oxidation of components. The CCW system then cools the primary system _ components providing a double barrier between contaminated systens and the service wate r. Chloride analysis for the RCS is accomplished by ion specific -

electrodes in liquid samples using procedures where dilution is not-4 necessary except in worst case conditions. Undiluted samples of chloride are retained for later analysis if deemed necessary.

Information pertaining to the initial scoping analysis of chloride and oxygen is presented in APCo response to Criteria (10). Alabama Power stated in letter dated February 9,1981 that chloride analysis can be performed within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.

NRC Criteria (6):

The design basis for plant equipment for reactor coolant and contain-ment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposures to any individual exceeding the criteria of GDC 19 (Appendix A,10 CFR Part 50) (i.e., 5 rem whole body, 75 rem extremities). Note that the design and operational review criterion was changed from the operational limits of 10 CFR Part 20 (NUREG 0578) to the GDC 19 criterion (October 30, 1979 letter f rom H. R.

Denton to all licensees).

NRC Clarification (6):

Consistent with Regulatory Guide 1.3 or 1.4 source terms, provide information on the predicted personnel exposures based on person-motion for sampling, transport and analysis of all required parameters.

APCo Response (6):

The design basis for reactor coolant sampling and containment atmos-phere sampling and analysis have provided for sampling under conditions of-TID-14844 source terms without exceeding GDC 19. Tine and motion studies have been conducted which conservatively estimate that a ' reactor coolant

! sample 'for hydrogen, ozygen, noble gases, iodines, cesiums, boron, chloride and non-volatile isotopes nay be collected without exposing any personnel to more than 350 millirem whole body dose or measurable internal . intake. Due to remote handling of the samples, the dose to the extremities of any personnel will be the same as the whole body dose. Containnert atmosphere samples can be collected without exposing any personnel .to more than 250 millirem whole body and 300 millirem' extremity dose. Sample preparation and analysis may be performed without exposing any personnel to more than 200 l millirem whole body or 300 millirem extremity dose. Shielding of the samples during transport and mobility of the transport device prevents personnel from receiving any significant radiation doses during transport of the sanples from the collection area to the laboratory. These dose rates were based on worst case radiation zone values assuming a worst case accident.

l 1

Attachment Page 7 NRC Criteria (7):

The analysis of primary coolant samples for boron is required.for PWRs. (Note that Rev. 2 of Regulatory Guide 1.97 specifies the need for primary coolant boron analysis capability at BWR plants).

NRC Clarification (7):

PWR's need to perform boron analysis. The guidelines for BWR's are to have the capability to perform boron analysis but they do not have to do 'so unless boron we injected.

APCo Response (7):

Boron analysis of primary coolent samples is performed on. undiluted liquid samples according to appropriate plant procedures which account for the possibility of Na0H in the sample due to containment spray.

NRC Criteria (8):

If inline monitoring is used for any sampling and analytical capability specified herein, the licensee shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing the samples.

Established planning for analysis at offsite facilities is acceptable.

Equipment provided for backup sampling shall be capable of providing at least one sample per day for 7 days following onset of the accident, and at-least one sample per week until the accident condition no longer exists.

NRC Clarification (8):

A capability to obtain both diluted and undiluted backup samples is requi red. Provisions to flush inline monitors to facilitate access for repair is desirable. If an off-site laboratory is to be relied on for the backup analysis, an explanation of the capability to ship and obtain analysis for one sample per week thereafter until accident condition no longer exists should be provided.

APCo Response (8):

No inline sampling and analysis is conducted at Farley Nuclear Plant for any parameter except hydrogen in containment atmosphere. This is described in APCo response to Criteria 2(b).

NRC Criteria (9):

The licensee's radiological and chemical sample analysis capability shall include provisions to:

(a) Identify and quantify the isotopes of the nuclide categories discussed above to levels corresponding to the source terird given in Regulatory Guide 1.3 or 1.4 and l'.7. Where necessary and

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l Attachment Page 8 i ~ practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel . exposure should be . i i _ provided. Sensitivity _of onsite liquid sample analysis. capability

- should be such as. to permit measurement ~ of .nuclide concentration in the range from approximately 1 micro Ci/g to 10 Ci/g.

(b) Restrict background ' levels ofL radiation in the radiological and-

!- chemical analysis facility from sources -such'that the sample ,

analysis.will provide msults with an: acceptably small error -'

4-(approximately a factor of 2).: 'This can be accomplished through-the use of sufficient shielding around samples and 7outside.

sources, and by the use of' a ventilation system design which will

control the presence of--airborne radioactivity.

j istC Clarification (9):  !

