ML20009E345

Transcript of G Martin Testimony on Behalf of Util Re Doherty Contention 40,10CFR100 Releases.Prof Qualifications Encl

ML20009E345
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	07/20/1981
From:	Martin G EBASCO SERVICES, INC., HOUSTON LIGHTING & POWER CO.
To:
References
NUDOCS 8107280066
Download: ML20009E345 (10)

Text

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. July 20, 1981

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UNITED STATES OF AMERICA  !; ,

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p G 3 0.?? ICE OF APPLIC.U n. ., fiUCLEAR REGULATORY COMMISSION i, 4 bbhb5-THE ATOMIC SAFETY AND LICENSING BOARD'- gQ7 g }

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6 H USTON LIGHTING & POWER COMPANY ) Docket No. 50-466 (Allens Creek Nuclear Generating ) @ N

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- ) p, 8 --i 27198 TESTIMONY OF GUY MARTIN, JR. ON PEHALF 1 p* 7 T 9 HOUSTON LIGHTING & POWER CO. Oh DOHER v.5. $sso'* ,3 CONTENTION 40 PART 100 RELEASES ,, / x 10

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11 Q. Please state your name and job position.

12 A. My I. me is Guy Martin, Jr. I am employed by 13 Ebasco as Supervising Engineer of Envirosphere's Radiological g Impact Assessment Department.

Q. Please state your educational background, work experience and professional qualifications.

A. A statement of my qualifications is attached to 17 this testimony as Attachment GM-1.

18 Q. What is the purpose of your testimony?

19 A. My testimony will address Doherty Contention 20 40, in which it is alleged that:

21 "

...the Allens Creek site is unsuitable for the 2,~ proposed nuclear plant, because the assumed fission product release from any accident considered credible will exceed the limitations of radioactivity dose 2a.

to the lcw population zone stated in 10 CFR 100.11, y03 24 (a) (1), (2), and (3). s Ii(

8107200066 810720 PDR ADOCK 05000466 T PDR __

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This Intervenor contends this because the 3 actual release of radioactivity from the Three Mile Island accident exceeded calculated release for any accident considered credible by a factor of 22, using the calculation suggestions of Regulatory Guide 1.4. The proposed ACNGS and Three Mile 5 Island are sufficiently similar in design such that the miscalculation in the ill-fated reactor's case 6 is the same for the subject of these proceedings, '

in regard to source terms, and other factors.

7 This contention is particularly relevant to 8 ACNGS construction license proceeding because the Applicant's proposed NPSS will use the largest BWR 9 core attempted,with the highest power core density, and greater minimum critical heat flux ratio than 10 any functioning BWR plant. Construction of the plant at the proposed site will injure Intervenor's 11 health and safety interest by exposing him to radiation in excess of the guidelines of 10 CFR 12 100.11."

Q. Are the TMI and Allens Creek designs substantially 13 sin l.lar as Intervenor Doherty alleges?

4 A. No. TMI is a Babcock and Wilcox PWR with a 13 reactor and containment design substantially different from the General Electric BWR at Allens Creek. These 17 differences will result in different isotopic source 18 terms and different containment release rates during an 19 accident. These differences are reflected in the fact that the NRC uses different Regulatory Guides to evaluate 21 the radiological consequences of a design basis accident (LOCA) for BWR's and PWR's (Regulatory Guide 1.3 addresses 23 BWR's; Regulatory Guide 1.4 addresses PWR's).

24 F. .

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Q. The Intervenor has based his allegation that 3

radiation releases for TMI were 22 times greater than 4

estimated on Board Notification BN-79-23. Would you 5 discuss your evaluation of BN-79-23?

6 A. BN-79-23 estimated the radioactive releases 7 from TMI to be 13 million curies of Xenon-133. The 8 concern raised by the notification was that it had been 9 previously estimated on the basis of Regulatory Guide 10 1.4 that the amount of Xe-133 released during a design 11 basis accident would be 600,000 curies. Thus, the actual release of this isotope was reported to have been 2

ubstantially in excess of that predicted under the 3

Regulatory Guide.

Q. Were 10 CFR 100 limits in fact, ever exceeded 15 for the Three Mile Island plant?

16 A. No. A detailed study in NUREG-0558 indicater 17 that an average dose of only 1.5 millirem was received 18 by the population surrounding TMI during the entire 19 course of the incident. The study also indicated that 20 the maximum estimated dose to one individual outside the

~l exclusion area was less than 100 millirem, or 1/250th of 22 the 10 CFR Part 100 limits.

