ML17301A024

Forwards Info Demonstrating Capability of post-accident Sampling Sys,Closing Out SER Open Items. Interim Procedure Guideline for Core Damage Assessment Also Encl

ML17301A024
Person / Time
Site:	Saint Lucie
Issue date:	03/04/1983
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML17213B100	List: ML17213B099 ML17301A024
References
RTR-NUREG-0737, RTR-NUREG-737 L-83-122, NUDOCS 8303090340
Download: ML17301A024 (39)

Text

RE.DLSTORvP.R.ATION OISTRIBOTION STN IRIDS>

ACCESSION  %<BR i 8303090300 DOC ~ DATE: 83/03/00 NOTARIZED; NO DOCKET FAOIL;50-389 St, Lucie PlantE Uni't 2E Florida, Power L Light Co. 05000389 AUTH ~ NAME AUTHOR AF F ILI ATION UHRIGg R,E ~ Flor,ida Power 8 Light Co,

.RECIP,NAIVE RECIPIENT AFFILIATION EISENHUTEDBG. Division of Licensing

SUBJECT:

Forwards info demonstrating capability of

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sampling sys closing out SER open i.tems.."Interim Procedure Guideline for Core Damage Assessment" also encl, DISTRIBUTION CODE: BOO IS COPIES RECEIVED:LTR .

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TITLE! Licensing Submittal: pSAR/FSAR Amdts 8 Related Correspondence'OTES:

REC IP IENT COPIES RECIPIENT COPIES ID CODE/NAMF LTTR ENCL ID CODE/NAME LTTR ENCL A/O LICENSNG 1 0 LIC BR 43 BC 1 0 LIC BR 03 LA 1 0 NERSESEV ~ 01 1 1 INTERNAL! ELO/HDS2 1 0 IE FILE 1 1 IE/OEP EPOS 35 1 1 IE/DEP/EPLB 36 3 NRR/OF/AEAB 1 0 NRR/OE/CEB 11 1 1 NRR/OE/EQB 13. 2 NRR/DE/CB 28 2 2 NRA/OE/HGEB 30 1 NRR/DE/MEB 18 1 1 NRR/OE/MTEB'7 1 1 NRR/DE/QAB 21 1 1 NRR/OE/SAB 20 1 NRR/DE/SEB 25 1 1 BIRR/OHFS/HFE000 1 1 NRR/DHFS/LQB 32 1 1 NRR/DL/SSPB 0 NRR/DS I/AEB 26 1 1 NRA/Ds I/ASB NRR/DS I/CPS 10 1' NRR/OSI/CSB 09 NRR/DSI/METB 12 1

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AB 2? 1 NRR/DSI/RSB 23 1 G F IL 04 1 1 RGN2 3 R /MIB 1 0 EXTFRNAL ~ ACRS Ql 6 6 BNL(AMOTS ONLY) 1 1 DMB/DSS {AMDTS) 1 1 FEMA"REP DIV 39 1 1 LPDR 03 1 1 NRC PDR 02 1 1 NSIC 05 1 NTIS 1 1 TOTAL NUMBER OF COPIES REQUIRED. LTTR 52 ENCL 45

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. BOX 14000, JUNO BEACH, FL 33408 FLORIDA POWER 8( LIGHT COMPANY March 4, 1983 L-83-122 Office of Nuclear Reactor Regulations Attention: Mr. Darrell G. Eisenhut, Director Division of Licensing U. S. Nuclear Regulatory commission Washington, D.C. 20055

Dear Mr. Eisenhut:

Re: St. Lu"ie Unit No. 2 Docket No. 50-389 Post Accident Sam lin S stem Ca abilit In Supplement 2 of the SER, the staff requested the applicant, by 5%

power, to demonstrate the capability of the Post Accident Sampling System to:

1) Provide for a chloride analysis within 4 days after the reactor coolant sample is taken;

2) Provide the capability to identify and quantify the activity for reactor coolant and containment atmosphere post accident samples;

3) Provide a procedure for relating radionuclide gaseous and ionic species to estimate core damage.

The attached information is provided to close out the above open items.

This information has been informally given to your staff for review and found to be acceptable. In addition, the present St. Lucie Unit 2 chemi" cal procedures have incorporated the methodology for estimating core damage per the description provided in item 3.

If you have any questions regarding this submittal, please contact us.

