ML20009E321

Transcript of SA Hucilk & Jj Boseman Testimonies on Doherty Contention 17,RL Huang Testimony on Tx Public Research Interest Group Addl Contention 41 & Jn Bailey Testimony on Doherty Contention 42.Prof Qualifications Encl

ML20009E321
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	07/20/1981
From:	Boseman J, Ryan Huang, Hucik S GENERAL ELECTRIC CO.
To:
References
NUDOCS 8107280045
Download: ML20009E321 (22)

Text

.. . '4 SECURIDES TJ,0 DC ; A ' C0'.* :LG il July 20, 1981 RECEIVED

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l01 ICE OF APPLIClii&!3 NUCLEAR REGULATORY COMMISSION

.i 't l ?! Jgg7 t-2 REPORT 3 SERVICE 3 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD p' ,g 3 \- s N ,,('

4 In the Matter of S ,,.

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HOUSTON LIGHTING & POWER COMPANY S Docket Ng\ 5'01 B' 5

(Allens Creek Nuclear Generating / ) D 6

Station, Unit 1) S ['d I,.U M L i 7 JUL 2 71901

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DIRECT TESTIMONY ON BEHALF 0 . u.s.gl3Y# ;

HOUSTON LIGHTING & POWER COMP t 4/ -

7 f1) STEVEN A. HUCIK AND JOHN J. BOSE DOHERM COMENMON 17 - SRV REMABE 10 (2) ROBERT L. HUANG ON TEXPIRG ADDITIONAL 11 CONTENTION 41 - REACTOR PRESSURE LIMIT /

SAFETY RELIEF VALVES 12 (3) JACK N. BAILEY ON DOHERTY CONTENTION 13 42 - POSITION INDICATION FOR SRV'S 14 I

15 Q. Panel, would each of you state your name, your 16 Position with your employer, and describe your professional 17 experience and educational background?

A. My name is Steven A. Hucik and I am employed by 18 yg the General Electric Company (GE) as Manager, Mark III Containment Engineering. My professional and educational 20 background is described in Attachment SAH-1.

My name is John J. Boseman and I am employed by GE as a Senior Engineer in Valves and Auxiliary Equip-ment Design. My professional and educational background i 24 l

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. i 1 described in Attachment JJB-1.

2 My name is Robert L. Huang and I am employed by GE as a Technical Leader of the BWR/6 Transient Design 3

4 Gr up. My educational and professional background is 5

described in Attachment RLH-1."

My name is Jack N. Bailey and I am employed by 6

Houston Lighting & Power Company (HL&P) as Supervising Project 7

Engineer of Mechanical, Nuclear and Health Physics. My educational and professional background is described in JNB-1.

Q. Mr. Hucik and Mr. Boseman, would you please state 10 the purpose of your testimony?

11 A. The purpose of our testimony is to address 12 Doherty Contention 17 which states that blowdown following 13 a Power Excursion Accident (PEA) , Loss of Coolant Accident 14 (LOCA) or Power Coolant Mismatch Accident (PCMA) combined 15 with a single or several relief valves stuck in either the 16 fully open or fully closed position may cause loads which 17 would crack the containment wall. In addition, the inter-18 venors question the reliability of the safety relief 19 valves used in a BWR and request that applicant be required i 20 to use the most reliable valve available.

21 Q. Mr. Boseman, please describe the safety relief 22 valve design being used in Allens Creek.

23 A. The SRV's for ACNGS are spring loaded, direct 24 l

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1 acting, dual function type valves. The current design 2 used at ACNGS is based upon the experience gained in over 3 100 reactor years of BWR operations. The design of ACNGS 4 SRV's, manufactured by Crosby Valve and Gauge Company, has 5 eliminated the causes of previously experienced undesirable 6

Performances ase.ociated principally with reverse seated type multiple stage pilot operated safety relief valves.

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The ACNGS SRV's do not use pilot or air operator diaphragms.

