ML19347F535

Testimony of Rl Baron Re Tx Pirg Addl Contention 21, Occupational Exposure.Prof Qualifications Encl.Related Correspondence

ML19347F535
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	05/11/1981
From:	Baron R HOUSTON LIGHTING & POWER CO.
To:
Shared Package
ML19347F516	List: ML19347F515 ML19347F520 ML19347F522 ML19347F524 ML19347F527 ML19347F529 ML19347F530 ML19347F531 ML19347F532 ML19347F534 ML19347F535 ML19347F536 ML19347F537 ML19347F539
References
NUDOCS 8105190556
Download: ML19347F535 (19)

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. 5-11-81 B_K.LEQie N N _

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"'" D UNITED STATES OF AMERICA .

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NUCLEAR REGULATORY COMMISSION O MAy y 3 ggg C i

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BEFORE THE ATOMIC SAFETY AND LICENSING BOARD g Dnice @hc ht

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' 4 In the Matter of S D

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5 HOUSTON LIGHTING & POWER COMPANY S Docket No. 50-466 S

6 (Allens Creek Nuclear Generating S Station, Unit 1) S 8 DIRECT TESTIMONY OF ROBERT L. BARON ON TEXPIRG ADDITIONAL CONTENTION 21 9 ON OCCUPATIONAL EXPOSURE 10 Q. Please state your name, your employer and your job 1.1 title.

12 A. My name is Robert L. Baron. I am employed by 13 Houston Lighting & Power Company as Supervising F.calth 14 Physicist.

15? Q. Please describe your educational background, work i 16 experience and professional qualifications.

_7 A. The statement of my background and qualifications 18 is attached as Exh2 hit RLB-1 to this testimony.

19 Q. What is the purpose of your testimony?

20 A. The purpose of this testimony is to addresc 21 TexPirg Additional Contention 21 which alleges that the 22 Final Supplement to the Final Environmental Impact Statement 23 (FSFES) underestimates projected radiation exposuLe to 24 workers at ACNGS.

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J _W" l Q. What is the projected occupational dose for workers 2 at ACNGS?

3 A. The Staff has estimated for Allens Creek a pro-4 jected occupational dose of 500 man-rem per year. (See 5 FSFES, p. 5-29). This is based on past experience from 6 operating light water reactor power plants similar in size 7 and type to the Allanr. Creek der,ign.

8 Q. TexPirg's contention implies that the estimate in 9 the FSFES is in err 3r because it is lower than actual cumula-10 tive doses cited in an article in Nucler: Engineering Inter-11 national (February, 1979). Is it proper to compare the 12 FSFES estimate with the recorded doses in the article cited 13 by TexPirg?

14 A. No. They are not directly comparable. All of the ,

15 recorded measurements in the Nuclear Engineering article are 16 stated in terms of a ratio between radiation dose and elec-17 tricity generated (man-rem /GW (e) year). The estimates in 18 tne FSFES are based on actual recorded annual doses measured 19 in man-rem per reactor year. The difference in units invali-20 dates any comparison between the two. This is most obvious 21 if one considers that there is generally an inverse rela-22 tionship between power production and accumulated dose. That 23 is, worker dose increases the more the plant is out of 24 operation cocause plant maintenance and modification is

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1 accomplished while the plant is shut down. Since the 2 numerator will be largar, and the denominator smaller, the 3 ratio of man-rem /GW(e) year will always be significantly 4 higher than the ratio of man-rem / reactor year.

5 Q. Leaving aside TexPirg's confusion as to the units 6 of occupational dose, is there any other evidence in the 7 Nuclear Engineering International article which indicates 8 that the estimate for Allens Creek is understated?

9 A. TexPirg's reliance on this article also fails to  !

10 account for the fact that the majority of the dose at operat-11 ing' plants is caused by plant maintenance and modification, 12 as explained above. As I describe later, the new generation 13 of plants such as Allens Creek have been specifically de-14 signed with this history in mind, and many steps i. eve been -

15 taken to speed maintenance and to provide better shielding 16 for workers during maintenance, modifications and daily 17 Operations. Accordingly, I believe that actual dose ratea 18 in the future, measured in terms of man-rems / reactor year, 19 will be less than the estimate in the FSFES.

20 0 How do past recorded total accumulated annual 21 doses compare with the estimate in the FSFES?

