ML20129E198

Forwards Safety Evaluation on Util 830609,840809 & 850530 Submittals Re post-accident Sampling Sys.Ten of 11 Criteria in NUREG-0737,Item II.B.3 Met.Procedure for Estimating Core Damage Acceptable on Interim Basis

ML20129E198
Person / Time
Site:	Pilgrim
Issue date:	07/01/1985
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Harrington W BOSTON EDISON CO.
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM NUDOCS 8507160753
Download: ML20129E198 (8)

Text

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July 1,1985

_ DISTRIBUTION CdDocket File

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• Docket No. 50-293 NRC PDR Local PDR ORB #2 Reading HThompson OELD Mr. William D. Harr_ington SNorris Senior Vice President, Nuclear PLeech Boston Edison Company ELJordan 800 Boylston Street JPartlow Boston, Massachusetts 02199 BGrimes ACRS(10)

Dear Mr. Harrington:

Gray File FWitt

SUBJECT:

NUREG-0737, ITEM II.B.3, JNorris POST-ACCIDENT SAMPLING Re: Pilgrim Nuclear Power Station We have completed our review of your submittals- of June 9,1983; August 9, 1984; and May 30, 1985, concerning the post-accident sampling system (PASS) at the Pilgrim Nuclear Power Station. As a result of this review, we find that ten of the eleven criteria in Item II.B.3 of NUREG-0737 have been met and that'your procedure for estimating core damage is acceptable on an interim basis.

On June 17, 1985, Mr. Kahler of your staff informed us by telephone that a more complete procedure for estimating core damage will be provided by September 20, 1985 for our review.

Enclosed is a copy of our Safety Evaluation of your submittals.

Sincerely, Original signed by MGrotenhuis for/

Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Enclosure:

As stated cc w/ enclosure:

See next page DL :@_B_#2 , DL:0R DL:0RB#2 SNofM s:ajs Pleech DVassallo 06/41/85 06ht] /85 / / /85 e507160753 850701 3 DR ADOCK 0500

Mr. William D. Harrington Boston Edison Company Pilgrim Nuclear Power Station cc:

Mr. Charles J. Mathis, Station Mgr. Thomas A. Murley-Boston Edison Company Regional Administrator RFD #1, P.ocky Hill Road Region I Office Plymouth, Massachusetts 02360 U. S. Nuclear Regulatory Commission 631 Park Avenue Resident Inspector's Office King of Prussia, Pennsylvania 19a06 U. S. Nuclear Regulatory Commission Post Office Box 867 Mr. A. Victor Morisi Plymouth, Massachusetts 02360 Boston Edison Company 25 Braintree Hill Park Mr. David F. Tarantino Rockdale Street Chairman, Board of Selectman Braintree, Massachusetts 02184 11 Lincoln Street Plymouth, Massachusetts 02360 Office of the Conmissioner Massachusetts Department of Environmental Quality Engineering One Winter Street Boston, Massachusetts 07108 Office of the Attorney General 1 Ashburton Place 19th Floor Boston, Massachusetts 02108 Mr. r ,rt M. Hallisey, Director Radi. Control Program Mass '

ts Department of Pubi fealth 150 Tremont Street Boston, Massachusetts 02111

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% UNITED STATES 8(  % NUCLEAR REGULATORY COMMISSION g j WASHINGTON, D. C. 20555

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION POST-ACCIDENT SAMPLING SYSTEM (NUREG-0737, ITEM II.B.3)

BOSTON EDISON COMPANY -

PILGRIM NUCLEAR POWER STATION, UNIT 1 DOCKET NO. 50-293

1.0 INTRODUCTION

Subsequent to the TMI-2 incident, the need was recognized for an improved post-accident sampling system (PASS) to determine the extent of core degradation following a severe reactor accident. Criteria for an acceptable sampling and analysis system are specified in NUREG-0737, Item II.B.3.

The system should have the capability to obtain and quantitatively analyze reactor coolant and containment atmosphere samples without radiation exposure to any individual exceeding 5 rem to the whole body or 75 rem to m the extremities (General Design Criteria (GDC) 19) during and following an ~

accident in which there is core degradation. Materials to be analyzed and quantified include certain radionuclides that are indicators of severity of core damage (e.g., noble gases, isotopes of iodine and cesium, and nonvolatile isotopes), hydrogen in the containment atmosphere and total dissolved gases or hydrogen, boron, and chloride in reactor coolant samples.