Very truly yours, Robert E. Uhrig Vice President Advanced Systems & Technology REU/RJS/cab 8303090340 830304 PDR ADOCK 05000389 PDR PEOPLE... SERVING PEOPLE

8303090340 POST-ACCIDENT SAMPLING SYSTEM SuPPLEMENTAL INFORMATION

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TABLE OF CONTENTS

~Pa e Criterion 1 1 Criterion 2 3 Criterion 3 7 Cri terion 4 7 Criterion 5 10 Criterion 6 ll Criterion 7 13 Criterion 8 14 Criterion 9 15 Criterion 10 17 Criterion ll 19 Interim Core Damage Procedure

Criterion:

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~ (1) The licensee shall have the capability to promptly obtain the reactor coolant samples and containment atmosphere samples.

The combined time allotted, 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.

Response: The attached Pert chart presents a conservative estimate of time required for one exerienced operator to perform PASS recirculation, analysis, and dilution procedures as described in the operating procedures of the PASS Technical Hanual. The times shown in the chart have been verified through field operation of a similar C-E PASS at the San Onofre Unit 2 plant. The PASS is described in Section 9.3.6 of the FSAR. The PASS Laboratory is located in the Reactor Auxiliary Building (RAB) on the ground level 19.50'levation. The Radiochemistry Sample Room is located in the RAB on the same elevation as the PASS Laboratory. The Sample Room and the PASS Laboratory are on the same hallway approximately 40 feet apart.

The PASS performs the chemistry analysis for boron, pH, total dissolved gas, hydrogen and oxygen offgas, and dissolved oxygen using on-line instrumentation with backup grab sampling capability.

Analysis of chlorides and radiological fission product analysis is performed through grab samples. These samples are analyzed in the Radiochemistry Sample Room., Grab sample transport to the Sample Room is performed by two means.

The first means of transport is the Grab Sample Facility. The Grab Sample Facility provides the capability to collect an undiluted reactor coolant liquid grab sample. The Grab Sampling Facility consists of an undiluted depressurized liquid sample vessel enclosed in a lead shielded cask, isolation and bypass valves with stem extensions penetrating the shielding for connection to the PASS,

a pallet lift hand truck, and double end shut-off quick disconnects interface connections to the PASS sample station liquid

'or sample return line. Sampling is performed as follows:

The hand truck containing the cask .is maneuvered to the PASS sample station liquid sample return line tees and the assembly is connected to the system by means of the double ended shut-off quick disconnects. Once connected, the grab sampling assembly is flushed with demineralized water via the PASS sample station connection. The liquid sample path pressure indicator (P-503) can be used to check for leaks before any contaminated fluid's allowed to enter. Following satisfactory pressurization, an undiluted sample is purged through the PASS. When a representa-tive reactor coolant sample is available, the sample vessel and connection tee isolation valves are opened to allow flow through the Undiluted Liquid Sample Vessel and the Grab Sampling Facility bypass valve is closed. When purging of the sample vessel is complete, the sample is isolated by opening the bypass valve and closing the isolation valves. Sample purge flow through the PASS is secured. Prior to disconnecting the cask, a demineralized water flush is accomplished by establishing flow via the PASS sample station connection through the normal liquid sample return line path. The connection tee isolation valves are opened and the Grab Sampling Facility bypass valve is closed, allowing a de-water flush of all tubing except for the sample vessel

'ineralized and isolation valves. This minimizes radiation exposure and the possibility fo contamination when the cask is disconnected. The cask is disconnected from the system by means of manually detaching the double end shut-off quick disconnects. The cask is then trans-ported on the pallet lift hand truck to the Radiochemistry Sample Room..

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The second means of transport is via sample syringes. The diluted samples are obtained through septum plugs located behind a small penetration in the shield wall using a syringe for sample withdrawal. Standard laboratory syringes are employed for this purpose. Following removal of a diluted sample the syringe is capped to prevent loss of the sample and placed in a hand held container and transported to the Radiochemistry Sample Room.

The response to Criteria ( 6) addresses the radiation protection aspects of Grab Sample handling.

The PASS is designed with the capability to sample under the condition of a loss of offsite power. The PASS is supplied with electric power from a 120/208 NAB power panel which is connected to the diesel generator.