22 A. The doses compare quite well for an obvious reason.

23 The FSFES clearly states that the figure of 500 man-rem was derived from historical experience. The data available at 24 l

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1 the time the FSFES was published shows that actual recorded 2 exposure from 1969 to 1976 ranged from 178 to 594 man-rem /

3 reactor year for LWR's. This information is contained in 4 " Occupational Exposure At Light Water Cooled Power Reactors",

5 NUREG-0323, p. 4 (1976).

6 Q. Is there more recent information on recorded 7 exposures since the publication of NUREG-0323?

8 A. Yes, there is. The latest source document pub-9 lished by the NRC for data on occupational dose is NUREG-10 0594, " Occupational Rediation Exposure at Commercial Nuclear 11 Power Reactors 1978". This report is a compilation of 12 occupational dose derived from reports submitted to the NRC 13 in accordance with Part 20.407 of Title 10, Chapter 1, Code 14 of Federal Regulations and plant technical specifications.

15 This edition (Nov. 1979) contains data from 64 light water

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16 reactors which had completed at least one year of commercial 17 operation as of December 31, 1978. Accordrag to this re-18 port, the average collective dose per recctor for LWR's was 19 497 man-rem, which is very close to the NRC estimate of 500 20 ran-rem for ACNGS.

21 Q. How do the newer, larger pl, ants compare to the 22 average?

23 A. For BWR's with rated capacity equal to or greater 24 than 750 FM, the average man-rem occupational dose per unit f

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1 per year is 510 man-rem. For BWR's of greater than 1000 MW, 2 the average occupational man-rem dose per unit per year is 3 Obviously, these numbers support the 500 man-460 man-rem.

4 rem per year estimate for Allens Creek. It should be stressed 5 that we believe that projections based on historical data 6 produce S.igher dose rates than will actually occur at ACNGS

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7 because nore of these plants received the manageracnt emphasis 8 and manpower dedication to the "As Low Ae Reasonably Achieve-9 able" (ALARA) philosophy during their design phase which 10 ACNGS has already received and will continue to receive.

11 Q. Is ALARA a new concept in reactor design?

12 A. Yes. In fact, the ALARA design philosophy was 13 adopted by the NRC because of the concern over increased 14 worker exposure dose at older plants which had not been de-15 signed to minimize worker dose d~ ring plant maintenance, 16 modification and daily operations.

17 Q. Based on industry experience, do you expoet that 18 annual doses to workers will continuously increase over the 19 life of the Allens Creek plant?

20 A. No. The dose will probably be far below 500 man-21 rem initially. The dose will increaue,, reach a plateau below 22 500 man-rem, and may then begin to decrease. For example, over l

23 the last recorded year (1978), the latest year of operation for 24 which there is data, six out of eight of the BWR plant sites l

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1 referenced reported decreases in occupational doce.

2 Q. What is being done by HL&P to reduce the dose rate 3 to a value bwlsw the 300 man-rem / year projected by the NRC?

4 A. Cumulative occupational dose at nuc' ear power 5 plants can be controlled or reduced by any mears which 6 either limits the amount of dose received and/or reduces the 7 number of people exposed to radiation. The ACNGS design is 8 thoroughly reviewed by the HL&P Health Physica Division to 9 ensure that expoeure to personnel at ACNGS will be As Low As 10 Reasonably Achievable. HL&P has an ALARA training program 11 designed to indoctrinate its engineering and health physics 12 personnel to the ALARA philosophy. This practice will 13 assure that ALARA is given appropriate consideration in 14 those areas for which engineering and health physics person-15 nel are responsible. The program consists of an ALARA 16 training course, with refresher courses on an annual basis.

17 Additionally, the engineering and health physics members of 18 the ACNGS team are instructed to follow the "ALARA Design 19 Review Manual", a dccument containing ALARA guidelines to be 20 considered for incorporation on the plant c'csign. Training 21 also includes visits to ope:., ting power plants by various 22 ACNGS personnel. This practice creates familiarity with the 23 operational and maintenance aspects of nuclear plants and u

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I 1 design more attuned to the operational aspects of the plant.

2 Q. How is Ebasco, the Architect-Engineer, assisting 3 HL&P in meeting the goal of reducing exposure?

J 4 A. Ebasco, with its many years of experience in the 5 desigt. of nuclear power plants and in its own commitment to 6 the ALARA design philosophy, is actively seeking new ways to 7 reduca exposure at the site. Key Ebasco/ACNGS engineere 8 make trips to plants in operation to observe actual problems 9 in plant design. From these observations they are able to 10 incorporate design resolutions into ACNGS. Additionally, 11 Ebasco has ALARA cr.gineers who are responsible for reviewing 12 the Allens Creek design. Comments issued by the ALARA 13 engineers are submitted to the Ebasco/ACNGS design team for 14 resolution. Thus, the ALAP.A engineers are actually inte-

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15 grated into the group of engineers and designers working on 16 the ACNGS design, which assures that ALARA consideratior.3 17 are addrescad at an early stage of the design. Finally, a 18 scale model of ACNGS is being built by Ebasco. It is used 19 by Ebasco's and HL&P's personnel to develop a visual im-l 20 pression of potential problem areas before and during con-21 struction of the plant. ,

22 Q. How is General Electric, the NSSS supplier, assist-23 ing HL&P in meeting these goals?

24 A. General Electric is also involved in the ALARA

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1 design process. Its experiences, which include plant design 2 and operational experiences, are reflected in its newer 3 In addition, GE is con-generations of the NSSS system.

