ML20082A719
| ML20082A719 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 10/11/1983 |
| From: | Burns L, Carter S, Cooke F, Ranganath S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20082A708 | List: |
| References | |
| NUDOCS 8311180232 | |
| Download: ML20082A719 (28) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION Before the Atomic Safety and Licensing Board In the Matter of )
Philadelphia Electric Company Docket Nos. 50-352 50-353 (Limerick Generating Station, Units 1 and 2)
STATEMENT OF SAMPATH RANGANATH, LLOYD S. BURNS, JR.,
FRANKLIN E. COOKE, AND STEPHEN E. CARTER IN SUPPORT OF MOTION FOR SUPMARY DISPOSITION OF CONTENTION I-62 Q.1. State your names, addresses and briefly describe your professional.
qualificationsrelatingtothesubjectofpressurizedthermalshock.
A.la. My name is Sampath Ranganath. I am Manager of the Mechanics Analysis group of the Nuclear Engineering Division, General Electric Company. In that position I am responsible for stress analysis work on all BWRs, including Limerick Generating Station. Additional responsi-bilities include fracture mechanics, fatigue evaluations, finite element analysis, stress corrosion cracking, residual stress analysis, and dynamic margin of components. I received a BSME from Bangalore University in 1965, an MS in Mechanical Engineering from Indian Institute of Science in 1967 and a Ph. D. in Engineiring from Brown University in 1971. I have been employed by GE for nine years in the area of solid mechanics.
The statement of my professional qualifications is attached hereto and incorporated by reference herein.
A.lb. My name is Lloyd S. Burns, Jr. I am a Senior Engineer and a Technical Leader of the Containment and Radiological Engineering group of I
8311180232 8311145 1 PER ADOCK 05
the Nuclear Engineering Division, General Electric Company. In that position I am responsible for the calculation of neutron flux and fluence en General Electric BWRs including Limerick Generating Station. I obtained a BA degree in physics from Kalamazoo College. I have been employed by General Electric for 27 years working in the areas of nuclear radiation analysis. The statement of my professional qualifications is attached hereto and incorporated by reference herein.
A.lc. My name is Franklin E. Cooke. I am a Principal Design Engineer in the Recctor Pressure Vessel and Internals Design group of the Nuclear Engineering Division, General Electric Company. In that position I am responsible for defining the reactor operating limits to assure adequate fracture toughness of the reactor pressure vessel, including that for the Limerick Generating Station.
I obtained a Bachelor of Science degree in Mechanical Engineering from Southern Methodist University in 1950. I have been employed by the General Electric Company for 28 years in the area of nuclear steam supply system design, testing and operation. I have been working in the area of fracture toughness requirements since 1974. The statement of my profes-sional qualifications is attached hereto and incorporated by reference herein.
A.1d. My name is Stephen E. Carter. I am an engineer in the Plant Materials Application group of the Nuclear Engineering Division, General Electric Company. In that position I provide support for design, procure-ment and quality control in the areas of reactor pressure vessel and 2
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piping and I as responsible for assuring compliance with the requiremegts 4 of 10 CFR Part 50, Appendices G & H. I received a B.S. degree in Metallurgy from the Pennsylvania State University in 1980. I have been employed by GE fer the past 3 years in the area of nuclear materials technology. The statement of my professional qualification is attached hereto and incorporated by reference herein.
+1 Q.2. Are you familiar with contention I-62 in the Limerick proceeding which alleges that the Limerick Generating Station can suffer a major breach of containment due to Pressurized Thermal Shock?
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A.2. Yes. '
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Q.3. What sections of the Final Safety Analysis Report for the Limerick Generating Station contain information that is related to the responses to Centention I-62?
A.3. Related reactor pressure vessel information is found, primarily, '
in FSAR Section 5.3. Fracture toughness and material surveillance x
aspects, for example, are detailed in Sections 5.3.1.5 through 5.3.1.9.