Criterion:

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(2) The licensee shall establish an onsite radiological and chemical analysis capability to provide, within three-hour time frame established above, quantification of the following:

(a) certain radionuclides in the reactor coolant and contain-ment atmosphere that may be indicators of the degree of core damage (e.g., noble gases; iodines and cesiums, and non-volatile isotopes);

(b) Hydrogen levels in the containment atmosphere; (c) dissolved gases (e.g., H2), chloride (time allotted for analysis subject to discussion below), and boron concentra-tion of liquids.

(d) Alternatively, have inline monitoring capabilities to per-form all or part of the above analyses.

(2a) Radionuclide analysis is performed through analysis of diluted and undiluted grab samples obtained from the PASS.

The capability for collecting and handling these grab samples are descr'ibed in the response to Criterion (1) above. The samples are analyzed using Germanium-Lithium semiconductor analyzers located in the Radiochemistry Sample Room. These Analyzers are calibrated for the radionuclides Xe 133, I-131,-

I-133, Cs-137, Cs-134, Kr-85 and Ba-140 among others. The analyzers can quantify all radionuclides required for core damage assessment.

The effcts of background radiation on sample radiochemistry analysis is reduced through the use of lead shielding around the sample chamber. As a backup, should the background radiation levels in the Reactor Auxiliary Building exceed the capabilities of this shielding, the samples can be transported to the radio-chemistry laboratory at the St. Lucie Unit 1 plant. St. Lucie Unit 1 has redundant analysis equipment to that provided in the Unit 1 plant. The two plants are adjacent and transport of the samples would not require them to be removed from the site. The means of transport is the same as that described in response to Criterion (1).

Provisions are provided to estimate the extent of core damage based upon both radionuclide concentrations and other physical parameters. The interim procedure for core damage assessment

, using only radionuclide data is attached., The comprehensive procedure based upon all available plant data is in preparation and will be provided by the scheduled initial criticality of the St. Lucie Unit 2 reactor. The comprehensive procedure will .be based upon the parameters of reactor coolant and containment

atmosphere hydrogen concentration, core exit thermocouple temperature, and containment dose rates in addition to the radionuclide data.

.(2b) There are two provision provided to analyze the containment building atmosphere for hydrogen content.

The first is the Containment Hydrogen Analyzer Subsystems.

This subsystem is described in detail in Section 6.2.5.2.1 of the FSAR. Two redundant subsystems are provided consisting of the sample and return piping, associated valves, hydrogen analyzer, grab sample cylinder, sample pump, moisture separator, cooler, instruments, calibration gas line and reagent gas line.

The subsystems have the capability to obtain samples from seven locations within the containment building. The subsystem valves and piping are designed and fabricated in accordance with ASME Section. III, Class 2 and N-stamped, the instrumentation is Class lE in accordance with the IEEE standards. The subsystem is designed with its own backup grab sample capability.

The second provision is grab sample analysis from the PASS.

The PASS provides a means to obtain diluted samples of the containment atmosphere. The means for handling these grab samples is described in response to Criterion (1).

The grab samples obtained from either the Containment Hydrogen Analyzer Subsystem or the PASS can be analyzed by the technique of gas partitioning in the Radiochemistry Sample Room.

(2c) The PASS provides the means to perform on-line measurement of reactor coolant total dissolved gas, hydrogen and oxygen offgas,

dissolved oxygen concentration, pH, and boron concentration.

PASS is described in Section 9.3.6 of the FSAR.

'he The provisions to provide chloride analysis is provided in response to Criterion (5) below=.

The range of chemical analysis capability is as follows:

(1) Total dissolved gas 0 to 2000 cc/kg (2) Hydrogen offgas 0 to 1005 of T.D.G.

(3) Oxygen offgas 0 to 25$ of T.D.G.

(4) Dissolved oxygen 0 to 20 ppm (5) Boron 0 to 5000 ppm (6) pH 3 to 12 (2d) As stated in the response to Criterion 2(c) the PASS employs on-line instrumentation for chemistry analysis. Demonstration of the applicability of the selected on-line instrumentation is shown through testing. under the anticipated radiation environment for post-accident use. A detailed description of the PASS instrumentation and of the applicability test program is provided in a report previously issued to the HRC Chemical Engineering Branch. This report is entitled "Engineering Evaluation And Functional Testing For The C-E PASS Components And Instrumentation",

CEN-229(L)-P dated November 1982, prepared by Combustion Engineering, Inc.

Particular attention is provided in the subject report concerning applicability of the PASS boron meter to analyze containment building sump samples in the presence of pH control additives.