To comply with NUREG-0737, Item II.B.3, Boston Edison Company (the licensee) must (1) review and modify his sampling, chemical analysis, and radionuclide determination capabilities as necessary and (2) provide the staff with information pertaining to system design, analytical capabilities and procedures in sufficient detail to demonstrate that the criteria are met.

2.0 EVALUATION By letters dated June 9, 1983, August 9, 1984 and May 30, 1985, the licensee provided information on the post-accident sampling system at Pilgrim Station, as follows:

Criterion (1):

The licensee shall have the capability to promptly obtain reactor coolant samples and containmen+ atmosphere samples. The combined time allotted for sampling and an i, sis should be three hours or less from the time a decision is made o take a sample.

The licensee has provided sampling and analysis capability to promptly obtain and analyze reactor coolant, suppression pool (from residual heat removal (RHR) loop) containment sump, and containment atmosphere samples within

three hours from the time a decision is made to take a sample. During loss of off-site power, alternate power sources are available for both gas and liquid sampling systems that can be energized in sufficient time to meet the three hour samplin meet Criterion (1)gand andare, analysis timeacceptable.

therefore, limit. We find that these provisions Criterion (2):

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

a) Certain radionuclides in the reactor coolant and containment atmosphere that may be indicators of the degree of core damage (e.g., noble gases, iodines and cesiums, and nonvolatile isotopes);

b) hydrogen levels in the containment atmosphere; c) dissolved gases (e.g., H7 ), chloride (time allotted for analysis subject to Discussion below), and boron concentration of liquids; d) alternatively, have in-line monitoring capabilities to perfom all or part of the above analyses.

The PASS provides grab sample capability for pH, chloride, boron, radionuclide analysis and dissolved hydrogen and oxygen in the reactor coolant, and grab samples and/or in-line monitoring of hydrogen oxygen and ganr.a spectrum in the containment atmosphere. The PASS provides the capability to collect diluted or undiluted liquid reactor coolant and gaseous grab samples.

The licensee has provided a core damage estimation procedure based on reactor coolant and containment atmosphere radionuclides and two other plant parameters.

This is acceptable as an interim procedure. The licensee has committed to provide a final core damage estimation procedure by September 20, 1985.

The final procedure should include information the from BWR Owners Group procedure which will provide a more realistic core damage estimate. We find that the licensee partially meets Criterion (2).

Criterion (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) to be placed in operation in order to use the sampling system.

Reactor coolant and containment atmosphere sampling during post-accident conditions does not require an isolated auxiliary system to be placed in operation in order to perfom the sampling function. The PASS valves which

( are not accessible after an accident are environmentally qualified for the conditions in which they need to operate. These provisions meet Criterion (3) and are, therefore, acceptable.

Criterion (4): ._

Pressurized reactor coolant samples are not required if the licensee can quantify the amount of dissolved gases with unpressurized reactor coolant samples. The measurement of either total dissolved gases or H,, gas in reactor coolant samples is considered adequate.

Measuring the 02 concentration is recommended, but is not mandatory.

Pressurized reactor coolant samples are cooled and degassed to obtain representative total dissolved gas samples at the PASS sampling station.

The hydrogen and oxygen concentrations are measured by gas chromatography.

Dissolved oxygen concentrations of 1 css than 0.1 ppm can be verified by measurement of a dissolved hydrogen residual of greater than 10 cc/kg.

Alternately, dissolved oxygen can be obtained by gas chromatography. We have determined that these provisions meet Criterion (4) of Item II.B.3 in NUREG-0737, and are, therefore, acceptable.

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 applicant 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 beir.3 taken. For all other cases, the applicant shall provide for the analysis to be completed within four days. The chloride analysis does not have to be done on-site.

Chloride analysis is performed on a diluted sample within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of sampling by using an ion chromatograph. Chloride analysis accuracy by this technique is less than 0.1 ppm. An undiluted sample will also be obtained and stored on-site for analysis within 30 days. These provisions meet Criterion (5) and are, therefore, acceptable.

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 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 frorr H.R. Denton to all licensees).