9(a) Provide a discussion of the predicted activity in the samples to .

be taken and the methods of handling / dilution that will be employed to reduce the activity sufficiently to perform the required analysis. Discuss

the range of radionuclide concentration which 'can be analyzed for, including

! an assessment of the amount of overlap between post accident- and normal-i sampling capabilities.

I 9(b) State the predicted background radiation levels in the counting room, including the contribution from samples which are present. =Also provide data demonstrating what the background radiation levels and ,

i radiation effect will be on a sample being counted to assure an accuracy

within a factor of 2.

F APCo Response (9):

l 9(a) The FNP post accident sampling system is the: normal pathway for ,

routine sampling and will accomodate activity levels up to-10 C1/gm. Thel j shielded sample vial (Figure 3) is not normally used for routine sampling;-  :

1 however, it is available for sampling whenever deemed necessary by Health Physics surveys. Dilution of the samples is accomplished by pipetting one ml of the sample into a beaker of 999 ml of demineralized water and Na0H l located behind a lead shield and lead glass window. _ If'necessary, a'second

dilution may be performed as described above to further redyce the sample l acti vity. The second. dilution results in a reduction ~of 100 in activity.

As the FNP analytical equipment has a ' range of SE-7 to 10 micro Ci/gm,,

samples with -an activity approaching ~ 10 micro C1/gm must be. diluted as j' described above-before an analysis can be performed.

Activity .in filter. samples can.be mduced by sampling less volume of.

..b the filter or- removing the used filter media. (i.e., charcoal granuals) and:

mixing with clean media for dilution. Alternatively, a sample of less -

volume can be collected as described in APCo response to Criteria 2(a).

1, 9(b)' The background activity levels for any isotope being: analyzed is-

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L .less than 0.1% of the. activity 1n the sample. The levelsl are ' determined by -

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.c Attachment Page 9 collecting a spectrum with no sample and the detector shield closed. The spectrum is then analyzed using the geometry and volume intended for the sample. Whenever the results of the background analysis indicate isotopes are present in quantities greater than 0.1% of the quantity measured in the sample, the sample will be moved to an alternate laboratory and another

, analysis performed. -Each sample will be removed f rom the counting room before another sample is brought in-for analysis to prevent any background effects in the counting room area. With background isotopes less than 0.1%

an accuracy within a factor of 2 is ensured.

NRC Criteria (10):

?

Accuracy, range and sensitivity shall be adequate to provide pertinent data to the operator in order to describe radiological and chemical status of the reactor coolant systems.

Clarification (10):

The recommended ranges for the required accident sample analyses are given in Regulatory Guide 1.97, Rev. 2. The necessary accuracy within the recommended ranges are as follows:

a) Gross activity, gamma spect rum: measured to estimate core damage, these analyses should be accurate within a factor of two across the enti re range.

b) Boron: measured to verify shutdown margin. In general this analysis should be accurate within i 5% of the measured value (i.e., at 6,000 ppm B the tolerance is + 300 ppm while at 1,000 ppm B the tolerance is t 50 ppm). For concentrations below 1,000 1

ppm the tolerance band should remain at i 50 ppm.

c) Chloride: measured to determine coolant corrosion potential.

For concentrations between 0.5 and 20.0 ppm chloride the analysis should be accurate within i 10% of the measured value. At

+

concentrations below 0.5 ppm the tolerance. band remains at t~ 0.05 ppm.

d) Hydrogen or Total Gas: monitored to estimate core degradation and corrosion potential of the coolant. An accuracy of i 10% is desirable between 50 and 2000 cc/kg but t 20% can be acceptable.

For concentrations below 50 cc/kg the tolerance remains at i 5.0 cc/kg.

e) 0xygen: monitored to assess coolant corrosion potential. For concentrations between 0.5 and 20.0 ppm oxygen the analysis should .

be accurate within i 10% of the measured value. At concentrations below 0.5 ppm the tolerance band remains at t 0.05 ppm.

Attachment Page 10 4

f) pH: measured to access coolant corrosion potential. Between a pH of 5 to 9, the reading should be accurate within + 0.3 pH units.