Specific testing was performed to verify this performance. A detailed description of that test program is provided as Appendix II of the subject report.

Criterion:

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(3) Reactor coolant and containment atmosphere sampling during post accident conditions shall not require an isolated auxiliary system (e.g., the letdown system, reactor water cleanup system (RMCUS)) to be placed in operation in order to use the sampling system.

Response: The PASS contains all valves, sample vessels, pumps and heat exchangers necessary to process'eactor coolant and contain-ment atmosphere samples during post accident conditions. The PASS returns all sample flow to containment to preclude un-necessary contamination of auxiliary systems and to ensure that high level waste remains isolated within the containment.

The PASS draws reactor coolant samples from the Reactor Cool-ant System (RCS) hot leg when RCS pressure is between 200 psig

,. and 2485 psig, or from a Low Pressure Safety Injection (LPSI) pump when RCS pressure is lower than 200 psig. The PASS draws containment samples via containment atmosphere sampling lines.

The PASS has complete sample processing capabilities and has sample lines that are not shared with nor originiating from an isolated auxiliary system. Piping and Instrumentation diagrams'or the PASS are provided in Figures 9.3-6a, 9.3-6b, and 9.3-6c of the FSAR. Therefore, reactor coolant and containment atmos-phere sampling do not require an isolated auxiliary system be placed in operation in order to sample with the PASS.

Valves which are not accessible after an accident are environmentally qualified for the conditions in which they must operate.

Criterion: (4) Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with un-pressurized reactor coolant samples. The measurement of either

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total dissolved gases or H2 gas in reactor coolant samples is considered adequate. Measuring the 02 concentration is recommended, but is not mandatory.

Response: The PASS provides a remote capability to obtain a pressurized reactor coolant sample and quantify both the amount of total dissolved gas in the coolant and the hydrogen and oxygen composition of that gas. The PASS also provides the capability to reasure the dissolved oxygen concentration below saturation in the range of 0 to 20 ppm.

The method used to measure and analyze the gas is as follows.

A sample of the Reactor Coolant System is initiated by opening system isolation valves (including containment isolation. valves using CIAS override, if necessary) and purging a reactor coolant sample. through the PASS sample vessel/heat exchanger, where it is cooled, through a throttle valve to reduce the pressure, through the in-line chemistry analysis e'quipment, to the reactor coolant drain tank. At reactor coolant pressures less than 200 psig, containment sump sample flow is purged'n the same manner using the LPSI pump discharge connection. After sufficient purging, a pressurized sample is then collected by isolating the sample vessel/heat exchanger. Total dissolved gas concentration is determined by degassing the sample. This is accomplished by de-pressurization and circulation by alternate operation of the burette isolation valve and the sample circulation pump. The resulting displacement of liquid into the burette is used to calculate the dissolved gas concentration. The collected gases, which have been tripped from the liquid, are then directed through

a float valve for moisture separation and circulated through hydrogen and oxygen analyzers. The H2 analyzer is a thermal conducitivity devide that determines and indicates the volu~e percent of H2 present in the gas removed from the liquid sample. The 0 analyzer is a para-magnetic device that determines and indicates the volume percent of 02 present in the gas taken out of, the liquid sample. After recording the hydrogen and oxygen gas concentrations, the gas sample vessel, which contains nitrogen, is placed on line to dilute the gas volume. This dilution operation reduces the radiation levels such that a sample can be drawn from the gas sample vessel by injection of a syringe through a septum plug mounted in the vessel. This sample can be transferred to the site laboratory for subsequent radioisotope and backup gas quantifi-cation. Prior to sample withdrawal, additional dilution, which may. be necessary for'his quantification, may be performed by further nitrogen addition, circulation and veniing. The gaseous measurement portions of the system are flushed with demineralized water and purged with N2 to reduce personnel exposure during withdrawal of the sample and to reduce contamination plateout between samples.

The dissolved ox'ygen concentration, below the saturation level for depressurized reactor coolant, is measured using a Orbisphere polarographic instrument. This instrument operates in the range of 0 to 20 ppm. A reactor coolant sample is purged through the system for analysis. The sample is depressurized, cooled, and the total gas is evolved down to saturation levels from the sample during depressurization. The evolved gas, if any, is separated and returned to the containment building.

'l The processed reactor coolant is then purged through the in-line dissolved oxygen analyzer for measurement.

1 Criterion: (5) The time for a chloride analysis to be performed is dependent upon two factors: (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 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.

