ML17334A975

Forwards Complete & Detailed Description of post-accident Sampling Sys Per Reqirements of NUREG-0737,Section II.B.3. Exemption from Section II.B.3 for Listed Items Requested. Fee Paid

ML17334A975
Person / Time
Site:	Cook
Issue date:	06/27/1986
From:	Alexich M INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM AEP:NRC:0678V, AEP:NRC:678V, NUDOCS 8607010188
Download: ML17334A975 (15)

Text

INDIANA8 MICHIGAN ELECTRIC COMPANY P.O. BOX 16631 COLUMBUS, OHIO 43216 June 27, 1986 AEP:NRC:0678V Donald C.

Cook Nuclear Plant Unit Nos.

1 and 2

Docket Nos.

50-315 and 50-316 License Nos.

DPR-58 and DPR-74 NUREG-0737, SECTION II.B.3 POST-ACCIDENT SAMPLING, SYSTEM Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

20555

Dear Mr. Denton:

Per Item I of our letter AEP:NRC:0678S, dated February 19,

1986, attached is a complete and detailed description of the Post-Accident Sampling (PAS) System relative to the requirements of NUREG-0737,Section II.B.3.

As a result of the evaluation completed by NUS Corporation, we are requesting exemptions from NUREG-0737,Section II.BE 3 for the following:

1. Undiluted backup grab samples for in-line monitoring.

2. Heat-tracing containment air sample line to prevent plate-out of iodine.

3. Transporting containment atmosphere sample for hydrogen analysis.

Our schedule for closing out NUREG-0737,Section II.B.3 (MPA 5F-12) is dependent on the NRC's response to the above requests for exemption.

Pursuant to the requirements of 10 CFR 170.12(c),

we have enclosed an application fee of $150.00 for review of this exemption request.

This document has been prepared following Corporate procedures which incorporate a reasonable set of controls to insure its accuracy and completeness prior to signature by the undersigned.

Very truly yours, M. P. Alexich Vice President 'lgSI cm Attachment cc:

John E. Dolan W.

G. Smith, Jr.

- Bridgman R.

C. Callen G. Bruchmann G. Charnoff NRC Resident Inspector

- Bridgman 8g07010188 860+2 PDR ADQCK 05000318 P

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• A 1

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Attachment to AEP:NRC:0678V 0

DE C.

COOK PLANT PAS PROGRAM As a result of NUREG-0578, NUS Corporation was contracted in May 1980 to design and fabricate a post-accident sampling (PAS) system for the Donald C. Cook Nuclear Plant.

In October 1980, during fabrica-tion of the PAS panels, NUREG-0737 was issued.

New requirements and clarifications for the PAS system were included in Item II.B.3. which had not been originally incorporated in the PAS panel design.

These requirements specifically included chloride analysis of reactor coolant samples.

Due to size constraints and stipulated installation deadlines, in-line chloride analysis could not be incorporated into the PAS sampling equipment.

Fabrication and testing of the equipment were completed in August 1981.

The equipment was shipped in September 1981.

Clarifications to NUREG-0737 Item II.B.3 were issued in June 1982 adding new require-ments to the PAS system.

To comply with these requirements, several modifications to the PAS system were initiated.

After completion of these modifications, NUS was contracted in 1985 to provide verification and compliance testing of the PAS system.

This test work was completed in May 1986.

The results of this test work and review of the PAS system to meet the intent of NUREG-0737 have resulted in a request for the following exemptions from NUREG-0737:

'1)

Exem tion from undiluted backu rab sam les for in-line In-line monitoring for pH and dissolved oxygen and in-line degassing foi dissolved hydrogen are preferable to and more accurate than grab sample analyses.

To insure the functional capability of these in-line apparatus, a preventive mainten-ance program is performed semi-annually.

Also, routine use of these systems is necessitated by the plant training program.

If failure should occur during long-term post-accident conditions, the capability exists to flush all sample lines and replace the failed components.

2)

Exem tion from heat-tracin containment air sam le line to revent.

late-out of iodine.

The PAS containment air samples are analyzed for isotopic noble gases and hydrogen only.

Isotopic iodines and non-volatile fission products are analyzed in the reactor coolant and containment sump for assessment of core damage.

Therefore, heat-tracing of the containment air sample line is not necessary and would serve no purpose.

