IR 05000214/2004016

Forwards Final Safety Evaluation Concluding That 850214-0416 Submittals Re post-accident Sampling Sys Meet Requirements of NUREG-0737,Item II.D.3 & Acceptable

ML20133D157
Person / Time
Site:	Yankee Rowe, 05000214
Issue date:	08/01/1985
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:	Papanic G YANKEE ATOMIC ELECTRIC CO.
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.3, TASK-TM LSO5-85-08-001, LSO5-85-8-1, NUDOCS 8508070249
Download: ML20133D157 (9)

Text

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* August 1, 1985 Docket No.

50-?9 LS05-85-08-001 Mr. George Papanic, Jr.

Senior Proiect Engineer-Licensing Yankee Atomic Electric Company 1671 Worcester Road Framingham, Massachusetts 01701

Dear Mr. Papanic:

SUPJECT: NUREG-0737 ITEM II.D.3 POST-ACCIDENT SAMPLING SYSTEM

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Re: Yankee Nuclear Power Station

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Enclosed is our final Safety Evaluation of your post accident sampling

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system (PASS) for Yankee. Our evaluation is. based on your submittals of February 14 and October 3,1983, and March 15 and April 16, 1985.

Based on the enclosed evaluation, we find that the Yankee PASS meets all the requirements of Item II.B.3 of NUREG-0737 and is, therefore, acceptable.

Sincerely,

44Aad signed by:

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John A. Zwolinski, Chief Operating Reactors Branch No. 5 Division of Licensing F

Enclosure:

Safety Evaluation cc w/ enclosure:

See Next Page

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DISTRIBUTION ~

Docket Files PVu

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ORB #5 Reading JClifford flRC PDR CJamerson L PDR OELD JZwolinski EJordan HThompson BGrimes Cdamerson JPartlow ACRS(10) 7

DL:0RB#5 DL:0RDf5[///kg DL:0RB#5 '

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CJamerson JClifford:jb' QZwolinski i/p/85 7/p/85 '\ /1 /85 g/[

op 8500070249 850001 PDR ADOCK 05000029 P ppg

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,s Mr. George Papanic, Jr.

Yankee Atomic Electric Company Yankee Nuclear Power Station .

cc:

Mr. James E. Tribble, President Yank e Atomic Electric Company 1671 Worcester Road Framinghan, Massachusetts 01701 -

Thomas Dignan, Esquire Ropes and Gray 225 Franklin Street Boston, Massachusetts 02110 Mr. N. N. St. Laurent Plant Superintendent Yankee Atomic Electric Company *

Star Route .

Rowe, Massachusetts 01367

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Chairnan -

Poard of Selectmen Town nf Rowe Rowe, Massachusetts 01367 Resident Inspector Yankee Nuclear Power Station ,

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c/o U.S. NRC Post Office Box 28 Ponroe Bridge, Massachusetts 01350 .

Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue .

King of Prussia, Pennsylvania 19406

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Robert M. Hallisey, Director Radiation Control Program Massachusetts Department of Public Health 150 Tremont Street, 7th Floor *

Boston, Massachusetts 02111 ,

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pn **:vq'o 4 UNITED STATES g

8 o NUCLEAR REGULATORY COMMISSION h ; W ASHINGTON, D. C. 20555

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SAFETY EVALUATION RY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO OPERATIOi! 0F YANKEE NUCLEAR POWER STATION

YANKEE ATOMIC ELECTRIC COMPANY DOCKET NO. 50-29 ,

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Post Accident Sampl kn System (NUREG-0737, II.B.3)

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1.0 ' Introduction ,,

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$usequenttotheTMI-2 incident,theneedwasrecognizedforanimproved post-accident sampling system (PASS) to determine the extent of core

'egradation d 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 the extremities (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 iodine of severity and cesium, of core damage and nonvolatile isotopes (e.g)., noble gases,

,, hydrogen in the isotopes containment atmosphere, and total dissolved gases or hydrogen, boron, and chloride in reactor coolant samples.

To comply with NUREG-0737, Item II.B.3, the licensee should (1) review and modify its sampling, chemical analysis, and radionuclide determination capabilities as necessary and (2) provide the staff with in'ormation pertaininn to system design, analytical capabilities and procedures in

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sufficient detail to demonstrate that the criteria are met.

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2.0. Evaluation

.By l'etters dated February 14 and October 3, 1983. with additional infomation nrovided March 15 and April 16, 1985, the licensee provided

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.,information on the PASS.

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-2-Criterion (1):

The licensee shall have the capability to promptly obtain reactor coolant samples and containment atmosphere samples. The combined time allotted for sempling and analysis should be three hours or less from the time a decision is nade to take a sample.