4 ducting on-going studies to better the reliability of these 5 Many items discussed below are design improvements systems.

6 resulting from General Electric's efforts. ACNGS has re-7 ceived and will continue to receive the benefits of these 8 efforts.

9 Please list some examples of the design improve-Q.

10 ments that have been made in an effort to try to reduce 11 worker exposure.

12 A. A few examples of design improve.nents which will 13 result in lower expecures are:

14 Underwater Steam Separator Transfer:

15 The BWR steam separator has to be transferred from the 16 vessel to a storage location at the beginning of refueling 17 and returned to its original position at the end of the 18 refueling. Typically, at other plants, these components are

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19 fully exposed when transferred. The design of ACNGS a> lows i

20 them to be moved underwater to provide shielding for the 21 operators and other personnel during the transfers, thus j 22 substantially reducing exposure to personnel.

23 Reactor Cavity C3eanup:

24 Suspended particulates (crud) in the reactor cavity

1 water are deposited on the cavity wall surfaces, leaving the 2 surfaces contaminated. Cleaning of cavity walls and floors 3 following fuel transfer and prior to final vessel head 4 closure is a manual operation. Many plants use a low polish 5 on the stainless steel cavity liner, which will prolong the 6

• time of the cleanup operations. ACNGS will use a higher 7 polish on the liner which reduces the time required for 8 de' contamination and therefore the dose to personnel.

9 Radwaste Evaporator Maintenance:

10 The radwaste evaporators at many existing plants use 11 type 304 stainless stee), which is subject to chloride 12 stress corrosion cracking. This has prompted the need for 13 replacement of individual components and piping as leaks 14 appear. Moreover, most of the evaporators in use today are 15 skid mconted, with the major components located in pro.ximity 16 to each other on a single skid. This closeness of-components, 17 coupled with the large amount of maintenance, has resulted 18 in a large personnel exposure at most plants. The waste 19 evaporators at ACNGS will use inconel tubes which are more 20 resistant to corrosion, thereby reducing maintenance ac-21 tivities. Also, the components of the evaporators are 22 located in separate shielded rooms to improve access and 23 shielding for maintenance personnel, thereby reducing ex-24 posure.

1 1 Solid Waste Handling:

2 Many older generation plants use cement or urea-form-3 aldehyde resin waste solidification with a minimum of 4 remote handling of the waste. ACNGS will use a bitumen 5 based waste solidification system, which will generate 6 smaller volumes of waste than the systems at older plants.

7 The ACNGS system will also have a higher degree of auto-8 mation compared to older systems. Both features will reduce 9 operator contact time with the waste and will thus decrease 10 exposure.

11 Control Rod Drive Removal:

12 The control rod drive mechanisms which are mounted on 13 the bottom of the reactor vessel must be maintained 14 periodically. Radiation exposure is accumulated for the 15 most part in removing the CRD mechanisms from the reactor l

16 vessel and in replacing them, and to a lesser extent in 17 rebuilding the CRD mechanisms. The Allens Creek plant will 18 have an under-vessel CRD servicing platform, which is an 19 improvement over the design for earlier BWR's. It is 20 anticipated that this will reduce the number of man hours 21 spent in the area under the vessel.

2; Control Rod Drive Maintenance:

23 The CRD maintenance area in the Fuel Handling Building incorporates a number of features to reduce exposure: a 24 1

1 shielded filter for the flush tank circulating water; a 2 shielded drum for disposal of contaminated components; a 3 shielded flush tank long enough to permit disassembly of the 4 CRD mechanism underwater; separate storage for the main-5 tained from the non-maintained underwater stored CRD; and 6 sufficient space to facilitate the work required to service 7 the CRD'c.

8 Recirculation and Feedwater System Maintenance:

9 Current generation BWR's have experienced significant 10 exposures due to repair of recirculation system piping.

11 This is a result of stress corrosion cracking of weld sensi-12 tized 304 stainless steel piping and corrosion-thermal cycle 13 fatigue cracking of feedwater spargers. ACNGS will use 316K 14 stainless steel in the recirculation piping and an improved 15 nozzle design to prevent cracking and thereby reduce maint-16 enance exposures.