3)

Exem tion from trans ortin containment atmos here sam le for h dro en anal sis.

The PAS system has the capability to obtain and analyze an undiluted containment atmosphere sample for hydrogen using an in-line gas chromatograph.

Otherwise, containment atmosphere samples are diluted and transported to the counting facility

for noble gas analysis.

Analysis of the diluted sample for hydrogen would not be within the range of the laboratory gas chromatograph.

PAS SYSTEM DESCRIPTION The D. C. Cook Plant PAS system consists of a liquid and gas sample panel (LGSP),

an Undiluted Liquid Sample Station (ULSS),

and related control panels.

The sample collection stations are located on elevation 587'f the auxiliary building.

The sample analysis facilities are located on elevation 609'f the auxiliary building approximately 220'rom the PAS sampling panel locations.

Grab samples from either sampling station are removed by lead-shielded syringe and transported in a shielded container to the analytical facilities.

Time and motion studies have been conducted to verify that samples of reactor coolant and containment atmosphere can be obtained and analyzed within three hours from the time a decision is made to take a

sample.

The PAS stations are powered from an emergency power supply or from the 250 V.D.C. station battery system, and therefore samples can be obtained during loss of offsite power.

The samples requiring laboratory analysis can be analyzed only after offsite power is restored.

Based on the number and high reliability of Cook Plant offsite power sources, the sample analysis can be made within the three hours from the time a decision is made to take a sample.

Table I shows the sampling and analytical capabilities of the PAS system.

The system has provisions for sampling the locations shown in Table I for either unit following a LOCA.

(3)

LGSP is designed for in-line analysis of undiluted samples and for dilution of grab samples.

In-line analyses are provided for pH and dissolved oxygen measurements.

Dissolved hydrogen in the ireactor coolant is analyzed by an in-line gas chromatograph.

Dilution of liquid grab samples is performed by isolating a 10 ml sample of reactor coolant in the LGSP and diluting with 10 liters of demineralized water.

Verification of the dilution factor was performed using lithium hydroxide and comparison of soluble isotopes I-131, I-133, I-135 and Na-24.

A comparison of all isotopes (such as Cs-134, Te-129 and Sr-89) was not possible because of the low levels of radio-activity in the test data LGSP diluted samples.

The mean dilution factor was determined to be 846.

The LGSP and ULSS have eight inches of lead shielding.

The LGSP has a ventilation system to remove airborne radiation should a leak occur in the panel.

This exhaust is filtered through a charcoal adsorber and a high-efficiency particulate filter into the auxiliary building ventilation system.

The system is designed to obtain samples under LOCA conditions without exposing any individual to radiation in excess of 5 rem to the whole body and or 75 rem to the extremities.

The ULSS is designed to collect an undiluted reactor coolant sample from Hot Legs 1 or 3 within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of the accident.

The sample is removed by a lead-shielded syringe in 30 days to provide backup chloride analysis.

The PAS system incorporates means for adequate purging of gas and liquid sample lines.

The purge rates for liquid and gas samples are S

dependent on the pressure drop and driving force of each sample point.

(4)

~

The minimum purge time for the furthermost point (the Unit 2 Hot Legs 1

a 3)is approximately 5 minutes at 4 gpm.

To minimize sample leakage@

the sample lines up to and from the LGSP and ULSS are welded.

Demineralized water or nitrogen is available to be used to flush all post-accident sample lines to minimize blockage.

The LGSP liquid sample inlet filter can be back-flushed with demineralized water.

Sample purge and flush fluids are routed to a PAS waste collection tank, then pumped to the containment of the unit where the accident has occurred.

These sample waste lines have remotely operated containment isolation valves to shut off waste flow if necessary.

All PAS valves not accessible after an accident have been modified to be environmentally qualified for conditions in which they must operate.

To operate the three air-operated sample isolation valves inside containment, procedures provide for opening the instrument air containment isolation valves after a containment isolation signal has been initiated.

Comparison of analytical results of samples from the various locations can be used to evaluate sample representativeness and adequacy of mixing.

In conjunction with this comparison, comparison of physical parameters, e.g.,

core temperature and reactor coolant system

pressure, is available to further ensure that representative samples are being taken.

