ML20024A705

Clarifies Info Provided in Re NUREG-0737,Item II.B.3 Post-Accident Sampling Sys (Pass), Per 830526 Request.Equipment Ordered to Allow Powering of Analyzer & Solenoid Valves from Reliable Source within 3 H

ML20024A705
Person / Time
Site:	Rancho Seco
Issue date:	06/17/1983
From:	Reinaldo Rodriguez SACRAMENTO MUNICIPAL UTILITY DISTRICT
To:	Stolz J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM RJR-83-512, TAC-44473, TAC-60632, NUDOCS 8306220075
Download: ML20024A705 (6)

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SACRAMENTO MUNICIPAL UTILITY DISTRICT O 6201 S Street, Box 15830, Sacramento, California 95813; (916) 452-3211 RJR 83-512 June 17,1983 DIRECTOR'0F NUCLEAR REACTOR REGULATION ATTENTION JOHN F STOLZ CHIEF OPERATING REACTORS BRANCH 4 US NUCLEAR REGULATORY COPNISSION WASHINGTON DC 20555 DOCKET 50-312 RANCHO SEC0 NUCLEAR GENERATING STATION UNIT NO 1 NUREG 0737 ITEM II.B.3 POST ACCIDENT SAMPLING SYSTEM (PASS)

During a conference call between District personnel and NRC' staff members on May 26, 1983, members of your staff requested clarification of infomation provided in our letter of May 2,1983 on this subject.

The enclosure to this letter provides the additional infomation requested.

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[U\\i R. J. Rodriguez Executive Director, Nuclear Enclosure i

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83062200 5 830617 I I PDR ADOCK 05000312

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__P PDR Ah ELECTRIC SYSTEM SERVING MORE THAN 600.000 IN THE HEART OF CAltf0RNIA

ENCLOSURE 1 SMUD RESPONSES TO NRC QUESTIONS ON POST-ACCIDENT SAMPLING SYSTEM-(PASS)

ITEM 1; Regarding Criterion 1:

NRC requested a date by which the District will complete the addition of a backup power source to allow sampling during a loss i

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of offsite power.

RESPONSE: Equipment has been-ordered which will allow powering the analyzer and solenoid valves from a reliable source within three hours of a loss of external power sources. Delivery of necessary equipment is anticipated in January,1984 and installation will be perfonned at I

the first outage of sufficient duration not later than the next refueling outage which is currently scheduled to begin in November,1984 Some PASS valves utilize pneumatic operators which receive motive power from instrument air compressors which are lost upon loss of offsite power. However, it will be possible to

, connect a diesel driven air compressor which is currently onsite to i

the instrument air header within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> of loss of external power if needed to support post-accident sampling.

l ITEM 2; Regarding Criterion 2 1

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. NRC requested an abstract of the District's core damage procedure RESPONSE: The procedure is currently under development by our consultant (Stone & Webster).- Currently no graphs or tables which will be included in the final procedure are available. An abstract of the procedure which is under development is provided as enclosure 2.-

A copy of our draft or final procedure can be provided when' i

available, if desired.

ITEM.3; Regarding Criterion 3:

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NRC requested additional supportive.infomation to support the District's conclusion that the PASS valves will operate satisfactorily in the accident environment.

RESPONSE: All valves required to operate and support the PASS inside containment are motor-operated and are qualified for the post-accident environment. Other valves, which were added 1

specifically for the PASS are pneumatic operated diaphragm valves with solenoid actuated air supplies. We have reviewed the results of the Plant Shielding Study perfomed in accordance with NUREG 0737_ item II.B.2.

The dose rates which were conservatively estimated by that study were integrated and the contribution from t

each of the active systems (e.g., HPI, LIP, Containment Spray, Waste Gas) was considered based on the time sequence into the accident for their relative contribution to the areas of the plant where the PASS pneumatic valves are located.

Based on this evaluationweesyimateanintegrated30dayaccidentdoseof approximately 10 Rads.

Since the valves were not purrhased as 4

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. ATTACHMENT 1 class 1, no environmental qualification documentation is available. However, based on infomation available in the industry literature for similar components, we have concluded that the PASS valves would operate as required for the accident duration should an accident occur.

ITEM 4; Regarding Criterion 10:

(1) NRC requested a definition of the range and accuracy for all instruments not previously stated in correspondence and a justification for a deviation from the ranges required by Regulatory Guide 1.97 Rev. 2.

(2) NRC requested that SMUD provide the basis for concluding that radiation levels which are expected to occur in accident conditions will not significantly alter the instrument accuracy over the required range.

(3) A third mquest was made by NRC during a subsequent conversation: to provide a comitment as detailed in the NRC letter of July 12, 1982 under criterion 10 to provide refresher training in use of the PASS for those personnel who will be assigned to operate the equipment.