- s FSAR Secticn 4.3.2.8 describes reactor pressure vessel neutron flux and fluence calculations'. i 4 I l
Q.4. Please define the term Pressurized Thermal Shock. ,
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A.4. Pressurized Thermal Shock (PTS) is a condition that may affect some PWR's but no BWR's. PTS results from the introduction of cold water intoahotpressure\esselwhilethepressureishigh. Therms.1 stresses are produced in the vessel walls when cold water is introduced into the vessel. Thesestresses,inconjunctionwithstresseswhichoccurasa result of high vessel pressure, have the potential to cause crack propaga-tion in vessel materials. The materials of which the reactor pressure vessel is made can become embrittled as a result of substantial neutron bombardment. This embrittlement could, under certain conditions, adversely affect the ability of the materials to withstand these combined stresses.
PTS has been recognized as a problem in some pressurized water reactors because(i)pressureforsomeeventscanremainhighinaPWRduringcold water injections, and (ii) the neutron flux' s high enough to cause significant vessel material embrittlement. '
e Q.5. Describe why the phenomenon of pressurized thermal shock is not significant for a boiling water reacto'r (BWR) such as the Limerick Generating Station.
A.5. PTS is not a problem for boiling water reactors since the necessary ingredients--high reactor vessel pressure during cold water injection and significant neutron irradiation embrittlement -- do not occur in a BWR. Furthermore,'the decrease in vessel material fracture toughness as a result of irradiation is substantially less in a BWR than that in a PWR. Specific reasons for these conclusions are as follows:
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- The pressure in a BWR follows the water-steam saturation curve.
Duringcoldwaterinjection,suchasfromtheHighPressureCoolant Injection (HPCI) system, the pressure in the BWR drops because the water and steam remain in equilibrium.
- The neutron fluence at the vessel wall in a BWR is very low compared with a PWR because of the presence of a large water-iilled annulus between the vessel and the shroud surrounding the. reactor core, and because of a substantially lower reactor core power density. Thus, radiation embrittlement effects are minimal in a BWR.
. The design, construction, testing, operation and surveillance, together with the physical behavior of the BWR, as stated above, assure that PTS is not a problem for the Limerick Generating Station.
y Q.6. Please describe the codes and standards to which the Limerick reactor vessels are designed and fabricated.
A.6. The Limerick reactor pressure vessels are designed, fabricated, Ltested, inspected, and stamped in accordance with the requirements of ASME Code Section III, Nuclear Power Plant Components, Class 1, including the Summer 1969 Addenda and ASME Code Section IX, Welding Specifications, including Summer 1969 Addenda.
Q.7. Describe the manner in which the Limerick pressure vessels were designed and constructed in accordance with ASME Sections III and IX.
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. 1 A.7. The reactor pressure vessels are vertical, cylindrical pressure <
l vessel of welded construction fabricated in accordance with ASME Code,Section III, Class 1 requirements. All RPV fabrication was performed in accordance with GE approved drawings, fabrication procedures, and test procedures. The shell and vessel head.were made from formed low-alloy steel plates, and the flanges and nozzles from low-alloy steel forgings.
Welding performed to join these vessel components was in accordance with ASME Sections III and IX requirements. Weld test samples were required for each procedure for major vessel full penetration welds.
All plate, forgings, and bolting were 100% ultrasonically tested and surface examined by magnetic particle methods or liquid penetrant methods in accordance with ASME Code Section III requirements. In addition, the pressure retaining welds were ultrasonically examined in accordance with ASME Code Section XI and Regulatory Guide 1.150 (Rev. 1) guidelines.
Fracture toughness properties were measured and controlled in ccordance with ASME Code Section III requirements to limits specified by General Electric Company.
Q.8. Define the meaning of fracture toughness as used in the design of,the Limerick reactor pressure vessels and describe how the fracture toughness of the ferritic pressure boundary materials of the Limerick reactor pressure vessels were determined.
