ML17258A193

Forwards Draft Evaluation of Radiological Portion of SEP Topic XV-19, LOCA Resulting from Spectrum of Postulated Piping Breaks within Rcpb. Removal of 2-minute Delay Timer Recommended

ML17258A193
Person / Time
Site:	Ginna
Issue date:	09/29/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-15-19, TASK-RR LSO5-81-09-072, LSO5-81-9-72, NUDOCS 8110020357
Download: ML17258A193 (11)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 September 29,=1981 Docket No. 50-244 LS05 09-072 Mr. John E. Maier Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649

Dear Mr. Maier:

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SUBJECT:

SEP TOPIC XV-19, LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECTRUM OF POSTULATED PIPING BREAKS WITHIN THE REACTOR COOLANT PRESSURE BOUNDARY - R. E.

GINNA Enclosed you will find our draft evaluation of the radiological portion of SEP Topic XV-19.

This evaluation compares your facility as described in

'ocket No. 50-244 with the criteria currently used for licensing new facil-ities.

Please inform us within 30 days if your. as-built facility differs from the licensing basis assumed in our assessment.

If no comments aie re-ceived within 30 days, we will assume the topic is complete.

You will notice that the assessment recommends removal of*the two minute delay timer which exists in the spray additive system.

This is due to the fact that the staff is concerned that the operator has been trained that the timer exists to allow overriding the addition of NaOH when it is not needed and would incorrectly turn off the additive when it is needed.

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An incorrect assessment of the extent of an accident and of.the need for NaOH additive either to wash down iodines in whatever form they appear or to keep iodines in solution in the containment sump is not an unlikely oc- //I currence.

For this reason, the SRP was revised to require the start of spray and addition of additive "without mechanical delays or manual over-"

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,Sii0020357 Si0929 "PDR ADOCK 05000244 Ip PDR OFFICE/

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OATE 0 I4RC FORM 318 (10.80) NRCM 0240 OFFICIAL R ECOR D COPY USGPO: 19&1 335-960

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This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to r eflect the as-built conditions at your facility. 'This assessment may be'=revised in the future if your facility design is changed or if NRC.criter ia relating to this subject are modified before the integrated assessment is completed.

Sincerely,

Enclosure:

As stated'ennis M. Crutchfield, Chief Operating Reactors Branch No.

5 Division of Licensing cc w/enclosure:

See next page OfFICEQ SURNAME/

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Mr. John E. Maier CC Harry H. Voigt, Esquire LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N.

W.

Suite 1100 Washington, D. C.

20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 World Trade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza

Albany, New York 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West

Ontario, New York 14519'esident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503 Lake Road

Ontario, New York 14519 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street,.N.

W.

Suite 600 Washington, D. C.

20006 C

U. S. Environmental Protection Agency Region II Office ATTN:

EI S COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,

Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comoission Washington, D. C.

20555 Dr. Ri char d F. Cole Atomic Safety and Licensing Board U. S; Nuclear Regulatory Commission Washington, D. C.

20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.

20555

R.E. Ginna Nuclear Power Station W

XV-19 LOSS OF COOLANT ACCIDENTS RESULTING FROM SPECT UM OF POSTULATED PIPING i

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I.

INTRODUCTION Loss-of-coolant accidents (LOCA's) are postulated breaks in the reactor coolant pressure boundary resulting in a loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system.

LOCA's result in excessive fuel damage'or melt unless coolant is replenished.

Excessive fuel damage can result in significant radiological consequences to the environment via leakage from the containment.

SEP Topic XV-19 is intended to assure that the radio-logical consequences of a design basis LOCA from containment leakage and leakage from engineered safety features outside containment are within the exposure guideline values of 10 CFR Part 100.

II. REVIEW CRITERIA Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction

'permit or operating license provide an analysis and evaluation of the design and performance of structures,

systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility.

The LOCA is one of the postulated accidents used to evaluate the adequacy of these structures,

systems, and components with respect to the public health and safety.

In addition, 10 CFR Part 100.11 provides dose guidelines for reactor siting

.gainst unicb calculated accident dose consequences may be compar d.