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Q. Please explain how the number of curies of 3

Xenon-133 could be larger than obtained under Regulatory 4

Guide 1.3 or 1.4 analyses, yet result in doses still 5 acceptable and lower than 10 CFR 100 limits.

6 A. Because i* assumes immediate release of fission 7 products to the containment atmosphere, the model used 8 by Regulatory Guides 1.3 and 1.4 is a conservative 9 estimate of doses that could result from a design basis 10 aceident with a high degree of core damage. This arsump-11 tion results in a higher ratio of shorter lived hich 12 energy gamma emitting noble gases such as Krypton-88 being released. At TMI, releases to the environment did 13 not occur until several hours into the incident, thus 4

allowing a significant decay time for the short lived lo_

high energy isotopes and resulting in a higher ratio of the Icw energy gamma emitter, Xenon-133, being released.

17 Although the total number of curies released from TMI 18 exceeded that of the Regulatory Guide source term, the 19 effect of the TMI source term was less because the high 20 energy gamma emitting isotopes are the major contributors 21 to the total dose.

22 Has the NRC calculational method, that is Q.

23 Regulatory Guides 1.3 and 1.4, with respect to post-accident 24 1

2 doses, been revised since the TMI incident to reflect 3

different calculational techniques due to the higher 4

than estimated release of XE-133 at TMI?

5 A. No.

6 Would you please discuss the method used for Q.

7 determining Allens creek's compliance with 10 CFR 100 8 limits?

9 A. The analytical procedure used to determine 10 such compliance is detailed in Regulatory Guide 1.3 and 11 in the Allens Creek PSAR Section 15.1.39. Basically the 12 method assumes that 25 per cent of the iodine and 100 p r cent of the noble gas core inventories developed 13 during aquilibrium maximum full power operation are immediately released to the containment and are available for leakage from containment. Next the containment is assumed to leak at the maximum rate allowed by the plant 17 technical specifications for the duration of the accident.

18 Then plant specific values for meteorology are applied 19 to determine the resulting doses.

Q. Does Allens Creek meet the 10 CFR Part 100 91

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'imits?

22 A. Yes. Using the methodology of Regulatory 23 Guide 1.3 the resultant doses are giver ,,TJ4R Table 24 15.1.39-3 and summarized below:

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1) The individual 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose at the exclusion 3

area is 4.9 rem whole body and 150 rem thyroid t'

2) The individual accident duration doce at the 5 low population zone is 1.2 rem whole body and 6 71 rem thyroid.

7 Part 100 establishes limits of 25 rem whole 8 body and 300 rem thyroid for either an individual two 9 hour done at the exclusion area boundary or an individual 10 accident duration dose at the low population zone. The 31 above doses are well within the 10 CFR 100 limits.

12 In addition, it should be noted that gross 13

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required to achieve the fission product inventories in containment that are assumed in Regulatory Guide 1.3.

16 An analysis has also been completed for a less 1 17 i

conservative amount of fuel damage and resultant 18 fission product release that would occur during a

, 19 design basic accident. Under this more realistic 90

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analysis (PSAL Table 15.1.39-3) the following doses 21 would be received:

1) Individual 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the exclusion area 23 -6 -5 of 5x10 rem whole body and 3.4x10 rem 24 thyroid.

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2) Individual accident duration dose at the low population zone of 9.8x10 -7 rem whole body and

-5 l.8x10 thyroid.

5 Q. How was the large amount (approximately 13 6

million curies) of Xe-133 released to the environment at 7 TMI?

8 A. According to the Kemeny Commission Report, 9 most radioactivity escaping from TMI-2 to the environment 10 was in the form of fissien gases transported through the 11 reactor coolant let-down/make-up system into the auxiliary 12 building and through the building filters, then to the vent header and to the outside atmosphere. The major 13 release of radioactivity on the morning of March 30 was caused by the controlled, planned ',enting of the make-up la_

tank into the vent header.

16 Q. In a scenario for radiot.ativicy release of 17 this type included in the dose assessment analyses of 18 Regulatory Guides 1.3 or 1.4?

l 19 A. No. Containment isolation is assumed to

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occur, and the containment is then assumed to leak at 91

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the maximum allowable plant technical specifications 22 leak rate.