For all other ranges t 0.5 pH units is acceptable.

g) To demonstrate that the selected procedures and instrumentation will achieve the above listed accuracies, it is necessary to provide information demonstrating their applicability in the post accident water chemistry and radiation environment.- This can be accomplished by performing tests utilizing the standard test matrix provided below or by providing evidence that the selected procedure or instrument has been used successfully in a similiar .

envi ronment.

Standard Test Matrix for -

Undiluted Reactor Coolant Samples in a Post Accident Environment Constituient, Nominal Concentration (ppm) Added as (chemical salt)

I- 40 Potassium Iodide Cs+ 250 Cesium Nitrate Ba+2 10 Barium Nitrate La+3 5 Lanthanum Chloride Ce+4 5 Ammonium Cerium Nitrate Cl- 10 B 2000 Boric Acid Li+ 2 Lithium Hydroxide NO3 150 NH4 + 5 K+ 20 Gamma Radiation 104 Rad /gm of Absorbed Dose (Induced Field) Reactor Coolant Notes

1) Instrumentation and procedures which are applicable to diluted samples only, should be tested with an equally diluted chemical test matrix. The induced radiation environment should be adjusted commensurate with the weight of actual reactor' coolant in the -

sample being tested.

2) For PWRs, procedures 'which nay be affected by spray additive chemicals must be tested 'in both the standard test matrix plus appropriate spray additives. Both procedures (with and without spray additives) are' requi red to be available.

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Attachment j' .Page 11 i

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_ 3) For BWRs, if procedures are verified with boron in the test j  ; matrix, they _do not ~have to be tested without boron.

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4) In lieu of conducting tests utilizing the standard test' matrix.fori instruments and procedures, provide evidence that- the selected:

instrument or- procedure has -been used successfully-in a f similar envi ronment.

l All equipment' and procedures which are used for- post- accident sampling s

i and analyses should be calibrated -or tested at a f requency which will ensure, to a high degree of reliability, that:it will-be available if~

requi red. Operators should receive initial ~ and ref resher training in i post accident sampling, analysis and transport. A minimum frequency for the j above efforts is considered to be every six months if 1.ndicated by _ testing.

These provisions should be submitted in mvised Technical Specifications:in l- accordance with Enclosure 1 of NUREG-0737. The staff will providef model: ,

t Technical Specifications at a later date.

i APCo Response (10):

l j a) Gross Activity Gamma Spectrum b The error for dilution of a high activity _ sample is maintained 3 at less than i 10%. Analytical accuracy which is adjusted by count ~

i rate and count time-is within i 25%. Worst case additive error is

maintained at less than or equal to i 45% which is within a factor _ of.

j' two across the enti re range.

b) Boron Analysis i Boron analysis at Farley Nuclear Plant is normally maintained with

! _ an accuracy of i 10 ppm at 1000 ppm. Boron ' dilution is mquired 1

whenever boron content is greater than 7000 ppm which_ creates an . . ..

additional error of i 10% per 1000 ppm. Boron. dilution is 'not .mquired 1' for radiological concerns. FNP equipment allows analysis of. undiluted i samples containing up to 10 C1/gm.

L

j. c) Chloride Analysis i

j Chloride analysis.at FNP is performed using the ion specifici

electrode method with a . range of 0.010 ppm _to 35,000 ppm.D Analysis

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' accuracy _ is Li 5.8% at 0.020 ppm. No dilution is :necessary to' perform

chlo' ride analyses. .

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d)- Hydro' gen or-Total Gas Analysis Hydrogen-analysis within thef reactor-coolant sytem at- FNPiis maintained at .i 10% accuracy:for 100 cc/kg' and above. A linear L

relationship -is established ~ using~. reference . gases of 90% and' 2%

hydrogen by volume,- which corresponds to approximately'.4365 cc/kg-and 100 cc/kg respectively. -For hydrogen analysis below 100 'cc/kg, the' i;

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Attachment Page 12 -

! concentration is established by comparing the unknown value.with the -

4 extrapolated linear reiationship such that an accuracy within +; 20% is -

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2- maintained for concentrations between 100 cc/kg and 50 cc/kg, and an~

accuracy within 5 cc/kg for concentrations below 50 cc/kg.

i' e) 0xygen Analysis d

Post accident dissolved oxygen analysis at FNP is performed by

injecting an aliquot of the gases' removed from the liquid reactor ' -

l coolant sample into a gas chromatograph. This 'results: in ~a lower.