A.8. " Fracture Toughness" is a measure of a material's inherent ability to resist unstable extension of flaws in the presence of applied, 6
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dynamic, impact, or other suddenly applied loads. Appendix G of 10 CFR Part 50 specifies minimum fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of water cooled power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime.
Section 5.3.1.5 of the FSAR demonstrates the compliance of the Limerick reactor vessel design with the requirements of 10 CFR Part 50 Appendix G.
As described therein, the ferritic pressure boundary materials of the Limerick 1 reactor pressure vessel (RPV) were qualified by toughness testing in accordance with the 1968 edition of the ASME Code including the Summer 1969 Addenda. In addition, these materials were tested to the augmented fracture toughness requirements specified by GE. All RPV components were impact tested by either the drop-weight test or the Charpy V-Notch impact test. Both impact tests were conducted on beltline plate material, closure flange material, top head material, feedwater and LPCI nozzle material forgings. Reference temperature nil ductility transition temperature (RTNDT) values for the RPV components were estab-lished using impact test data and procedure that meets the requirements of 10 CFR 50 Appendix G. Similarly, compliance of the Limerick Unit 2 reactor with 10 CFR Part 50 Appendix G will be demonstrated prior to issuance of an operating license for that unit.
Q.9. Describe the effect of neutron irradiation over the life of the facility on fracture toughness properties of the vessel.
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A.9. The effect of neutron fluence on pressure vessel materials is to cause a decrease in its fracture toughness. The decrease in toughness is, however, significant only at high fluences--well above those expected in the Limerick vessels based upon conservative calculational techniques as verified by operational experience.
Q.10. Describe how the fluences for the Limerick vessels were determined.
A.10. The neutron fluence calculations were carried out on the basis of analytical models incorporating reactor core data, geometric arrangement, and basic physical data. These items are incorporated into computer programs to solve the transport equations for nuclear radiation. These programs were verified in strict compliance with the quality assurance provisions of 10 CFR 50 Appendix B. The results of these calculations have been compared to field measurements and found to conservatively overpredict the neutron fluence. -
The BWR vessel geometry from the view point of neutron vessel fluence calculations can be described in terms of a cross sectional view of concentric cylinders. The inner core region consists of 764 square fuel bundles of identical geometrical size arranged in a pattern simulating a cylinder. This region is surrounded by an annulus of water of an average thickness of approximately 8 inches. The core water is separated from the downcommer water by a stainless steel metal shroud 2 inches thick.
The downcommer region containing the jet pumps is a water region 22.1 inches thick. The vessel is approximately 6.2 inches thick in the beltline region. See FSAR Figure 4.3-29.
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l Q.11. What were the effectq of the calculated fluence on reactor vessel operating limits?
A.11. A maximum fluence for the core beltline material of 1.1x1018 n/cm2 at one quarter of the vessel wall thickness from the inside diameter was conservatively calculated. This fluence was applied over the entire length of the core beltline plates and welds in order to determine the shift in fracture toughness which was, in turn, used to determine the reactor operating limits at the end of reactor service life. For critical nozzles such as the LPCI nozzle, the fluence was uniquely calculated for each nozzle. The resulting core beltline operating limitations are less restrictive than those operating limitations established as a result of other reactor vessel parts. Such vessel parts are located well away from the core beltline in a region of insignificant fluence with respect to fracture toughness properties of the vessel material.
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The results are depicted on FSAR Figure 5.3-4, a copy of which is attached hereto. This figure will become part of the Technical Specifications for Limerick Unit 1 and will define the minimum temperature versus reactor pressure required to assure adequate fracture toughness. A similar curve will be developed for Unit 2. The dashed lines represent the limits for the core beltline region.after exposure to the fluences described above.
1 It may be seen that the solid curves to the right of the respective dashed lines require a higher temperature for a given pressure. This figure, therefore, demonstrates that the core beltline region is not limiting with regard to the setting of temperature and pressure limits 1
for the reactor.