Process auxiliary systems carrying reactor coolant or containment atmosphere gases which are isolated post accident (i.e., the letdown system or the reactor water cleanup system) are not required to be placed in operation, but portions of the existing Nuclear Sampling System must be placed in service to obtain post-accident samples.

A PAS preventive maintenance program has been developed and will be performed every six months by an outside contractor, commencing approximately October 1986.

This program includes service of all PAS components and comparison testing to normal sample data.

The PAS training program consists of classroom and on-the-job training.

Qualification for the PAS System operation is performed by classroom and panel operation testing.

Chemistry technicians are requalified every six months.

PAS ANALYTICALCAPABILITIES Table II depicts the analytical capabilities of the post-accident sampling system.

Matrix testing for boron, dissolved oxygen, chloride and pH analysis were performed by NUS Corporation.

A description of each analysis follows.

Boron Boron analysis is performed using a fluoroborate selective ion electrode.

The range of analysis is 375 to 2000 ppm boron with an accuracy of + 11.4%.

This range is adequate to confirm minimum shutdown margin during post-accident conditions.

Comparison testing showed the PAS analysis to be within 11.7% of the analysis of a normal (undiluted) sample.

Dissolved ~sdno en Dissolved hydrogen analysis is performed by isolating 10 ml of undiluted reactor coolant in the LGSP.

The sample is stripped and (6)

degassed using nitrogen (revisions are to be made to use argon for this purpose) and collected in a 150 ml panel-mounted flask.

(A design change currently being processed will substitute argon as the stripping gas.)

The stripped gas is swept through the evacuated system into a Baseline Gas Chromatograph (GC).

The hydrogen content is recorded on a strip chart.

The GC is calibrated over a 0-2000 cc/kg range.

The accuracy of the GC was determined to be

+ 11% for 15 cc/kg and

+

5.6% for 750 cc/kg.

The comparison test showed the PAS analysis to be within

+

5 cc/kg of the analysis of a normal sample (40cc/kg).

Dissolved

~Ox en Dissolved oxygen analysis is performed using an in-line Yellow Springs Instrument dissolved oxygen probe and remote analyzer-recorder.

The range of the analyzer is 0-20 ppm with an accuracy of

+ 10%.

A PAS analysis showed 2.4 ppm and 0.05 ppm compared to the analysis of a normal sample (2 ppm and 0 ppm, respectively).

pH The pH analysis is performed using an in-line Horizon analog pH meter equipped with a gel-filled combination pH sensing electrode.

The s

range of the analyzer is 0-14 pH with a accuracy of +.15 pH units.

Comparison testing with normal sampling analyses was within.01 to.05 pH units.

Chloride Chloride analysis can be performed on a diluted reactor coolant grab sample within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of the initiation of the LOCA.

The lower

limit of detectability is approximately 10 ppm using an Ion Chromatograph.

Additionally, ULSS is used to obtain an undiluted backup grab sample from the reactor coolant Hot Legs 1 or 3 for chloride analysis.

This sample is collected in an in-line 75 ml flask in the sample station.

(A design change has been initiated to install the in-line flask which is on site.

Presently, samples are collected in a 50 ml bottle inside a lead pig).

Within 30 days, a sample can be obtained from the flask using a lead-lined syringe and analyzed in the Ion Chromatograph.

The undiluted sample's lower limit of detectability is 10 ppb chlorides.

Comparison testing could not be performed due to the low level of chloride in the reactor coolant.

Radionuclides Radionuclide analysis of reactor coolant is performed on a diluted grab sample.

The sample is removed using a lead-lined syringe and taken to the counting room for gamma analysis using the Series 85 MCA with pdp ll/24 computer and Ge(Li)/HPGe detector.

The detector is completely surrounded by a 4-inch-thick lead cave which acts as a shield from external radiation fields.

The detector must be shielded, since at radiation fields greater than 10 mR/hr, the detector does not function properly.

An external background radiation, field of 50 mR/hr, the expected dose rate from unisolated letdown flow will be reduced by a factor of 5000 to approximately 0.01 mR/hr, which will not affect the operation of the detector.

However, in the unlikely event of counting room ventilation failure, the airborne activity in the counting facility could potentially rise to a dose rate of up to 981 mR/hr.

Therefore, samples could not be analyzed as the (8)

dose rate inside the cave would exceed the 10 mR/hr operational limit for the detector due to airborne contamination.