17 Recirculation Pump Maintenance:

18 Radiation levels in the vicinity of the reactor recircu-19 lation pumps / motors tend to be elevated due primarily tc 20 activated corrosion products deposited on the inside wall of l

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21 the system piping. One of the more time consuming jobs is 22 replacement of the pump shaft seals. In order to reduce the 23 frequency of seal replacement caused by wear, and thus 24 reducing exposure, ACNGS uses filtered water from the

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1 control rod drive system to flush the seals.

2 Inservice Inspection - Reactor Coolant System:

3 Prior to inspection of piping welds, thermal insulation 4 must be removed. In many cases removing the insulation, 5 locating the weld and replacing the insulation results .in 6 more radiation exposure than the inspection of the weld. A 7 significant fraction of the man-rem associated with an

8. outage is incurred while performing in-service inspection on 9 the vessel and reactor coolant piping, especially due to the 10 number of locations which require physical (including ultra-11 sonic) testing and inspection.

12 The design of ACNGS includes many features to speed in-13 service inspection. Platforms are provided for easy access.

14 Insulation is attached to a permanent structure t to the 15 reactor vessel) so removal is not required, and an automatic 16 ultrasonic test probe will be used. These featurca cctve to 17 expedite inspections and reduce exposure.

18 Safety Relief Valva Maintenance:

1 19 Several safety relief valves are removed, tested and 20 replaced during each refueling. Removal and installation 21 within the drywell are responsible for a large portion of 22 the exposure received from this cask. ACNGS has a permanent 23 hoist system in the drywell to aid in removal of the relief valves. Less time will be spent in *he drywell thus re-24

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1 ducing exposure.

2 Saubber Inspection and Maintenance:

3 All safety-related hydraulic snubbers must be 100 4 percent visually inspected during a refueling outage. A 5 portion of them must also be removed from the drywell area 6 to a more remote inspection facility for testing. A rotat-7 ing inspection sequence is provided to ensure that all 8 snubbers are periodically removed and tested. The hydraulic 9 snubbers tlut are used will be equipped for in-place testing 10 so they will not require time consuming removal and re-li installation, thus reducing exposur. to personnel performing 12 this job function.

13 Fuel Handling:

14 The ACNGS fuel handling platform incorporates the 15 latest state-of-the-art electrical and mechanical improve-16 ments. This will speed handling and thus reduce operator

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17 exposure during refueling.

l j 18 Reactor Water Cleanup (RWCU) Pump Maintenance:

19 The RWCU pumps require periodic maintenance to replace 20 shaft seals because of wear on the seals. Improvesents in 21 the RWCU pump and design and in the s9al cooling system will i

22 be incorporated in the ACNGS to reduce seal wear.

23 Regenerative Heat Exchanger Maintenance:

24 Leakage frcm the heat exchangers has been a problem at

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1 Operating plants have corrected the problem times in BWR's.

2 by replacing the gasket with a welded seal. Gasket main-3 tenance and the conversion to a welded seal during commer-4 cial operation have resulted in high exposure. These 5 problems will not be encountered at ACNGS because the design 6 calls for a welded seal to be used.

7 Traversing Incore Probe Repair Work:

8 Repairs to the cabling and drives to the traversing 9 incore probe (TIP) have resulted in some exposure at operat-10 ing plants. As the detector and cabling become irradiated, 11 the replacement of detectors entails exposure of maintenance 12 personnel. The ACNGS TIP system incorporates improvement in 13 mechanical design, one of which has greater remote capabil-14 ities. Remote cable cutting and disposal techniques also 15 serve to reduce operator exposure.

16 Reactor Water Sample Station:

17 Most existing BWR's have the reactor water sample I

i 18 station located inside containment. Sampling requires about 19 1250 man-hrs /yr of time inside containment. This sample

20 station will be located in a lower dose rate arza in the 21 Reactor Auxiliary Building at ACNGS which will reduce the 22 i man-rem expended per year associated with the relocated 23 sample station.

24 General Plant:

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1 Other additional ALARA concepts ar? being incorporated 2 on a plant N ecific basis. They include, for example: (a) 3 installation of permanent scaffolding for easier access; (b) 4 placing instrumentation in low radiation areas; and (c) 5 reviewing drawings and plant models specifically to ensure 6 that piping, valves, pumps and other components are arranged 7 to facilitate access and to minimize interference ta personnel 8 performino maintenance and other necessary functions. This 9 will reduce the time spent in doing maintenance, thus re-10 ducing exposure to personnel.