The PASS has sampling and analysis capability to promptly obtain and analyze reactor coolart sanples and containment atmosphere samples within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the time a decision is nade to take a sample.

We determined in our draft safety evaluation (SE), dated December 21, 1983, that the PASS partially met criterion (1). Additional information was provided in the licensee's March 15, 1985 letter, which included a

discussion of. alternate power supplies that could be energized in sufficient '

c. time, during a loss of offsite power, to meet the 3,-hour sampling and D analysis tire limit.

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Containment atmosphere samples can be obtained by using the pressure within the containment to provide the sample flow. If enough flow is not obtained, the hydrogen analy::er can be used to increase the flow. It is powered from a normal 480 V station service bus. With e less of offsite power, this -

bus can be manually conrected to an emergency 480 V bus, and powered directly by an emergency diesel generator. Procedures are in effect to perform this *

operation. The erercency power provided by the connection can be made available in ample time to meet the 3-hour sampling and analysis time limit.

The Post-Accident Sampling Panel (PASP) used to obtain the reactor

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coolant samples is also powered from a normal 480 V station service ,

bus. The connection arrangement discussed for the containment atmosphere samples is also availeble for emergency power for the PASP, and can be connected within the 3-hour sampling and analysis time limit. We find that these provisions meet Criterion (1) and are, therefore, acceptable.

Criterion (2):

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

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

h) hydregen levels in the containment atmosphere; c) dissolvedgases(e.g.,H,,), chloride (timeallottedfor analysis sub,iect to discbssion below), and boron concentration of liquids; d) Alternatively, have in-line monitoring capabilities to perfern all or part of the above analyses.

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The PASS provides the capability to collect undiluted liquid and gaseous grab samples that can be transported to the radiochemical laboratory for hydroaen, boron, chloride, and radionuclide analyses.

In our December 21, 1983 draft SE, we found that the licensee partially

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met Criterion (2) by establishing an onsite radiological and chemical analysis capability. Additional information provided in the licensee's March 15, 1985 letter, which provided a procedure, consistent with the staff's clarification of NUREG-0737, Item II.B.3, Post-Accident Sampling System, to estimate the extent of core damage based on radionuclide concentrations and taking into consideration other physical perereters such as core temperature data and sample location.

We find that these provisions meet Criterion (2) and are, therefore, acceptable.

Criterion (3):

Reactor coolant and containnent atmosphere sampling during post-accident conditions shall not recuire an isolated auxiliary system -

(e.g., the letdown system, reacto'r water cleanup system) to be placed in operation in order to use the sampling systen.

Reactor coolant and containment atmosphere sampling during post-accident conditions dces not require an isolated auxiliary system to be placed in operation in order to perform the sampling function. The PASS provides the ability to obtain samples from each reactor coolant cold leg, the residual heat removal system, the containment sump, and the centainment atmosphere without using an isolated auxiliary system.

The PASS valves which are not accessible after an accident are environmentally qualified for the conditions in which they need to operate. Ve find that these provisions meet Criterion (3) and are, therefore, acceptable.

Criterion (4):

Pressurized reactor coolant samples are not required if the licensee can quantify the arount of dissolved gases with unpressurized

.- reactor coolant samples. The measurement of either total dissolved cases or P, gas in reactor coolant samples is considered adequate.

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Measuring the 07 concentration is recommended, but is not nandatcry.

The PASS has the capability of obtaining pressurized reactor coolant samples. Additionally, the quantity of dissolved gases is determined as a normal part of the post-accident sampling procedure (OP-9450).

We determined that these provisions meet Criterion (4) of Item II.B.3 in NUREG-0737 and are, therefore, acceptable.

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-4-Criterion (5):

The time for a chloride aralysis 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 harrier between prinary 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.

In our letter dated May 26, 1981, we stated that the requirements of NUREG-0737. Item II.B.3, clarifications Nos. 2C and 5; to monitor chloride within N or 06 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (site dependent) is intended to provide information

,. .to the operator on the potential for chloride stress corrosion cracHog

.(CSCC) of the reactor' coolant stainless steel pressure boundary during .

the post-accident outage period. The two primary staff concerns are; a. CSCC during a long outage may affect integrity of a critical systen, b. During recovery, an assessment will be made of chloride /oxynen/pH history to determine the extent of examination required for CSCC, prior to approving a restart.

Due to the nultiple potential sources of chloride (plant cooling water, makeup water, chemical additives, resin degradation, etc.), we consider it likely that chloride contamination will exist at some point during the accident, as is the case at Ti1I-2 where 2-6 ppm chloride exists in the reactor coolant system. Therefore, the only means of assessing its effect is to be able to monitor chloride.