Instead, the ACNGS SRV's consist of a direct-acting safety valve with an electro-pneumatic actuator asseu.bly to provide for two separate and independent modes of operation 11 (safety and relief). This design improvement combined with 12 existing manufacturing control of critical dimensions and 13 clearances between all moving parts, stringent production 14 testing and inservice maintenance and inspection will make 15 it extremely unlikely that any of the 19 SRV's on ACNGS will 16 stick open or closed.

17 Q. What has been the operating experience of Crosby SRV's?

19 A. Crosby SRV's have been used in both nuclear and 20 non-nuclear applications for many years with an excellent 21 service and performance record. The present SRV design 22 with the previously stated improvements have undergone

! 23 extensive qualification testing and are expected to have an

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i 24 even better service and performance record. To date the l

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1 operational history at Browns Ferry III and at Chinshan 2 nuclear plants have shown this to be true.

3 Q. Mr. Boseman, please describe the operation and 4 function of the safety relief valves.

5 A. Safety / relief valves (SRVs) protect against 6 overpressurization of the Reactor Coolant Pressure Boundary 7 (RCPB) by opening automatically in either the relief or 8 safety modes of operation when the pressure setpoints are exceeded. Allens Creek has 19 safety / relief valves. The 9

pressure setpoints for the automatic relief function of 10 the valves are in the range of 1103 psig to 1123 psig.

Of the 19 valves, eight specially selected SRVs, which are part of the ADS system, open automatically as part of the Emergency Core Cooling System for small breaks in the RCPB where depressurization of the reactor vessel is necessary to 15 permit operation of the low pressure coolant systems.

16 The present design of the system is such that the 17 19 SRV's open at different pressure levels via the relief 18 function. At 1103 psig, 1 valve opens; at 1113 psig, 9 19 more valves open; at 1123 psig, the remaining 9 valves 20 open. In the relief mode the valve is opened by pres-21 surizing an air cylinder which moves an actuating lever 22 thereby lifting the valve stem, as shown on PSAR Figure 23 5.2.6. The air cylinder is pressurized when either of 2 24 J

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1 solenoid valves is energized. The solenoids are auto-2 matically energized by an Instrumentation and Control 3 signal generated by high reactor pressure or they can be l 4 manually energized by the operator.

5 If the reactor pressure exceeds 1123 psig and one 6 or more of the safety / relief valves are not open, the 7 valves will open automatically in the safety mode of opera-8 tion when the pressure underneath the valve overcomes the spring force holding the disc closed. The safety setpoints 9

range fr m 1165 psig to 1190 psig. All SRVs will be open 10 by the time reactor pressure reaches 1190 psig.

Q. Mr. Hucik, please describe how the discharge through the SRVs cause loads on the containment?

A. The SRV discharge piping routes reactor steam from the relief valves to the suppression pool. The discharge 15 piping is arranged so that the quenchers which are attached 16 to the end of the SRV discharge piping are uniformly distri-17 buted in the suppression pool. When a safety relief valve l

18 is opened, there is a rapid pressure build-up in the discharge 19 pipe. This rapid compression of the column of air in the 20 pipe caused by the release of reactor steam causes a sub-21 sequent acceleration of the water column in the submerged 22 portion of the pipe. During this water clearing process, 23 the pressure in the pipe builds to a peak as the last of the 24 water is expelled. At this point, the highly compressed J

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1 cushion of air between the water ,slug and the reactor steam 2 begins to leave the pipe. As the highly compressed air 3 exits, it forms an air bubble which expands and contracts, or oscillates, as it rises to the surface of the suppression 4

pool. This oscillation of the air bubble causes a pressure 5

disturbance throughout the suppression pool which is transmittec.