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The portion of the curve labeled " Vessel Discontinuity Limits" represents the limits for those portions of the reactor vessel away from the core beltline. They are established by stress analyses which include the effects of cold water injections and other mechanical loads that may be applied in addition to reactor pressure loading. The evaluations are in accordance with ASME Code Section III and 10 CFR 50 Appendix G.
5 Q.12. What confirmatcry fracture mechanics evaluations have been performed?
A.12. A fracture mechanics evaluation of the effects of a potential loss of coolant accident (LOCA) in a BWR/6 reactor vessel of similar design to Limerick has been performed. The purpose of that analysis was to determine if cracli propogation could occur as a result of stresses inducedinthevesselbytheinjectionofcoldwaterbytheEmergency Core Cooling System (ELCS). The analysis included determination of thermal stresses in the vessel and calculation of resulting applied stress intensity factors which would exist at the tip of a potential crack in the vessel wall. The stress intensity factors were compared to the available materials fracture toughness which was calculated considering vessel temperature and neutron fluence effects on vessel material properties.
It was found that the available vessel material fracture toughness always exceeded the applied stress intensity factors for all postulated crack depths. It was therefore concluded that crack propogation would not occur, even for large initial flaws with depths approaching the vessel wall thickness.
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h Limerick specific fluence, material properties, and NDT shift data confirm that the results of the generic analysis are applicable to the Limerick Reactor Pressure Vessels.
Q.13. What surveillance program is planned for the Limerick vessels?
A.13. Surveillance specimens were fabricsted from heats of materials (i.e., both weld and plate) that are actually used in the beltline core region. The coupon orientations are equivalent to the orientations of
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the specimens used to establish unirradiated impact properties. Three surveillance specimen capsules which contain both Charpy V-Notch and tensile specimens are placed in the vessel. Each capsule also includes an Fe, Ni, and Cu flux wire. A separate neutron dosimeter is attached alongside one of the capsules and contains 3 Cu and 3 Fe wires. These 4
dosimeters can be used to periodically check the actual flux field to which the capsules are exposed. The specimens will be removed at intervals over the life of the vessel in order to confirm the adequacy of the predicted irradiation effects.
Q.14. What are your overall conclusions concerning the effect of PTS on the Limerick Generating Station pressure vessels?
A.14. The conditions necessary for the occurrence of PTS in PWRs cannot' occur in BWRs such as Limerick because of differences in design and operation. The integrity of the Limerick vessels are assured by conservative design, selection of materials, construction, testing, operation, and surveillance in accordance with regulatory requirements.
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l MATERIAL ERIAL MADfUS MATERIAL DMW j NO. NAME WATER 0.274g/cm 3 3
93.56 UO 2 2.642 s/cm 1 REACTOR CORE ZlRCONIUM 0.896 g/cm 3
WATER 0.74 g/cm WATER 101.4 2
334L STAINLESS STEEL FROM ASME SHROUD 103.4 3 SA 24G WATER 0.74 g/cm 3 l
4 WATER 125.5 CARBON STEEL FROM ASME $33 5 VESSEL 131.8B 1.3 X 10-3g/cm 3 AIR '
s AIM UMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT MODEL FOR ONE OlMENSIONAL
, TRANSPORT ANALYSIS OF
- VESSEL FLUENCE FIGURE 4.3 29
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I I E A. SYSTEM HYDROTEST LIMIT l l l Wim NEL m VEMEL CORE BELTtiNE '
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l a LIMITS f f ASSUMED 300F TEMP. SHIFT FROM AN INITIAL PLATE RTNOT OF 200F, g l CURVES ARE NOT LIMITING (SHOWN w h* FOR INFORMATION ONLY) 8a- 1 I I ti N I I l j -
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CURVES A,B.& C ARE PREDICTED b / TO APPLY AS THE LIMITS FOR 8 7 40 YEARS (32 EFPY) OF OPERATION y
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psig APPENDIX G LIMITS BOLTUP 800F 200 --
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TEMPERATURE 0 F LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS RfPORT UNIT 1 MINIMUM TEMPERATURE I
REQUIRED VS REACTOR PRESSURE REV. 22,07/83 FIGURE 5.3 4
+
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION Before the Atomic Safety and Licensing Board 4
In the Matter of )
)
Philadelphia Electric Company ) Docket Nos. 50-352
) 50-353
)
-(Limerick Generating Station, )
Units 1 and 2)
AFFIDAVIT OF SAMPATH RANGANATH, MANAGER, MECHANICS ANALYSIS GROUP, NUCLEAR ENGINEERING DIVISION GENERAL ELECTRIC COMPANY, REGARDING CONTENTION I-62 STATE OF CALIFORNIA ) ss:
COUNTY OF SANTA CLARA)
Sampath Ranganath, being duly sworn according to law, deposes and says:
- 1. My name is Sampath Ranganath. I am Manager, Mechanics Analysis group, Nuclear Engineering Division, General Electric Company.