Provisfor.s ere incorporated in the licensee's PASS to obtain a sample of primary coolant and perform an analysis on the sample after an accident occurs. The licensea has comitted to perform a chloride analysis within the required 4 days after the accident. The licensee is purchasing an ion chronatography which will be delivered later in 1985. The testing will be completed after the fall 1985 refueling, .

and the procedures and documentation for chloride analysis should be available by ilune 1, lop 6. 1.'e find that these provisions and comitments 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 10any CFRird bidual Part exceeding 50) (i.e., the criteria 5 rem whole body, of 75GDC-19 (Appendix A,(Fote rem extrenities).

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.)

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-5-The licensee has performed a shielding analysis to ensure that operater exonsure while obtaining and analyzing a PASS sample is within the acceptable limits. This operator exposure includes entering and exiting the semple panel area, operating sanple panel manual valves, positioning the grab sample into the shielded transfer carts, and performirp manual sample dilutions, if required, for isotopic analysis. PASS personnel radiation exposures from reactor coolant and certainment 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 sarples for boron is required for PWRs. .

..d Boron analysis of the reactor coolent will be performed by standard.

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mannitol potentiometric titratinn nethod with a measurement range of 250 ppm to 10,000 ppm under accident conditions with an accuracy of about 5% at the lower concentration. We find that this provision meets Criterion (7) and is, therefore, acceptable.

Criterion (8):

If in-line monitoring is used for any sampling and analytical cepability specified herein, the. licensee shall provide backup sampling through grab sanples, 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.

No in-line monitoring instruments are used in the PASS. All analyses will be made on grab serples. We find that this provision meets Criterion (8) and is, therefore, acceptable.

Criterior (n): ,

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

i a) identify and quantify the 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 ,

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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 rance from approximately 1 Ci/g to 10 Ci/g.

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-6-b) Restrict background levels of radiation in the radiological and chemical analysis facility from sources such that the sample analysis will provide results with er ecceptably 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 airborne radioactivity.

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

The PASS can perform radioisotope analyses at the levels corresponding to the source term given in Peculatory Guides 1.4, Rev. 2,and 1.7.

• Radiation background levels will be restricted by shielding and ventilated in the radiological and chemical analysis facilities such .

. that analytical results can be obtained within an acceptably small a error (approximately a factor of 2). We find 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 systems.

The accuracy, range, and sensitivity of the PASS instruments and analytical procedures, under normal circumstances, are consistent with the recorrendations of Regulatory Guide 1.97, Rev. 2, and the clarifications of flVREG-0737. Item !!.B.3, Post-Accident Sampling Capability, transnitted to the licensee on June 30, 1982. In our December 21, 1983 draft SE), we found that this information partially metCriterion(10). The licensee provided additional information in #

its April 16, 1985 letter that addressed the equipment and procedures used for post-accident sampling.

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The orly chemical analysis currently performed by YAEC on a liquid sample during PASS operations has been evaluated using the constituents

'of the standard test matrix, with the exception of the induced gamma radiation field. The results of the evaluation indicate the rethodology meets the Preuracy requirements of Criterion 10.

The effect of an induced gamma radiation field (10 rads /gm of coolant)

on the analytical method has beer independently evaluated. The pH electrodes of the type used in tbc mannitol7 titration method for boron have been shrwn to be stable up to 10 rads of cumulative

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-7-radiation exposure with higher levels tolerated with appropriate calibration. Based upon the diluted matrix source term, the PASS (

sample volures at YAEC and times of analysis, the cumulative exposures will be substantially less than 10 rads.

Currently, PASS operators receive initial training and semiannual refresher training in post-accident samplino, transport, and analysis.

Testing is a part of this training. We find that these provisions meetCriterion(10)andare,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 for purging samp1.e lines, for reducing plateout

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• in sample lire, for ninimizing sample loss or distortion, for preventing blockage of sanple lines by loose material ir the RCS or containment, for appropriate disposal of the samples, and for fiev restrictions to limit reactor coolant loss frcm a rupture of the sample line. The post-accident reactor coolant and containment atmosphere samples should be represen-tative of the reacter ecolant in the core area and the contain-nent atnosphere following a transient or accident. The sample lines should be as short as possible to ninimize the volume of fluid to be taken fron 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 to ensure samples are representative, size of sample line to limit reactor coolent loss fron a runture of the sample line, and ventilation exhaust from PASS fil.tered through HEPA filters. To linit iodine slateout, the containment air sanple line is heat traced. We deternined t1at these provisions

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meet Criterion (11) of Item !!.B.3 of NUREG-0737, and ar,e, therefore',

acceptable.

3.0 Con,clusion on Based on the above evaluation, we conclude that the post-accident sampling system meets all the requirenents of Item !!.B.3 of NUREG-0737 and is, therefore, acceptable.

4.0 Acknowledoment This Safety Evaluation was prepared by P. Wu.

Dated: August 1, 1985.