6 as a dynamic load to the containment. This air clearing 7

process takes about 0.75 seconds. After the air clearing 8 .

process, the quencher is acting only as a condenser as the steam from the RPV is discharged into the suppression pool and condensed. This will begin to heat the suppres-11 sion pool. If the relief valve sticks open, it will 12 continue to heat the pool but will not impart any signifi-13 cant dynamic loadings on the containment. This heat-up 14 of the suppression pool will be controlled by the Residual 15 Heat Removal (RHR) System which will keep the pool tempera-16 ture within acceptable limits.

17 Mr. Hucik, would you also de ribe the load Q.

18 combinations, pertinent to this contention, for which the 19 containment has been designed?

20 A. There are two load combinations pertinent to this 21 contention. They are as follows:

22 1. LOCA plus single SRV actuation.

23 2. Automatic Depressurization System (ADS) 24 1

1 actuation.

2 For the case of a relief valve stuck open, the worst case for containment design is a combination of LOCA 3

blowdown loads with the loads due to a single SRV actuation.

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The dynamic loadings due to this load case would be the same as for LOCA plus a stuck-open relief valve.

6 For the case of relief valves failing in the 7

Closed position, load Case #2 demonstrates that only 8 of the 19 relief valves are necessary to rapidly depressurize the reactor. Thus the failure of up to eleven relief valves 10 in the closed position will not cause the violation of a 11 safety limit.

12 i The pressure loadings on the containment due to 13 load cases 1 and 2 have been provided to HL&P by GE and 14 l

are part of the design basis for the containment as 15 described in Chapters 3 and 6 of the ACNGS PSAR.

16 Mr. Huang, what is the purpose of your testimony?

Q.

A. My testimony addresses TexPirg Adcitional Conten-1 1

18 tion 41, which alleges that there is inadequ. ate protection 19 against overpressurization of the ACNGS Reac:or Coolant i

20 Pressure Boundary (RCPB) resulting from pressure increase l

21 transients. This concern arises, according to TexPirg, 22 because the Nuclear Pressure Relief System (NPRS) is not 23 designed adequately to ensure that during the most severe l 24 1

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1 abnormal operational pressure increase transient, pressure 2 is maintained below the limit allowed by the American 3 Society of Mechanical Engineers (, A SME) Boiler & Pressure 4 Vessel Code. TexPirg's concern stems from the fact that 5 the analysis of the most severe operational transient 6 resulting in the highest nuclear system pressure rise 7 assumes that the reactor is shutdown by the high-neutron 8 flux SCRAM. TexPirg has asserted that reliance on the high 9 flux signal as a major contributor to the termination of the 10 pressure transient, and hence as a critical design input into the NPRS, does not provide aa adequate assurance 11 12 against overpressurization of the RCPB because there is a 13 history of poor performance in the BWR flux instrumentation 14 systems with inaccuracies of 5.4 percent.

Q. Would you briefly describe the purpose and design of the NPRS?

16 A. As described by Mr. Boseman, the NPRS consists of 19 safety / relief valves located on the main steam lines I between the reactor vessel and the first isolation valve 19 within the drywell. These valves protect against over-20 pressurization of the RCPB by opening automatically upon 21 geceipt of pressure signals (relief apcration) to limit a 22 pressure rise or by self-actuation (safety operation), if 23 not already automatically opened for relief operation. The 24 J

1 events that lead to actuation of the safety / relief valves 2 result from sudden reduction of steam flow while the reactor is operating at power. Major pressurization transients are 3

4 caused by the closure of the MSIVs or the turbine control valves or turbine stop valves. The closure of these valves 5

uts off the steam flow path and isolates the reactor vessel 6

from the condenser while steam is still being formed. The 7

pressure inside the vessel thus increases rapidly.

The ASME Boiler & Pressure Vessel Code requirements 9

of Article NB-7000 on setpoints of safety / relief valves are 10 conservatively satisfied by (1) setting all setpoints at or 11 below the reactor vessel design pressure and (2) by setting 12 the setpoints so that the peak vessel pressure does not 13 exceed 110 percent of the design pressure during the limiting 14 The ACNGS safety / relief valves are set pressurization event.