- 2. I participated in the answer to Question la, which includes a Statement of my Professicnal Qualifications, and the answers to Questions 4, 5,,12 and 14. The statements therein are true and correct to the best of my knowledge, information and belief.
10 II 83 %- !*'--
Date Sampath'tanganath V l
i Subscrioed and sworn before me on ll Odo B E R. 1923 OF I [ y ggf
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SANTA CLARA COUNTY 2 ai 3 Nrory eueiic a
W cow & Expwes Dec. 21,1984 ft
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION i Before the Atomic Safety and Licensing Board l
In the Matter of )
)
Philadelphia Electric Company ) Docket Nos. 50-352
) 50-353
)
(Limerick Generating Station, )
Units 1 and 2)
AFFIDAVIT OF LLOYD S. BURNS, JR. SENIOR ENGINEER, CONTAINMENT AND RADIOLOGICAL ENGINEERING, NUCLEAR ENGINEERING DIVISION, GENERAL ELECTRIC COMPANY, REGARDING CONTENTION I-62 STATE OF CALIFORNIA ) ss:
COUNTY OF SANTA CLARA)
Lloyd S. Burns, being duly sworn according to law, deposes and says:
- 1. My name is Lloyd S. Burns, Jr. I am a Senior Engineer in the Containment and Radiological Engineering group, Nuclear Engineering Division, General Electric Company.
- 2. I participated in the answer to Question Ib, which includes a Statement of my Professional Qualifications, and the answers to Questions 2, 3, 10, 11, and 14. Tne statements therein are true and correct to the best of my knowledge, information and belief.
- 3. I am familiar with the contents of FSAR Section 4.3.2.8 as they pertain to the above answers. The statements, tables, and figures in that section of the FSAR, as amended through Rev. 22, 7/83, are true and correct to the best of my knowledge, information and belief.
Geb /o . /973 Date N) A Y h' Lloyd S. Burns, Jr.
Subscribed and sworn before me on h /0 /V3
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OFFICIAL SEAL L
) RUTHE M KfNNAMON
, NoTAltf PUBLIC - CALWORNIA 1 /1 '
< SMfTA CLARA COUNTY , Notary Public
, My comm. agires AP:t *C,19!3 ? -
. -_ .. : m U5 Cwinst Auenes, San Jose, CA 95125 110102-2
UNITED STATES OF AMERICA NUCLEAR REGULATORY C0 MISSION Before the Atomic Safety and Licensing Board In the Matter of Philadelphia Electric Company Docket Nos. 50-352 50-353 (LimerickGeneratingStation, Units 1 and 2)
AFFIDAVIT OF FRANKLIN E. COOKE, PRINCIPAL DESIGN ENGINEER, REACTOR VESSEL AND INTERNALS DESIGN, NUCLEAR ENGINEERING DIVISION, GENERAL ELECTRIC COMPANY, REGARDING CONTENTION I-62 STATE OF CALIFORNIA) ss-
- COUNTY OF SANTA CLARA)
Franklin E. Cooke, being duly sworn according to the law, deposes and says:
- 1. My name is Franklin E. Cooke. I am a Principal Design Engineer in the Reactor Pressure Vessel and Internals Design group, Nuclear Engineering Division, General Electric Company.