, sensitivity of 5 to 7 ppm, an upper bound equivalent to air saturated

water (18 - 20 ppm),1and an accuracy of- 10%.

f 1-f) pH Analysis j pH Analysis at FNP is performed with

• 0.05 pH units per pH

!: increment ever the buffered span (i.e., if the pH is buffered at f4 and

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! 10, the accumulated error will ' be 0.3 pHf units). .

i g) Standard Test Matrix d

l The criteria for testing equipment in a gamma induced. field 'of 104 i Rad /gm of reactor coolant has not been performed in the past at FNP.

i Alabama Power Company is confident that existing equipment and procedures would provide satisfactory results during post accide'nt j analysis and the analysis of such test matrix would only be for;

confi rmatory purposes. Such confirmation would be incompatible with ,

] the Farley ALARA program. In addition, since the Farley analytical j equipment and practices reflect current industry standards, Alabans' Power Company does not feel that there is sufficient justification for .

i 4

performing such confirmation testing.

NRC Criteria (11): H r

i In the design of the post accident sampling and analysis capability, consideration should be given to the following items:

j a) Provisions for purging sample lines, for reducing plateout in

{ sample lines, for minimizing sample loss or distortion, for i preventing blockage of sample lines by loose material in the RCS 1

or containment, for-appropriate disposal of the samples,- and for flow restriction to limit reactor coolant . loss from a: rupture of the sample'line. The post accident reactor coolant ~ and containment atmosphere samples should be representative of the ~

reactor coolant in the core area 'and the containment atmosphere following a transient or accident. -. :The sample lines shouldJbe as j: short'as possible to minimize the volume' of. fluid to be taken f rom-containment. . The residues of sample, collection should be' returned to containment sor to a closed sytem.-

} . .

4-

' b) The ventilation exhaust f rom the sampling station should be filtered with charcoal absorbers and high-efficiency particulate air -(HEPA) filters.

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Attachment i Page 13 NRC Clarification (11): -

11(a) A description of the provisions which address each of the items in Criteria 11.a should be provided. Such items as heat tracing and purge velocities should be addressed. To demonstrate that samples are representative of core conditions, a discussion of mixing, both short and long term, is needed. If a given sample location can be rendered inaccurate due to the accident-(i.e., sampling from a hot or cold leg loop which nay-have a steam or gas pocket) describe the backup sampling capabilities or

, address the maximum time that this condition can exist.

BWRs should specifically address samples which are taken from the core shroud area and demonstrate how they are representative of core conditions.-

Passive flow restrictors in the sample lines any be replaced by redundant, environmentally qualified, remotely operated isolation valves to limit potential leakage f rom sampling lines. The automatic containment isolation valves should close on containment isolation or safety injection signals.

11(b) A dedicated sample station filtration system is not required, provided a positive exhaust exists which is subsequently routed through charcoal absorbers and HEPA filters.

APCo Response (11):

11(a) The FNP post accident sampling system is a continuous use system

, with a flow rate of 0.6 gallons per hour. Sample lines are 3/8 inch stainless steel tubing which provide a delay of 1.2 minutes of sample flow and 10 half lives of N-16. Plateout in the sample lines is minimized by a linear flow velocity of approximately 132 feet per minute. Large radius bends and short piping has been utilized in the design of the system to minimize the volume of flow from containment. Redundant sample lines f rom the RHR pumps is provided to ensure a representative sample can be obtained if blockage of one of the sample lines occur. A flow monitor on the gross failed fuel detector and periodic switching of the sample lines assure unrestricted flow through the lines. Loss of coolant f rom a rupture of the lines is restricted to 0.6 gallons per minute which, along with any spillage or sample residue, is collected in the waste hold up tank. Sample volume is-limited to 50 cc to prevent any excessive sample spillage. . Sample residue is drummed for storage or processed in the waste processing demineralizer.

All lines and valves in the penetration room and containment have been qualified and tested to meet NRC requirements. Two hydrogen dilution fans and four air cooler fans are utilized to obtain a representative sample of containnent atmosphere.

During an accident which has been deemed not serious by Health Physics surveys, samples are taken from the vent stack and no heat tracing is required due to the proximity of the sample location. ~During accident conditions which have been deemed serious by HealthL Physics surveys, samples are remotely taken (RE 298) and heat tracing'is utilized.

l

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'j Attachment i Page 14 I j~' - ll(b) The ventiliation exhaust from the sampling station is routed to

.the gaseous radwasta handling' system which is equipped with HEPA- filters and -

charcoal absorbers.

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