15 to operate in the relief mode from 1103 to 1123 psig and in i 16 the safety function from 1165 to 1190 psige This satisfies l

17 the ASME Code requirements because all valves open at less 18 than nuclear system design pressure (1,250 psig).

19 Q. What is the limiting pressurization event and how 20 does it relate to SRV capacity?

21 A. The pressure transient resulting from the closures 22 of all main steam line isolation valves (MSIV) represents 23 the most severe pressurization transient when credit is taken only for an indirectly derived SCRAM. The analysis 24

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1 of this transient conservatively assumes the failure of the 2 direct, safety-grade main steam isolation valve position 3 SCRAM. In this event the reactor is shutdown by the backup, 4 indirect neutron flux SCRAM. Consequently, the analysis 5

introduces a significant delay between the initiation of the The 6 transient (MSIV closure) and the initiation of a SCRAM.

7 Probability of this event with failure of the safety grade scram on MSIV closure is very low.

8 The required safety / relief valves capacity is-9 d termined by analyzing the pressure rise from such a con-10 servatively postulated transient. The plant is assumed to be operating at 105 percent of nuclear boiler rated steam flow conditions at a maximum operating reactor vessel dome pres-13 sure of 1,045 psig. It is further assumed that only one-half 14 of the safety / relief valves operate in the pressure relief 15 mode (setpoints are conservatively assumed to be in the 16 range of 1,115 to 1,155 psig), and the other half is assumed 17 to operate in the backup safety mode (spring setpoints are 18 conservatively ast,umed to be in the range of 1,175 to 1,215 19 psig). The analysis indicates that the design valve capacity 20 is capable of maintaining the reactor vessel pressure well 21 below the ASME Code allowable pressure in the nuclear system 22 (110 percent of design pressure or 1,375 psig). The peak 23 pressure at the bottom of the reactor vessel is 1,294 psig.

24 Therefore, the most severe over-pressure transient is termi-

1 nated well below the pressure limit required by the ASME Lade.

2 Q. What are the conservative assumptions built inte 3 this analysis?

4 A. Under the general requirements for protection 5 against overpressure, as given in Article NB-7,000 of Section III f the ASME Boiler & Pressure Vessel Code, analysis of 6

7 overpressure transients can consider the effects of an g

appropriate SCRAM from the Reactor Protection System (RPS) .

Thus, the above-described overpressure transient analysis 9

could have considered the effects of a SCRAM resulting from an RPS signal initiated by MSIV closure since this automatic SCRAM qualifies as an acceptable protection device under the 12 provisions of the ASME Code. There are four main steam lines 13 with two isolation valves per line. Position switches 14 mounted on the eight main steam line isolation valves signal 15 Each of the MSIV closure to the reactor protection system.

16 switches is arranged to provide a signal to the reactor 17 protection system before the valves are more than 10% closed.

18 This provides early positive indication of closure. The 19 logic for generating a SCRAM signal from isolation valve 20 closure is as follows: Closure of two main steam line 21 isolation valves on the same steam line will not initiate 22 SCRAM. (To allow for MSIV testing.) However, clostre of 23 one main steam line isolation valve (MSIV) in two or more 24 steam lines will cause a SCRAM. In other words, one of four l

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n 1 steam lines may be closed without initiating SCRi> but 2 closure of more than one steam line will initiate SCRAM 3 assuming any single failure.

However, even though the Code allows for considera-4 g tion of the immediate SCRAM generated by MSIV closure, the General Electric Company conservatively assumes the failure 6

of the direct, safety-grade MSIV position SCRAM signals. The 7

General Electric analysis relies only upon tha delayed SCRAM g

signal generated by high-neutron flux.

When the MSIV's close and the vessel internal 10 pressure rises, the steam bubbles in the core region collapse.