Response: A four step process is used in the analysis of reactor coolant for chlorides. These four steps are: (1) initial analysis of a diluted grab sample for chloride concentration, (2) measurement of reactor coolant dissolved oxygen concentration in the range of 0 to 20 ppm using on-line instrumentation, (3) collection of an undiluted grab sample in the shielded Grab Sample Facility with retention for analysis within 30 days consistent with ALARA, and (4) verification that dissolved oxygen is less than or equal to 0.1 ppm by measurement of a dissolved hydrogen residual of greater than or equal to 10 cc/kg.

The initial grab sample for chlorides is obtained using the PASS with sample dilution of a factor of 1000. This sample is transported to the Radiochemistry Sample Room for chloride analysis using the mercuric nitrate manual titration technique.

The capability to perform chloride analysis of grab samples using low chromotography is planned for the future but will not be

available at the time of initial reactor criticality.

The dissolved oxygen concentration is measured by the PASS using an Orbisphere dissolved oxygen analyzer. The measure-ment is made on-line with an instrument range of 0 to 20 ppm on an undiluted depressurized reactor coolant sample. This measurement is described in response to Criterion (4) above.

The undiluted grab sample is collected using the shielded grab sample facility. The procedures for operating and transporting this device are described in response to Criterion (1) above.

The undiluted sample is stored in the shielded grab sample facility until the dose rates from the sample have dropped sufficiently to C

allow analysis consistent with ALARA.

Verification of low oxygen concentration ip the reactor coolant by measuring the residual hydrogen concentration is performed, using the PASS as described in response to Criterion (4):

Criterion: (6) The design basis for plant equipment for reactor coolant and containment atmosphere sampling and analysis must assume that it is possible to obtain and analyze a sample without radiation exposur'es 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 from H. R. Denton to all licensees)),

Response: The PASS performs the chemistry analysis for boron, pH, total dissolved gas, hydrogen and oxygen offgas, and dissolved oxygen using remote operation of on-line instrumentation. The time required to perform these analyses is less than three hours.

As shown in Figure 12.3A-2 of the FSAR the dose rate in the PASS Laboratory following a plant accident consistent with the Regulatory Guide 1.4 source terms is less than 15 mr/hr.

Therefore, it is anticipated that the personnel dose obtained during these sample procedures is less than 45 mrem.

The initial scoping analysis for reactor coolant chloride analysis requires a diluted grab sample. The attached Figure 1 provides the dose rat'es calculated for a syringe sample obtained 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following an accident consistent with the Regulatory Guide 1.4 source term assumptions. For chloride analysis the grab sample dilution is taken at a factor of 1000. The dose rate for this case is 450 mrem/hr from.a 2.5 cc sample to the extremities of an operator holding the syringe. For chlorides a larger sample is required. The required sample is 25 cc. There-fore, the extremity .dose rate from this sample is 4.5 R/hr. The time required to handle the sample during chloride analysis is a maximum of 15 minutes. The resulting integrated extremity dose is 1.125 Rem. This is well below the 10 CFR Part 50 GDC 19 criteria of 75 Rem to the extremities.

The shielded Grab Sample Facility is designed to collect and retain a 75 cc undiluted sample of the reactor coolant liquid.

The dose rates calculated for the Grab Sample Facility assuming a Regulatory Guide 1.4 source term with no credit taken for decay are 76 mrem/hr at four feet from the Grab Sample Facility and 4.6 Rem/hr at the valve handwheels and quick disconnects. The time

, required for an operator to operate the valves and disconnects is approximately 2 minutes. Therefore, the operator extremity does during collection of an undiluted sample is 150 mrem. This is well below the 10 CFR Part 50 GDC 19 criteria of 75 Rem to the extremitites. A separate shield wall is provided to reduce

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For backup chemistry analysis and for radiochemistry analysis using diluted grab. samples the dose rates calculated are shown

- in Fi,gures 1, 2, and 3 attached. The anticipated values on contact to the sample syringes two hours following an accident yielding Regulatory Guide 1.4 source terms are as follows:

360 mrem/hr for a depressurized reactor coolant sample, 600 mrem/hr

. for a reactor coolant gas sample, and 15 mrem/hr for a containment atmosphere sample. These values are obtained from the figures using the anticipated dilution factors. In each case the sample contact dose rates are sufficiently low to ensure that the 10 CFR 50 GDC 19 dose criteria will be maintained, As shown in Figure 12.3A-7 two alternate routes for access to the PASS Laboratory are available. The maximum dose received using the preferred route is approximately 1 mrem. The maximum dose received using the alternate route is 750 mrem. In each case transit time through the radiation zones are approximately 30 seconds.