III. RELATED SAFETY TOPICS Topic II-2.C, "Atmospheric Transport and Diffusion Characteristics for Accident Analysis" provides the meteorological data used to evaluate the offsite doses.

Topic III-5.A, "Effects of Pipe Breaks on Structures, Systems and Components Inside Containment" ensures that the ability to achieve safe shut-down or mitigate the consequences of an accident are maintained.

Various other topics examine such areas as contai:nment integrity and isolation, post accident chemistry, ESF

systems, combustible gas control and control room habitability.

IV.

REVIEW GUIDELINES The review of the radiological consequences of a LOCA was conducted in accordance with the Appendices A, B, and C to Standard Review Plan 15.6.5 and Regulatory Guide 1.4 with the exception noted below.

The plant is adequately designed against a

LOCA and the dose mitigating features are acceptable if the resulting doses at the exclusion area and low population zone boundaries are within the guideline values of 10 CFR Part 100.

V.

EVALUATION The assumptions used in this evaluation are listed in Table XV-19-1.

In addition to decay, radioiodines were assumed to be removed from the containment by sprays and recirculation ventilation filter systems.

These latter removal processes are assumed to be applicable to 78Ã of the containment volume.

Leakage to the environment through the containment and through the,r esidual heat removal system were modeled; the. purge valves were assumed to be closed since the licensee has coranitted to purging less than 90 hour0.00104 days <br />0.025 hours <br />1.488095e-4 weeks <br />3.4245e-5 months <br />s/year in a letter dated 12/14/79.

Credit was given for removal of iodine by the recirculation filter

system, and by the spray, which was terminated when 99/ of the elemental iodine in the sprayed volume had been removed.

However, Ginna presently has a delay timer on the NaOH addition to the containment spray.

This is contrary to Section 6.5.2 of the Standard Review Plan, which requires automatic initiation of containment spray and spray additive "without mechanical delays or manual overrides."

This evaluation was done assuming the two minute delay prior to addition of NaOH to the boric acid spray.

The FSAR states th'at the temperature of water leaking from the recirculating system in the auxiliary building,would be less than 200 F; it was conserva-tively assumed that 10K of the contained radioiodines would be released to the auxiliary building atmosphere.

The EAB and LPZ thyroid and whole body doses are given in Table XY-19-2.

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CONCLUSIONS

.,',The plant',is,.:adequately designed against a

LOCA and the dose mitigating r.

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', features are acceptable, However, since the exi'stence of the two minute de-

, lay,'imer, is contra'ry to current Regulatory Criteria, we recommend that the rr JC licensee remove 'the delay timer from the spray injection system.

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' the delay timer would prevent inadvertant overriding of the containment spray system'n the event of an actual loss-of-coolant accident.

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TABLE XV-19-1 Assumptions Made in Analysis of the Radiological Consequences of the Loss-of-Coolant Accident 1.

Reactor Power - 1551 tQ 2.

Release Fractions 25K Radioiodines 1005 Noble Gases 3.

Containment Leakage Fraction 0.2/ / day First 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 0.1X / day Remainder of accident 4.

Recirculation loop leakage 2 gal /,hour 5.

Seal failure leakage (at 24 hrs) 50 gal / min 6.

Atmospheric Dispersion Factors (see Topic II-2.C)

Time Period 0-2 hour 0-8

,nour 8-24 hour 1-4 day 4-30 day Distance EAB LPZ L'PZ LPZ LPZ x/0 sec/m3 4.8 x 10 5

3.0 x 10 2.1 x 10 8.6 x 10 6

2.5 x 10""

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2.06 7.

Total iodine removal coefficients i 0 -

2 min

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'. 55 min - 30 days 8.

Containment free volume 9.97 x 10 ft 5

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4. 53 4.53 TABLE XV-19-2 EAB AND LPZ DOSES FOR DELAYED ADDITION OF NaOH EAB

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Containment Leakage Contribution ECCS Leakage Contr ibution Total Doses 110 20 130 3.0 0.1 3.1 21 10 31 0.5 0.0 0.5

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