- 2. I participated in the answer to Question 1c, which includes a Statement of my Professional Qualifications, and the answers to Questions 2, 3, 6, 7, 11 and 14. The statements therein are true and correct to the best of my knowledge, information and belief.
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- 3. I am familiar with the contents of FSAR Section 5.3 as they l
pertain to the above answers. The relevant statements, tables, and figures in that section of the FSAR, as amended through Rev. 22, 7/83, are true and correct to the best of my ~. wledge, information and belief.
$Date& 'b b l W 3 tranklin t. Cooke 0 -
i subscribed and sworn to Before Me on 7 October 1983.
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Nethry Public
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g OFFICIAL SEAL KARIN 5. VOGELHUB9t NOTARY PUBUC.CAUFORNIA I Mye SANTA CLARA COUNTY
' . Expires Dec. 21,1984 000000000 06
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Board In the Matter of )
)
Philadelphia Electric Company ) Docket Nos. 50-352
) 50-353 .
)
(Limerick Generating Station, )
Units 1 and 2) -
AFFIDAVIT OF STEPHEN E. CARTER, ENGINEER, PLANT MATERIALS APPLICATION, NUCLEAR ENGINEERING DIVISION, GENERAL ELECTRIC COMPANY, REGARDING CONTENTION I-62 STATE OF CALIFORNIA ) ss:
COUNTY OF SANT4 CLARA)
Stephen E. Carter, being duly sworn according to law, deposes and says:
- 1. My name is Stephen E. Carter. I am an engineerin the Plant Materials Application group, Nuclear Engineering Division, General Electric Company.
- 2. I particTpated in the answer to Question Id, which includes a Statement of my Professional Qualifications, and the answers to Questions 2, 3, 8, 9 and 13. The statements therein are true and correct to the best <
of my knowledge, information and belief.
- 3. I am familiar with the contents of FSAR Sections 5.3.1 and 5.3.3 as they pertain to the above answers. The relevant statements, tables, and figures in those sections of the FSAR, as amended through flev. 22, 7/83, are true and correct to the best of my knowledge, informa-tion and belief.
io/n/n na hm Stephen'E. Certer Date' '
Subscribed and sworn before me on 1I Oc_hR F AL. I W_3 omCIAL SEAL 4 A .af &
Kdtary Public
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KAREN 5. VOGEHUBat 9 NOTAAY PuBUC.CAUFORNIA SANTA CLARA COUNTY My Comminion Expires Dec.2L 1984 .
3
[ntervenor Iewi4 R7buttal in Affidavit of Ellict and Hodge3 dated 11/03/83
)nly material that ic new shall be commented upon. Please see previous filing of 10-24-83 for clarificationof prior material.
- 1. No comment.
- 2. No comment.
- 3. No ennent.
4 No comment.
5 I agree ,1-t "Barry Elliot ... lack (s) the information necessary to confirm the accuracy 'of those statements" and " a number of those statements are not material to PTS considerations." These quotes refer to the statemento of material facts by the Applicant.
- 6. No comment.
- 7. No comment.
- 3. Sea discussion of the ' direction'in this filing on Inge 2. Herein Mr Elliot refers to n " postulated crack." See Page 7 of Intervenor's filing of 10-24-83 Q12 A12(2) and footnots.
- 9. Ses discussion on page 2 of this fiiLing. Also even if' the 10% mixing sta.ted in the Elliot answer is correct, there is no 31stification to assumethat 10% is adequate sixing.
- 10. No comment.
- 11. These screenlag criteria Sere chosen by the staff and have not been theconcern of any court action. Inotherwords the staff proposes that there chosen criteria should-be ace:tpted without affording anyone a day in court to challenge that criteria.