11 With more water in the core, the neutron flux increases 12 above the high-neutron flux SCRAM setpoint. This sequence 13 takes considerably longer than the direct MSIV position SCRAM, 14 and its use in the overpressure transient analysis is, 15 therefore, very conservative.

16 Thus, TexPirg's allegation that the General 17 Electric analysis is nonconservative neglects completely 18 the conservatism of taking no credit for the earlier scram.

19 Q. TexPirg also asserts that the safety / relief valve 20 capacity analysis does not account for an alleged history 21 of unreliability in high-flux SCRAM signal circuitry. Is 22 this assertion valid?

23 A. No, there are 4 divisions of Average Power Range 24 Monitors (APRMs) which measure neutron flux and thus reactor 1 power levels. The APRM signals are monitored by the 2 Reactor Protection System (RPS). If the APRM measurement 3 exceeds the setpoint, a SCRAM will be initiated. The 4 4 divisions of APRMs are amply redundant and are routinely 5 calibrated and checked through heat balances at power 6 operations.

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In addition, the high APRM scram setpoint assumed in the overpressure protection analysis-is at a conserva-6 9

tively high level above the nominal setpoint. This accounts f r initial setpoint errors and setpoint drift that may occur 10 during operation. Typically, the assumed setpoint in the analysis is about 4% above the nominal setpoint. Furthermore, the neutron flux increase in the overpressure transient caused by MSIV closure with assumed failure in the direct j

scram is very rapid. So, the alleged APRM uncertainty of

+5.4% could only delay the scram initiation time by less than 16 50 milliseconds, and its effect on the peak vessel pressure 17 is still well below the ASME Code limit.

18 Finally, further indepth protection offered by the 19 safety / relief system is demonstrated by the fact that 20 pressure limits will not be exceeded even if the high-21 neutron flux scram (the second or delayed SCRAM) is assumed 9

to fail. The SCRAM under these conditions is initiated by 23 the high reactor pressure trip signal. The probability of 24 the simultaneous failure of the MSIV position SCRAM and high-

1 neutron flux SCRAM signals is obviously extremely low. But 2 even assuming these incredibly unlikely events, the peak 3 reactor vessel pressure for this transient is still below the 4 ASME Code limit of 1,375 psig. Hence, even if TexPirg's 5 unfounded claims of high-flux signal unreliability were true, 6 the RCPB would still be adequately protected against over-7 pressurization.

8 Q. Mr. Bailey, what is.the purpose of your testimony?

9 A. The purpose of my testimony is to address Doherty Contention 42 which alleges that the design of the SKV 10 position indicators is inadequate. He bases his allega-11 12 tion on events at TMI and the main thrust of his contention is that HL&P has not explained how it will comply with the 13 ocommendations that evolved from the TMI incident.,

4 Q. Are you familiar with HL&P's commitments to meet 5

the new regulatory requirements that evolved from the TMI incident?

17 A. Yes, I am. I was the engineer at HL&P who had primary responsibility for the preparation of HL&P's detailed response to NUREG-0718. Accordingly, I am intimately 20 familiar with the details of the steps taken to comply with 21 the new regulatory requirements that evolved from TMI. ,

22 Q. Turning to Mr. Doherty's contention, does the ACNGS 23 design provide direct indication of SRV position in the main 24 control room?

1 A. Yes. As indicated in Appendix 0, page 0-9" of 2 the PSAR, Allens Creek will provide direct indication 3 of SRV position in the main control room. This commitment 4

was made in response to Item II.D.3 of NUREG-0718 " Licensing 5

Requirements for Pending Applications for Construction Permits and Manufacturing License", which requires a demon-6 stration that design and implementation can be completed 7

g prior to the issuance of an operating license.

0 Please describe the SRV position indication design 9

to be incorporated in the Allens Creek design.