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Q. Are any design improvements incorporated in 3

the Allens Creek design that would preclude the occurrence 4

of a release such as that which occurred at TMI?

5 First of all, ACNGS does not utilize a coolant A.

6 let-down/make-up system, as provided at TMI, so a release 7 path of this type is not possible at ACNGS. Moreover in 3 rerponse to the TM: incident, design modifications were 9 developed for the containment isolation system. These 1C modifications are required by NUREG-0718, " Licensing 11 Requirements for Pending Applicttions for Construction 12 Permits And Manufacturing License." As detailed in S tions II.E.4.2, II.E.4.4, and III.D.l.1 of the ACNGS 3

PSAR, Appendix 0, Allens Creek has incorporated all suggested modifications, such as containment isolacion for non-essential systems, that were not already part of 16 the plant design.

17 These design modifications were developed in response 18 to the TMI incident to contain radioactive contaminants 19 within the containment building. Incorporation of these 20 modifications will help assure tP.t releases such as 2^1 occurred at TMI will not occur at Allens Creek.

2~7 Q. What are your conclusions?

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A. The release mechanism which occured at TMI 3

cannot be duplicated at Allons Creek, nor will the 4

estimated doses from a design basis accident at Allens Creek exceed the 10 CFR Part 100 limits.

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10 11 12 13 14 15 16 17 10 19 20 21 22 23 24 g_

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Attachment GM-1 Guy Martin, Jr.

I received a ME from the City College of the City of New York in 1974. I received a MS in Nuclear Engineering from Polytechnic Institute of New York in 1976. I have been employed by Ebasco since 1973. I have eight years' experience in preparation of Safety Analysis and Envirion!; ental Reports sections dealing with the impact analysis of toxic chemicals and radiological releases. Such analyses are performed for both routine plant operation and accident conditins. In this regard, I conduct reviews of radwaste handling systems, air handling and cleanup systems and estimate radionuclide releases from plant eff3uents and calculate and calculation of implant dose rates to equipment and personnel from air borne radionuclide exposure and I have performed ALARA of air cleanup systems. I have performed safety reviews of engineered safety systems, which included a review of the specifications and operation from the radiation protection viewpoint and have provided design recommendations based on assessed radiological doses and established nuclear safety criteria. I have performed analyses of the transport of

toxic chemicals postulated to be released accidentally and calculated the concentration in critical locations of the power plant. I have provided technical feedback to the designers on required protection levels. In this regard I have assisted in making the determination of toxic chemical detector specifications based on worker and equipment protection criteria.

I have responsibility for the preparation of radiological envirionmental surveillance programs wherein I have prepared detailed surveillance program description based on site specific critical pathways of exposure. I have established the sampling requirements of the frequency and types of analyses to be performed.

I have also participated in preparation of a study regarding the establishment of a comprehensive data base regarding high level waste disposal and I have supervised the health physics activities related to decontamination work at the Kellex Labcratory.

Prior to my employment with basco, I was employed as a cost analyst by Equitable Life Assurance Socity of the US.

r. . .

1 I am a member of the American Society of Mechanical' Engineers, a member of the Health Physics Society, and a member of the American Nuclear Society, and Intern Engineer of New York State. I have written the following publfcations:

Marti., G. and J. Thomas 1978. Meeting the dose requirements of 10CFR100 for site suitability and general design criteria 19 for control room habitability: a parametric approach.

Transactions of American Nuclear Society 24th Annual Meeting.

Vol. 18.

Martin, G. D. Michlewicz and J. Thomas 1978. ~ Fission 2120:

a program for assessing the need for engineered safety feature grada air cleaning systems in post-accident environment.

Proceedings of 15th DCE Nuclear Air Cleaning Conference.

Letizia, A. P., G. Martin and J. F. Silvey 1979. - Implice.tions for nuclear facilities of changes being initiated in the NRC standard atmospheric diffusion model. Proceeding of the 41st Annual Meeting of the American Power Conference.

Bhatia, k. K., Mauro, J., Martin, G. - Effections of Containment Purge on the Consequences of a Loss-of-Coolant Accident.

Transactions of the American Nuclear Society 1980 Annual Meeting.

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