A
- 12. p3Cuthrieformulaisdevelopedfromonlyafewhundreddatapointsashas be:n answered in discovery. Whether this limited amount of data for so many variables; contaminants , temperature, time, thermal histvry, gamma flux etc, is sufficient should really be decided in front of the ASIB. The probability limit merely refers to how
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the data falls and is really not a prediction of how different material wil1 fall cn the curve. Therefore there is a genuine issue of material fact to determine if the ataff's Guthrie curve is representative of " archival" material or material actually ;
used in the IGS. , j 13 Same argument 4s 12 above.
- 14. " expected" Either there will be or there will not be FIS transients in the IGS.
Why ";xpected?" If the staff cc.nnot make a definitive , positive statement about the possibility of PIS in BWRe such as IGS, surely that lack of definitire statement backed up with scientific fact is enough to demonstrate a genuine issue of material fact to be worthy of litigation.
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PROFESSIONALQUALIFICATIONS l
SAMPATH RANGANATH MANAGER, MECHANICS ANALYSIS NUCLEAR ENGINEERING DIVISION GENERAL ELECTRIC COMPANY My name is Sampath Ranganath. My business address is 175 Curtner Ave.,
San Jose, CA., 95125. I am the Manager of the Mechanics Analysis group of the Nuclear Engineering Division of the General Electri.c-Company. I have been in this position for five years. I am responsible for all analytical work on the Limerick Generating Station in the areas of stress analysis, fracture mechanics, fatigue evaluations, finite element methods development, stress corrosion cracking, residual stress analysis, and dynamic margin of components.
I received a BSME from Bangalore University in 1965, an MS from Indian Institute of Science in 1967, and a Ph. D. in Engineering from Brown University in 1971.
I have been employed by General Electric Company for 9 years in the area of solid mechanics. I have been a key contributor to several-EPRI programs including those related to improved fatigue design rules for carbon steel and fracture mechanics evaluation of the pressure vessel
. .. 4 under LOCA conditions. I am a member of the Subgroup on Standards and Evaltation,Section XI, ASME Code, contributing in the development of rules for evaluating flaws in nuclear pressure vessel components.
I am a member of the ASME and a registered Professional Engineer in the state of California. IamanadjunctlecturerattheUniversityofSanta Clara and have taught courses in fracture mechanics and pressure vessel design at the University of Santa Clara. I am the author of several papers in the fields of dynamic behavior, stress analysis plasticity, fracture mechanics and fatigue.
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PROFESSIONALQUALIFICATIONS LLOYD 5. BURNS, JR.
TECHNICAL LEADER, CONTAINMENT AND RADIOLOGICAL ENGINEERING NUCLEAR ENGINEERING DIVISION GENERAL ELECTRIC COMPANY My name is Lloyd S. Burns and my business address is 175 Curtner Ave. ,
San Jose, CA., 95125. I am a Senior Engineer and Technical Leader in the Containment and Radiological Engineeping group of the Nuclear Engineering Division of the General Electric Co. I have the responsibility for
. calculation of the neutron flux and fluence calculations for General Electric BWR's.
I received my B. A. in Physics from Kalamazoo College, Kalamazoo, Michigan in 1950.
I have been employed by the General Electric Company since 1956. All of my work experience with General Electric has been in the area of radiation analysis. I have worked with commercial reactor power plants since 1968.
Prior to that time I worked on mobile reactor designs primarily for government applications. I have participated in the development of the currently used neutron transport methodology.
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I am currently a Technical Leader in the Containment and Radiological Engineering group. I have overall technical responsibility for radiation analysis work within General Electric's scope of responsibility for BWR plants. One portion of the radiation analysis work is the calculation of neutron flux and fluence on BWR reactor pressure vessel. I have performed neutron flux and fluence calculations for BWR vessels and I have prepared the neutron fluence data for the Limerick Generating Station. I have actively followed the measurement and data available on vessel fluence of operating BWR reactors.
l I am a registered Professional Engineer in California.