A. SRV position indication will be determined by pressure measurement in the SRV discharge pipe. Data from plants presently in operation demonstrates that this method 13 provides an adequate indication of SRV position. The actual 14

, essure setpoint to be used at ACNGS will be determined from 15 a combination of analysis and field test data, and will be 16 submitted with the FSAR. Indication in the Main Control 17 Room will be on two light matrices, one for each division of 18 position measurement, on the Reactor Core Cooling Systems

' benchboard above the manual control switches for the relief 20 valves. The indication will be redundant, safety grade, 21 seismically and environmentally qualified, and powered from 22 a Class IE power sovice. An alarm indicating that an SRV 23 is open will be provided.

24 There are no questions regarding technical feasi-

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• 1 bility or state-of-the-art of the SRV position indication .

2 design, nor is there any concern that it cannot be implemented 3 prior to OL issuanc - In fact, this design concept has been 4 approved by the NRC for use on the Hatch nuclear plant 5

perated by Georgia Power Company. Accordingly, there is no d ubt that we can demonstrate that our design concept is 6

techrically feasible.

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Attachment JNB-1 Jack N. Bailey ,

I graduated from Georgia Tech in 1972 with a degree in Electrical Engineering. From June 1972 through November 1977, I served in the U.S. Navy. During this time I attended the Naval Nuclear Power School, served on the USS Long Beach, a nuclear-powered guided missile cruiser, and served as a supervisor in the Naval Nuclear Power Training Unit, DlG Prototype.

I have been employed by HL&P since December 1979.

Initially I was a member of the ACNGS Engineering Team with responsibility for piping program organization, painting requirements, inservice inspection access require-ments and other balance of plant systems. From February 1980 to August 1980 1 was Chairman of HL&P's TMI Design Task Force. The task force studied the TMI accident and made recommendations for needed studies and design changes.

From August 1980 to March 1981 I was Engineering Team Leader for Allens Creek with responsibility for all NSSS systems and a variety of balance of plant systems. I was also responsible for HL&P's response to NUREG-0718, " Licensing Requirements for Pending Applications for Construction Permits and Manufacturing License." I was promoted to my current position as Supervising Project Engineer in March 1981, and in this position I am responsible for supervising the mechanical, nuclear and health physics design review of ACNGS.

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Attachmant JJB-1 RESUME John J. Boseman 4

Position: Senior Engineer Emp! oyer: General Electric Co., San Jose, CA.

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! Principal Duties and Responsibilities:

Responsible for the design, development, qualification iests, programs, and related technological growth of equipment assigned including all necessary technical and liaison assistance to support installation, .

testing, inspection and maintenance of the equipment. Formulated, directed, and participated in the design, tests, evaluations and programs to improve the reliability and optimization of Sc' *.y/ Relief Valve designs for BWR applications.

Backaround 5/68 - 1/77 -

Product Engineer with General Electric Co., MAO, Schenectady, N.Y. - As cognizant engineer liaisoned and performed design, development, tests, manufacture, evalua-tions, installation and field support functions for various, types of Naval nuclear power plant fluid components. As Product Engineer performed and provided technical direction for the design, applications, qualifications, development, installation and maintenance of assigned equipment.

(e.g. - Valves (checks, gates, globes, relief, motor operated,

hydraulic operated, etc.; magnetic separators, filters, pres-l sure vessels, pipe and fittings, demineralizers, reactor view-

! ing devices, etc.)).

1 Assigned to participate in the Navy's 1970 Valve Design Review

( Task Force.

6/66 - 4/68 - Associate Engineer with Lockheed Missile & Space Co.,

Sunnyvale, CA. - Analyzed, proposed, designed, developed, j tested and liaisoned the manufacture of advanced electro-mechanical microwave antenna and antenna systems for Satellite and Polaris / Poseidon Missile applications.