RESPONSE

(1) The following ranges and accuracies have been providad by the equipment manufacturers:

Range Accuracy pH 0 - 14 units

+ 2% meter j

}1%deoutput B

0-6000 ppm

+ 50 ppm (0-1000 ppm)

}5%(1000-6000 ppm)

Cl-0 - 20 ppm

+ 5%

Total 0-2000 cc/kg

+ 100% (0-100 cc/kg)

Dissolved T 20% (100-1000 cc/kg)

Gas T 10% (1000-2000 cc/kg)

The ranges are consistent with the guidance provided in Regulatory Guide 1.97 Rev. 2.

(2) The District has estimated the dose expected at the location of analyzers in the same manner as described in our response l

to item 3.

Although the equipment was not purchased as L

class 1 and therefore no documentation exists to demonstrate the acceptability, we have concluded that the expected radiation levels will not significantly alter the accuracy of the PASS instruments as a result of the integrated accident dose after examination of available industry literature of effects upon similarly designed equipment.

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. ATTACHMENT 1 (3) The District will connit to provide refresher training involving use of the PASS equipment for a sufficient number of designated equipment users to insure the availability of trained personnel to support post-accident sampling requirements.' The frequency of such testing / training will be every six months + 25%.

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ENCLOSURE 2

. ABSTRACT OF SMUD CORD DAMAGE ESTIMATE PROCEDURE BACKGROUND The procedure, currently under development, will provide Rancho Seco personnel with a manual technique for estimating the extent of core damage utilizing data.obtained from the Post-Accident Sampling System (PASS), supplemented with supporting infomation from other station indicators (e.g., pH, conductivity,

'in-core temperature, etc.).

The results of the procedure will classify the fuel core damage into one, or more, of the following categories:

i No Damage l

Cladding Rupture Fuel Overheating Fuel Melting The last three categories are further subdivided into one of three ranges:

(1) less than 10%, (2) 10% - 50%; and (3) greater than 50%.

It should be emphasized that the procedure is intended for accident analysis only, and is not applicable-for detemining fuel cladding failure of less than or equal ta 1% (i.e., Tech. Spec. Limit).

Furthemore, the procedure is

' estimate the extent of fuel core damage. Therefore, it is being developed in

- intended to provide Rancho Seco personnel with an efficient technique to j

a " cook-book" manner relying heavily on graphs and straightforward calculational techniques. The detailed theoretical background calculations, and other supporting infomation will also be available for reference.

METHODOLOGY The basic methodology used in developing the procedure consists of the following:

A.~ Key plant parameter infomation (e.g., in-core temperature monitors, coolant levels, containment hydrogen indicator, etc.) has been evaluated and a procedure is being fomulated to establish (a) preliminary indication (s) of possible core damage.

B.

The Rancho Seco PASS is capable of obtaining representative diluted or undiluted samples of the containment air and reactor coolant during post-accident conditions. A radionuclide analysis will be perfomed on the PASS samples and the results will be nomalized, using equations in the procedure, for dilution, temperature, pressure, radioactive decay and plant power history to establish the extent of core damage.

Nomalized radionuclide concentrations will be developed for Xe133, I131, Cs134, Cs138, Rul03, Mo99, and other isotopes; with this data, an initial estimate of core damage can be made using a series of graphs.

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. ATTACHMENT 2 s

'!l If the results of the graphical analyses' indicate that there is core -

damage, anC the' results are supported by information obtaine 1 from other plant indicators, the next step 'n the procedure is to calet. late the extent 'of core damage, the first step is to determine the primary sourte,'

or sources, of activity by calculating'the rstfos of the key isotopes (e.g., Kr87 to Xe133 and I134 to I131)., With the activity ratios

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established..the procedure provides equations to determine the contributions from clad rupture and fue1Ldegradation (i.e., overheat and melt).

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Finally, the calculated resul,ts are compared with the results of other plant--indicators and an estimate of core damage based on all available plant infomation and 'radionuclide analyses is formulated..

C.

The major assumptions used in devel'oping' the p,rocedure is as follows:

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l 1.7 The extent of core damage is proportional to radioactivity released to the reactor coolant.

i3 2.

Onihcladrupture.andfueloverheat-melttakeplace.

Oxidation and vaporization considerations are omitted since a successfully terminated LOCA is assumed.

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bnifom' mixing of radionuclides in the sampled' medium is assumed. A periodic sampling analysis may be required to verify this assumption.

4.

10d5ofthenoblegasesareassumedtoescapethecoolantandare>

released to containment.

100% of all other isotopes including iodines are, assumed to remain in the coolant.

9 D.

The inpist. data used in developing the procedure are:,

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1. ' Cord inventory for Equivalent Full Power (EFP) activities.

2.

Release fractions for clait gap and fuel matrix, PWR irradiated fuel by chemical group (e.g., nob 1'e gases, halogens, alkali metals, noble

. metals)

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3.

Isotope'ratM i f. el and gap for noble gases and halogens (e.g.,

Kr87/Xe133 i. p/T 'f); _

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4.

Rancho Seco plant specffic data:

Design Power Fuel Cycle Reactor Building Volume Coolant Volume Containment Volume s

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