Criterion: (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 BHR plants).

Response: The PASS provides the capability to perform inline boron analysis, in addition the PASS provides the capability of obtaining diluted and undiluted grab samples for backup boron

,analysis.

A detailed description of the on-line boron analyzer is provided in a report previously transmitted to the NRC, CEB.

This report is CEN-229(L)-P, "Engineering Evaluation And Functional Testing For The C-E PASS Components And Instrumentation" dated November 1982. This report also describes testing done to demonstrate performance of the boron meter in a radiation environment and to demonstrate accuracy in analysis of both reactor coolant and containment sump water with pH control additives.

Criterion: (8) If inline monitoring is used for any sampling and analytical capability specified herein, .the licensee shall provide back-up 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.

Response: The capabilities of the PASS system to provide backup sampling through grab samples was discussed in conjunction with grab sampling capabilities in the response to Criterion 1. The PASS provides a depressurized grab sample facility with lead shielding for undiluted samples and multiple dilution capability for diluted samples collected via syringe. The PASS provides the following backup grab sample capability: 1) undiluted reactor coolant, 2) diluted reactor coolant, 3) diluted reactor coolant off gas and 4) diluted containment atmosphere gas.

The undiluted reactor coolant sample is obtained via the Grab Sampling. Facility (see response to Criterion 1 for description). The diluted samples are obtained through septum plugs located behind a small penetration in the shield wall us-ing a syringe for sample withdrawal. All gas samples are diluted before withdrawal for the purpose of radiation protection.

Although it is possible to obtain undiluted gas samples, radiation protection is ensured by procedures and mechanical arrangements which deliberately'iscriminate against possible a'ccidental exposure to undiluted radioactive gas.

The equipment provided for backup sampling is capable of providing at least one sample per, day for an unlimited period following an accident.

Grab samples are provided as backup to the in-line analyzers used by the PASS for measurement of hydrogen gas, oxygen gas, pH and boron. The analysis of hydrogen and oxygen gas grab samples is performed using a Foxboro Gas Partitioner located in the Radiochemistry Sample Room. The gas partitioner is equiped with a supply system designed to accept the sample directly from the syringe used to remove the sample from the PASS. The analysis of boron is performed using manual Mannitol Titration on a 0.15 ml sample.

The capability to perform boron analysis using IOI'l chromatography is planned in the future. The analysis for pH will be performed with undiluted samples at such time following the accident that is consistent with ALARA.

Criterion'. (9) The licensee's radiological and chemical sample analysis capability shall include provisions to:

(a) Identify and quantify the isotopes of the nuclide catego.ies discussed above to levels corresponding to the source terms given in Regulatory Guide 1.3 or 1.4 and 1.7.'here necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided. Sensitivity of onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from approximately lx10 Ci/g to 10 Ci/g.

(b) Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with acceptably small error (approximately a factor of 2). This can be accomplished through the use of sufficient shielding around samples and outside sources, and by the use of ventilation system design which will control the presence of airborne radioactivity.

Response: (ga) The PASS provides the capability to obtain dilute samples for the purpose of identifying and quantifying the isotopes present in the reactor coolant and containment building atmosphere samples.

The responses to Criterion 1, 6, and 8 describes the PASS provisions for handling grab samples.

The PASS provides the capability to obtain the following grab samples: (1) undiluted, depressurized, cooled reactor coolant liquid, (2) diluted depressurized cooled reactor coolant liquid, (3) diluted, depressurized cooled offgas from the reactor coolant, and (4) diluted containment building atmosphere. The locations from which these samples amy be obtained include the reactor

coolant system hot leg, the containment building sump, and the containment building atmosphere. In each case the PASS provides the means to continuously vary the grab sample dilution factor to suit the needs of the analysis.

The grab samples will be analyzed in the Radiochemistry Sample Room using semiconductor detectors with multichannel analyzers.

Predicted specific activity of the samples are based upon Regulatory Guide 1.4 and 1.7 assumptions. The Figures 1 through 3 provided in response to Criterion 6 indicate the predicted gross specific activity and corresponding dose rates for grab sample analysis. The semiconductor detectors will be calibrated to permit measurement of nuclide concentration in the range of lxl0 Ci/g to 10 Ci/g.

