ML19270D619

Discusses Installation,Testing & Associated Training for Diesel Generator Panel Isolation Switches.Commits to Complete Installation Prior to 5% Power.Commitment Obviates Need for App R Exemption Requests

ML19270D619
Person / Time
Site:	Waterford
Issue date:	08/09/1984
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	Knighton G Office of Nuclear Reactor Regulation
References
WP384-2209, NUDOCS 8408130342
Download: ML19270D619 (2)

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I Docket No. 50-412 e MEMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM: William V. Johnston, Assistant Director Mate.ials, Chemical & Environmental Technology Division of Engineering

SUBJECT:

SAFETY EVALUATION REPORT FOR BEAVER VALLEY POWER STATION, UNIT No. 2 Plant Name: Beaver Valley Power Station, Unit No. 2 Suppliers: Westinghouse Electric Corporation; Duquesne Light Company Licensing Stage: OL Docket No.: 50-412 Responsible Branch and Project Manager: LB #3; M. Ley Reviewers: J. Wing, B. Turovlin Description of Task: Operating License Review Status: SER Complete - Two open items In our Draft Safety Evaluation Report, we concluded that Sections 6.1.1, 6.1.2, 9.1.2, 9.1.3, 9.3.2A, 9.3.4, 10.3.5, 10.4.1, 10.4.6 and 10.4.8 were acceptable. Section 9.3.2B (Post-Accident Sampling) met seven of the eleven criteria in Item II.B.3 of NUREG-0737.

By letter dated April 18, 1984, the applicant provided additional information.

Enclosed is our safety evaluation. Based on our evaluation, we now conclude that nine of the eleven criteria in Item II.B.3 of NUREG-0737 are met. Section 9.3.28 should be replaced with the attached evaluation. All other sections in the draft safety evaluation remain unchanged.

Input for the SALP process is also enclosed.

William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering

Enclosures:

As stated

Contact:

J. Wing B. Turovlin x27278 x28556 cc: See next page B408310342 B40827 CF ADDCK 05000412 CF

Thomas M. Novak cc: R. Vollmer D. Eisenhut V. Benaroya C. McCracken G. Knighton T. Sullivan S. Pawlicki M. Ley J. Weeks (BNL)

B. Turovlin J. Wing

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Safety Evaluation Report by the Office of Nuclear Reacto.- Regulation for Duquesne Light Company Beaver Valley Power Station, Unit No. 2 Docket No. 50-412 9.3.2 Process and Post-Accident Sampling Systems B. Post-AccidentSamplingSystem(NUREG-0737,II.B.31 Introduction In our draft safety evaluation, we concluded that the post-accident sampling system met seven of the eleven criteria in Item II.B.3 of NUREG-0737. The four criteria which were unresolved are:

(2) Provide a plant specific procedure to estimate the extent of core damage.

(8) Provide backup sampling capability via grab samples.

(9) Provide capability to measure radionuclide concentrations in the range of 1 pCi/g to 10 Ci/g with an error of a factor of 2.

(10) Provide information on the accuracies, sensitivities, and performance of the PASS instrumentation and analytical procedures in the post-accident water chemistry and radiation environment. Provide the frequency for demonstrating operability of procedures and instrumentation and retraining of operators on semi-annual basis.

By letter dated April 18, 1984, the applicant provided additional information.

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Evaluation Criterion (2):

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

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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 non-volatile isvtopes);

b) hydrogen levels in the containment atmosphere; c) dissolved gases (e.g. , H 2

), chloride (time allotted for analysis subject to discussion below), and boron concen-tration of liquids; d) alternatively,havefn-linemoni;oringcapabilitiestoperform all or part of the above analyses.

The PASS provides for in-line analysis of the reactor coolant and containment sump samples for total dissolved gas and oxygen, pH, chloride and boron concen-trations, and gross radioactivity. Hydrogen and oxygen in containment air are analyzed by in-line instrumentation. Radionuclide gamma spectrum analysis will be performed via grab samples at the onsite emergency response facility laboratory.

The applicant will adopt the Westinghouse Owners Group post-accident core damage assessment methodology.

We determined that these provisions partially meet Criterion (2) of Item II.B.3 in NUREG-0737. The applicant should provide a plant-specific procedure for estimating the de[ ee of core damage.

Criterion (8):

If in-line monitoring is used for any sampling and analytical capa-bility specified herein, the applicant shall provide backup sampling through grab samples, and shall demonstrate the capability of analyzing

the samples. Established planning for analysis at offsite f acilities is acceptable. Equipment provided for backup sampling shall be capatle of providing at least one sample per day for 7 days following onset of the accident and at least one sample par week until the accident condition no longer exists.

The PASS provides in-line analyses for gross radioactivity, boron, chloride, dissolved gases, and pH in reactor coolant and containment sump, and hydtgoen and oxygen in containment air. Backup sampling capability through grab samples is available for these in-line analyses.

We find that thesa provisior.s meet Criterion (8) and are, therefore, acceptable.

Criteion (9):

The applicant's radiological and chemical sample analysis capability shall include provisicas to:

a) Identify and quantify isotopes of the nuclide categories discussed 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 onsite liquid sample analysis capability should be such as to permit measurement of nuclide concentration in the range from

. approximately 1 pCi/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 an acceptably small error (apprc>imately 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 airborne radioactivity.

_4-The radionuclides in both the primary coolant and the containment atmosphere will be identified and quantified. Provisio-e Ice available for diluted reactor coolant samples to minimize personnel exposure. The PASS can perform radioisotopes analyses at the levels corresponding to the source terms given in Regulatory Geides 1.4, Rev. 2 and 1.7. These analyses will be accerate within a factor of two. We find that these provisions meet Criterion (9) and are, therefore, acceptable.

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 system.

The accuracy, range, and sensitivity of the PASS instruments and analytical procedures are consistent with the recommendations of Regulatory Guide 1.97, Rev. 3, and the clarifications of NUREG-0737, Item II.B.3, Post-Accident Sampling Capability, transmitced to the applicant on August 31, 1983.

Therefore, they are adequate for describing the radiological and chemical status of the reector coolant. Equipment used in post-accident sampling and analyses will be calibrated or tested at least every six months.

Retraining of operators for post-accident sampling is scheduled at a frequency of once evety six months. We determined that these provisions partially meet Criterion (10). The applicant should provide information on the performance of the PASS instrumentation and analytical procedures in the post-accident water chemistry and radiction environment.

. Conclusion We conclude that the post-accident sampling system meets nine of the eleven criteria in Item II.B.3 of NUREG-0737. The ft.llowing two criteria are unresolve 4

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u (2) Provide a plant specific precedure to estimate the extent of core damage.

(10) Provide information on the perforcence of the PASS instrumentation and

, analytical procedures in the post eccident water chemistry and radiation environ 5ent.

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Input to the SALP Process A. Functional Area: Chemical Technology

1. Management involvement in assuring quality Throughout the review process, the applicant's activities exhibited evidence of prior planning. Policies for quality assurance of protective coating systems were adequately stated and understood.

Rating: Cateogry 2

2. Approach to resolution of technical issues from a safety standpoint The applicant's approach to resolution of the post-accident sampling capability and secondary water chemistry displayed clear understanding of our concern. Conservatism was generally exhibited. The issues were resolved in a viable and sound manner.

Rating: Category 2

3. Responsiveness to NRC initiatives The licensee frequently requires extensions of time to respond to our request for additional information. There are still a number of unresolved issues.

Rating: Category 3

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