The licensee has perfomed a time-person-motion study to ensure that operator exposure while obtaining, transporting, and analyzing a PASS sample is within the acceptable limits. This operator exposure includes entering and exiting the sample panel area, operating sample panel manual valves, positioning the grab sample into the shielded transfer casks,. transporting casks and performing sample analyses. PASS personnel radiation exposures from reactor coolant and containment atmosphere sampling and analysis are within 5 rem whole body and 75 rem extremities, which meet the requirements of GDC 19 and Criterion (6) and are, therefore, acceptable.

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 BWR plants).

A diluted grab sample of the reactor coolant will be analyzed for boron ~ by plasma spectrometry or spectrophotometry in the range of 0 to 1000 ppm with an accuracy of 10%. This provision meets the recomendations of Regulatory Guide 1.97, Rev. 2 and Criterion (7) and is, therefore, acceptable.

Criterion (8):

If in-line 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 off-site facilities is acceptable. Equipment provided for backup sampling shall be capable of providing at least one sample per week until the accident condition no longer exists.

In-line monitoring is used to determine containment atmosphere hydrogen and oxygen levels. These are redundant safety related systems required by Item l

II.F.1.6 of NUREG-0737, and, therefore, it is acceptable not to have backup grab sample capability. Radionuclide analysis, dissolved hydrogen and oxygen, pH, conductivity, and boron are obtained on grab samples of reactor coolant.

The PASS can obtain both diluted and undiluted samples. We find these provisions meet Criterion (8) and are, therefore, acceptable.

Criterion (9):

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

l a) Identify and quantify isotopes of the nuclide categories discussed i

above to levels corresponding to the source term given in Regulatory Guides 1.3 or 1.4 and 1.7. Where necessary and practicable, the ability to dilute samples to provide capability for measurement and reduction of personnel exposure should be provided. Sensitivity of on-site liquid sample analysis

capability should be such as to permit measurement of nuclide concentration in the range from approximately 1 pCi 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 an 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 a ventilation system design which will control the presence of tirborne radioactivity.

The radionuclides in both the primary coolant and the containment atmosphere will be identified and quantified. Provisions are available for obtaining diluted reactor coolant samples to minimize personnel exposure.

The PASS can perform radioisotopes analyses at the levels corresponding to the source term given in Regulatory Guides 1.3, Rev. 2, and 1.7. Radiation background levels will be restricted by shielding. Ventilated radiological and chemical analysis facilities are provided to obtain results with an roximately a factor of 2). We find these acceptably provisions small meet error (app (9) and are, therefore, acceptable.

Criterion Criterion (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.

The accuracy, range, and sensitivity of the PASS instruments and analytical procedures are consistent with the recomendations of Regulatory Guide 1.97, Rev. 3, and July 27, 1983 clarifications of NUREG-0737, Item II.B.3, Post-Accident Sampling Capability. Therefore, they are adequate for describing the radiological and chemical status of the reactor coolant. The analytical methods and instrumentation were selected for their ability to operate in the post-accident sampling environment. The licensee has not provided standard test matrix testing information on similar equipment; however, the standard test matrix and radiation effect evaluation indicated no interference in the PASS analyses. Technician refamiliarization training will occur at least every 6 months. The equipment and procedures used in the PASS will be tested or calibrated to maintain a high level of reliability. We determined that these provisions meet Criterion (10) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.

Criterion (11):

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

a) Provisions should be made 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

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reactor coolant system (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 6r 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 adsorbers and high-efficiency particulate air (HEPA) filters.

The licensee has addressed provisions for purging and recirculation back to containment to ensure that samples are representative, and for flow restrictions and/or isolation valves to limit reactor coolant loss from a rupture of the sample line. To limit iodine plateout, the containment atmosphere shinple line is heat traced. The post-accident sampling stations are ventilated by the standby gas treatment system which contains charcoal adsorbers and HEPA filters. The post-accident reactor coolant and the containment atmosphere samples will be representative of the reactor coolant and the containment atmosphere respectively. We determined that these provisions meet Criterion (11) of Item II.B.3 of NUREG-0737, and are, therefore, acceptable.

3.0 CONCLUSION

On the basis of our evaluation, we conclude that the post-accident sampling system at Pilgrim Station meets ten of the eleven criteria in Item II.B.3 in NUREG-0737. The procedure for estimating reactor core damage is acceptable on an interim basis. The licensee has committed to providing a plant specific procedure for estimating the extent of core damage based on the BWR Owners Group Procedure.

Principal Contributor: Frank Witt Dated: July 1,1985 l

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