(gb ) The effects of background radiation on sample radiochemistry analysis is reduced through the use of lead shielding around the sample chamber. A background count rate will be recorded prior to sample analysis. As a backup, should the background radiation levels in the Reactor Auxiliary Building exceed the capabilities of the shielding, the samples can be transported to the radio-chemistry laboratory at the St. Lucie Unit 1 plant. St. Lucie Unit 1 has redundant analysis equipment to that provided in the Unit 2 plant. The two plants are adjacent and transport of the samples would not require them to be removed from the site. The means of- transport is the same as that described in response to Criterion (1).

Criterion: (10) Accuracy, range, and sensitivity shall be adequate to provide pertinent data to the operator in order to'describe radiolo-gical and chemical status of the reactor coolant systems.

Response: Instrumentation accuracies and ranges requirements for the boron, hydrogen, oxygen and pH meters are provided in System descrip-tion section of the PASS technical manual. The accuracy and range of the boron meter is +21 and 0-5000 ppm boron, respecti-vely. The accuracy and ranges of the hydrogen meter are +2K -..

and 0-10Ã by volume H2 and 0-100Ã by volume H2, respectively.

The accuracy and ranges of the oxygen meter are +2< and 0-5%

by volume 0 and 0-24K by volume 02. The accuracy and range of the pH meter are +.3 pH units and 3-12 pH units. These values are stated under normal environmental conditions.

Demonstration of the applicability and accuracy of the selected on-line instrumentation under post accident environmental conditions is shown through testing under the anticipated radiation environment consistent with Regulatory Guide 1.4 source terms.

The results of this testing are described in a r port previously provided to the NRC. This report is CEN-229(L)-P "Engineering Evaluation And Functional Testing For The C-E PASS Components And Instrumentation" dated tlovember 1982. This report was prepared by Combustion Engineering, Inc.

With regard to personnel training and familiarity with the PASS equipment and procedures, all FPRL Chemistry Department technicians will be trained both in the classroom and in actual hands-on operations, as a function of the Chemistry Department training program. Operating procedures will be developed and they will be consistent with the recommendations of the PASS supplier (Combustion Engineering).

Combustion Engineering is supplying video taped refresher training programs- to cover" both PASS Calibration and Maintenance, and Operations and Post Accident Chemistry'.

Criterion: (ll) In the design of the post accident sampling and analysis capability, consideration should be given to the following items:

(a) Provisions for purging sample lines, for reducing plateout in sample lines, for minimizing sample loss or distortion, for preventing blockage of sample lines by loose material in the RCS or containment, for appropriate disposal of the samples, and for flow restrictions 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 should be as short as possible to minimize the volume of fluid to be taken from containment.

The residues of sample collection should be returned to containment or to a closed system.

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

Response: (lla) The design of the PASS includes many features which serve to maintain sample integrity and limit radiological ex-posure or release. Many of these features were explained in response to earlier questions, namely dilution capabilities, operability independence with respect to isolated auxiliary .

systems, sample disposal via returning sample to containment

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thereby precluding unnecessary contamination of external environments and remote calibration sampling and purging capabilities. The PASS provides additional features which serve to maintain sample integrity and limit radiological exposure or release. Piping lengths are kept as short as possible thus limiting platout in sample lines. Pipe diameter downstream of exis'ting sample paths is 1/2" to 3/8" thus providing high velocity turbulent sample and purge flow at achievable flow rates. Sample and purge flow velocity and Reynolds number for the reactor coolout sample at the recommended flow rate of 1 gpm are of the order of 1.5 ft/sec and 10 ,

respectively. Sample purge and sample velocity for the contain-ment atmosphere'sample at recommended flow rates are 7.5 to 10 ft/sec and 1.5 ft/sec, respectively.

A strainer upstream of the sample vessel heat exchangeer is designed to remove insoluble particles which may cause sample station chemistry instrumentation to become plug-The strainer can be backflushed, with demineralized

'et.

water remotely by operation of valves at the control panel.

(lib) The PASS cabinet is equipped with louvers sized to pass up to 333 scfm of airflow from the surrounding room to the ventilation system exhause connection in the upper portion of the enclosure. Air flow is routed through charcoal absorbers.

This air flow precludes any buildup of radioactive gas or hydrogen gas and provides for removal of heat generated by internal components.