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PROFESSIONAL OUALIFICATIONS FRANKLIN E. COOKE PRINCIPAL DESIGN ENGINEER REACTOR PRESSURE VESSEL & INTERNALS DESIGN NUCLEAR ENGIN ERING DIVISION GENERAL ELECTRIC COMPANY My name is Franklin E. Cooke. My business address is 175 Curtner Ave.,
San Jose, CA.,.-95125. I am 2. Principal Engineer in Reactor Pressure Vessel & Internals Design group of the Nuclear Engineering Division of General Electric Company. In this position I am responsible for defining the reactor operating limits to assure adequate fracture toughness of the reactor pressure vessel.
I received my Bachelor of Science degree in Mechanical Engineecing in 1950 from Southern Methodist University. ,
I have been employed by General Electric in its nuclear energy business for twenty eight years. My duties have included the design, testing and operation of the nuclear steam supply system for General Electric boiling water reactor designs. I have been working in the area of fracture toughness requirements since 1974. My current responsibilities include definition of fracture toughness operating limits for the reactor vessel,
definition of standard plant reactor specifications, review of product safety standards and regulatory guides, thermal cycle evaluations, and definition of reactor water level operating limits.
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PROFESSIONALQUALIFICATIONS STEPHEN E. CARTER ENGINEER, PLANT MATERIALS APPLICATION NUCLEAR ENGINEERING DIVISION GENERAL ELECTRIC COMPANY My name is Stephen E. Carter. Mybusinessaddressis175Curtner[ Avenue, San Jose, CA., 95125. I am an engineer in Plant Materials Application, In that position
' Nuclear Engineering Division, General Electric Company.
I provide support for design, procurement, and quality control in the areas of the reactor pressure vessel, piping and pipe whip restraints.
I am also responsible for assuring compliance with the requirements of 10 CFR Part 50, Appendices G & H. In addition, I am Program Manager for
" Process and Product Specifications".
I received a Bachelor of Science degree in JMetallurgy from the Pennsylvania State University in 1980. I have been a member of the General Electric nuclear materials staff for over ~ years. My past responsibilities at GE have included applied corrosion research of low alloy steels, austenitic stainless steels, and Inconel alloy 600.
PROFESSIONAL QUALIFICATIONS CRAIG D. SAWYER MANAGER, SYSTEMS INTEGRATION NUCLEAR ENGINEERING DIVISION GENERAL ELECTRIC COMPANY My name is Craig D. Sawyer. My business address is 175
'Curtner Avenue, San Jose, California 95125. I am Manager of Systems Integration Engineering in the Nuclear Energy Engineering Division of the General Electric Company. In this position I am responsible for all multisystem or plant technical issues affecting BWR's, including the Limerick Generating Station.
I received a B.S. in Chemical Enginee ring from the Massachusetts Institute of Technology in 1960, an M.3. in Nuclear Engineering from MIT in 1964.
From 1964 to 1966, I worked in the U. S . Army Nuclear Power Program as a project manager on several small nuclear reactor projects. From 1966 to 1972, I worked in the Space Nuclear Reactor Program, first for General Electric, then for the Jet Propulsion Laboratory. I performed conceptual core designs and overall plant designs for compact direct energy conversion, and advanced reactor concepts in the 100 KWe to 10 MWe range.
Since 1972, I have been involved in BWR plant technolo-gy, all but one year at General Electric. My career has exposed me to almost all aspects of plant performance. I
- _s have performed and managed advanced fuel element and core design studies to increase performance of the uranium fuel cycle. I have also performed plant system performance studies such as cost / performance studies of varying core
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power density, ECCS performance and studies to enhance plant availability.
After the TMI accident, I headed a task force which performed all related studies, including risk assessments and evaluations, of proposed design changes. I have also managed all transient and accident analyses for BWR's under construction.
I am a member of the American Nuclear Society and a Registered Professional Engineer (Nuclear) in the State of i California. I have authored several papers on BWR design and operating characteristics.
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