11 These few examples demonstrate that the ALARA philosophy 12 and the reduction of radiation exposure has been, and will 13 continue to be, an emphasir.ed consideration in the total 14 design process for ACNGS.

15 Q. Would you please summarize your testimony?

16 A. An estimated occupational dose for Allens Creek of 17 500 man-rem / year is conservative, because it is based on the 18 historical records of occupational exposures. I believe 19 that it is reasonable to expect that ACNGS will fall shcrt 20 of the 500 man-rem / year estimate because we have improved 21 the design in a number of exposure problem areas obser ved in 22 ott-ratir.g plants. The ongoing ALARA efforts, the exposure 23 reduction lessons learned from industry operating experience, 24 and the NSSS vendor and AE experience that is being in-l l _

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1 corporated into the ACNGS design demonstrates HL&P's commit-2 ment to the ALARA concept. The reduction of occupational 3 doses for Allens Creek has and will continue to receive more 4 attention and emphasis than the d7 sign of power plants now 5 in operation. Thus, estimates based on historical data 6 overestimate the probable doses at ACNGS.

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1 Exhibit RLB-1 2 EDUCATION AND PROFESSIONAL QUALIFICATIONS 3 Robert L. Baron 4 Born: Gonzales, Texas, June 15, 1946 5 Education: Gon Nles High School, Gonzales, Teas - 1964 6 Hendereon County Jr. College, Athens, Texas, 7 AS (Physics) - 1966 8 Texas A&M University, College Station, Texas, 9 BS Physics - 1968 10 Texas A&M University, College Station, Texas, 11 MSNE - 1971 12 Memberships: South Texas Chapter Health Physics Society 13 South Texas Chapter American Nuclear Society 1483 Experience 15 ]981-Present: Supervising Health Physicist, Design Services 16 Section, Health Physics Division, responsible 17 for managing the ALARA design review effort on 18 both ACNGS Unit 1, and STPEGS Units 1 and 2.

19 1979-1980: HOUSTON LIGHTING & POWER COMPANY - Lead Health 20 Physicist responsible for assuring that both 21 the ACNGS and STPEGS designs are reviewed to 22 assure compliance with RG 8.8. Supervised the 23 work of two teams of professional health 24 physicists / engineers (a total of six personnel) 25 performing the ALARA design review function

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26 for ACNGS and STPEGS including SDD's, flow 27 disgrams, P&ID's, composite drawings, equipment 28 specifications and construction activities, as

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9 1 2 appropriate., and shielding. Provided technical 3 assistance in the development of the ALARA 4 Design Review Training course (for engineers),

5 videc tapes and Student' Handbook (all now 6 copyrighted).

7 1977-1973; HOUSTON LIGHTING & POWER COMPANY - Senior Fealth g Physicist on the design of both the Allens Creek g Nuclear Generating Station and South Texas 10 Electric Generating Station. On the ACNGS 11 resconsible for supervising overall review of 12 the entire plant design to meet the intent of 13 Regulatory Guide 8.8 (ALARA). Duties included

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14 supervising three professional health physicists /

15 engineers. As health physica team leader on 16 STPEGS, had direct responsibility for the 17 Radiation Monitoring System, Breathing Air 18 System and Decontamination Facility.

19 1975-1976: HOUSTON LIGHTING & POWER COMPANY - Associate 20 Health Physicist responsible for all Health 21 Physics " Personnel Qualifications and Training" 22 activities to include establishing HL&P's 23 Radiation Worker Medical program and supervising 24 the development of a comprehensive training 25 Program for all HL&P health physics technical

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26 personnel. In addition, developed the entire

! 27 nuclear security program for .5TPEGS to meet 28 Regulatory Guide 73.55. This included l

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1 2 development and writing of the formal Security 3 Plan submitted with the STPEGS FSAR and super-4 vising the seleetion and procurement of the 5 equipment for the computerbased access control 6 and surveillance system for both the construction

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7 and operational security requirements. Super-8 vised five professionals, three technicians, 9 and the writing of the ALARA Design Review 10 Manual, now copyrighted.

11 1972-1975: RADX CORPORATION (1390 W. Belt N., Houston, Texas) 12 Corporate Healdt Physicist - Primary responsi-13 bility was the management of both the company's 14 radiologica.'. and industrial safety programs 15 and serving as the primary interface with 16 involved state and federal regulutory agencies.

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17 Additional duties included research and develop-18 ment for new nuclear-related products, super-19 vision of engineering and maintenance on all 20 equipment using, measuring or storing radio-21 active materials. Also established and managed 22 company's formal QA program.

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