2/66 - 5/66 - Assistant Engineer with The Boeing Company, New Orleans,

LA. - Performed environmental and simulated testing of fluid power components and sub-systems applicalbe to the Saturn IV & V Booster System. Established cause for LOX and lift check valve failures and recommended corrective action to preclude recurrence.

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7/64 - 1/66 -

Assistant Engineer with Delta Steamship Lines, Inc.,

Nen* Orleans, LA. - Operated, tested, maintained and repaired steam and diesel power plant systems and equipment including the review and liaison of design proposals and major shipyard overhauls and repairs. Conceived and demonstrated an emergency technique to repack stern gland while underway.

Education /Trainina 1964 - B.S. Marine. Engineering - U.'S. Merchant Marine Academy.

Nuclear Radiation & Environmental Effects / Navy Structural Design Basis-63/ Reactor Plant Technology /ASME' Pressure Vessel Codes /

Kepner-Tregoe " Problem Solving & Decision Making Course.

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Publications / Articles ASME 80-C2/PVP-29'- OPERABILITY ASSURANCE TESTING OF ASME CODE, CLASS 1, SAFETY / RELIEF VALVES.

EVALUATION OF 3 THERMALLY SHOCK TESTED 1/2-INCH GLOBE VALVES (MEDF iS4) - U.S. Navy Document (Restricted).

U.S. NAVY NUCLEAR VALVE DESIGN MANUAL (VDM-71) (Classified).

Plant Equipment Design Memorandum No. 126 SEALING MECHANISM FACTORS.

Attachment RLH-1 Robert L. Huang I received a Bachelor of Science Degree in Nuclear Engineering from National Tsing Hua University in Taiwan, China in 1968.. During 1969 to 1975 I attended Columbia University in New York and received a Master of Science Degree in Nuclear Engineering in 1970 and a Doctor of Engineering Science Degree in Nuclear Engineeringt in 1975.

I joined the General Electric Company's Nuclear Energy Division in September, 1974. I was responsible for BWR/6 transient safety design, which includes reactor vessel overpressure protection and reactor fuel overpower protection.

Shortly thereafter, I performed transient safety analyses

' for the General Electric Standard Safety Analysis Report (GESSAR), and for the Grand Gulf FSAR. Since then, I have performed design and analysis studies for all BWR product lines, and led the transient design efforts which esta'alished I the BWR/6 product line.

In my current position as a Technical Leader, I provide technical guidance and work direction to engineers to perform the reactor system transient safety design analyses and licensing evaluations on BWR plants. I am also responsible i

for the establishment and specificatici. of reactor system l hardware and reactor protection system functional requirement.

Attachment SAH-1 Steven A. Hucik Mr. Hucik is manager of the Mark III Containment Engineering Unit of the Nuclear Power' Systems Engineering Department in the General Electric Company. His employment with General Electric began in 1973 and his experience has been mainly in the containment loads area. His unit is responsible for all Mark III containment analysis and dynamic load definition for loss-of-coolant accident (LOCA) and safety / relief valve discharges.

Mr. Hucik's experience in Mark III containment includes responsibility for the dynamic loads development and application, documented in the Mark III Containment Loads Report. He was responsible for the analysis of the Caorso SRV test data used to support the SRV load reduction defined in the final Mark III Containment Loads Report.

Mr. Hucik's Mark III experience is also supported by his previous involvement in the Mark I Containment Program.

His involvement included applicatiin of test data, load definition development, customer and Architect / Engineer interface, and US regulatory presentations.

l Mr. Hucik has provided support to the Mark III Customers and Architect / Engineers in the load application area for both LOCA and SRV discharge events. He has also made presentations of the design and licensing osses to the US regulatory agencies, customers, and Architect Engineers.

l Mr. Hucik is a 1973 graduate of Washington State l University with a B.S. Degree in Mechanical Engineering. In t

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1976, he received an M.s. Degree in Mechanical Engineering f

from the University of California, Berkeley. He is also a Registered Professional Engineer in the State of California.

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