In that case, radio-nuclide analysis would have to be delayed until the condition was corrected.

The range of the counting equipment for radionuclides will be within the 1 uCi/g to 10 Ci/g with an accuracy of + 10%.

Dilution of PAS samples will reduce the activity to normal sample levels.

Compar-ison to daily surveillance samples showed a deviation range from -7.7%

to 10.5%.

h Analysis of noble gas from the lower containment atmosphere is performed on a diluted grab sample.

Containment atmosphere samples are captured in the LGSP in either a 0.5 cc or 5.0 cc isolation chamber.

The sample is then purged into an evacuated 1 liter panel-mounted vessel using argon for dilution.

Samples are then removed from the vessel using a lead-shielded syringe and transported to the counting room for noble gas analysis.

The hydrogen analysis is performed on an undiluted containment atmosphere sample using the in-line Baseline Gas Chromatograph.

Gas dilution system testing was performed by NUS Corporation.

The dilution factors obtained from the testing demonstrated a dilution of 1:225 for the 5.0 cc sample and 1:1997 for the 0.5 cc sample with an accuracy of -0.2% to +11.25%.

TABLE II AMLYSIS Reaator Coolant Boron Dissolved Hydrogen Dissolved Oxygen pH RANGE 0.375 2.0 ppm 0

2000 cc/kg O.l 20 ppm 0-14 ACCURACY

+ 11.4%

5.6%

50 cc/kg

+

11%,

50 cc/kg

+ 0.15 pH 5

9

+ 0.15 pH 5

9 METHOD Fluoroborate Selective Ion Electrode Gas Chromatography In-line probe In-line probe Chloride Radionuclide ntainment Atmos here Noble Gas Hydrogen

'> 10 ppm (diluted)

> 10 ppb (undiluted) 1 pCi/g 10 Ci/g Factor of 2

1 pCi/g - 10 Ci/g Factor of 2

0-4%

Ion Chromatography Ge (Li) Detector Ge (Li) Detector Gas Chromatograph

TABLE I SAMPLE LOCATION RADIONUCLIDEB BORON CHLORIDE D.O.

Hot Leg l or 3

D,G D,G D or U,G U,I U,I U,I N/A Pressurizer Steam Space RHR E or W

Lower Containment Sump Lower Containment Atmosphere D,G D,G D,G D,G D,G D,G D,G D,G D,G D,G U,I U,I U,I N/A N/A N/A N/A U,I U,I U,I N/A N/A N/A

~

N/A U,I N/A Not Applicable G

Grab Sample I In-Line Analysis D Diluted U

Undiluted D.O.

Dissolved Oxygen D.11.

Dissolved Hydrogen

\\r e

June 24, 1986 DOCKET NO(S).

50-315/316 ttr. Johh Dohan, ViceePresident Indiana and Michigan Electric Company c/o American Electric, Power Service Corporation 1 Riverside Plaza

Columbus, Ohio 43216

SUBJECT:

D.

C.

Cook Plant, Units 1 and 2

.r/ ~,~/ "/ r /e r

4 i /

DISTRI ION: w/o enclosure

~Docket-F-i.le~

PRC System NRC PDR Local PDR PWR84 Rdg BJYoungblood Rdg t1Duncan Dlligginton OGC ACRS (10)

JPartlow BGrimes EJordan NThompson The following documents concerning our review of the subject facility are transmitted for your information.

D Notice of Receipt of Application, dated D Draft/Final Environmental Statment, dated D Notice of Availabilityof Draft/Final Environmental Statement, dated D Safety Evaluation Report, or Supplement No.

, dated D Notice of Hearing on Application for Construction Permit, dated D Notice of Consideration of Issuance of Facility Operating License, dated IIP Monthly Notice; Applications and Amendments to Operating Licenses Involving no Significant Hazards Considerations, dated Jul 16 1986 (See page 25770)

D Application and Safety Analysis Report, Volume D Amendment No.

to Application/SAR dated D Construction Permit No. CPPR-D Facility Operating License No.

, Amendment No.

, Amendment No.

dated

, dated D Order Extending Construction'Completion Date, dated D Other (Specify/

Enclosures:

As stated Office of Nuclear Reactor Regulation cc:

See next page

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