ML20009E315

Transcript of Rl Call 810720 Testimony on Doherty Contention 26 (Stud Bolts).Prof Qualifications Encl

ML20009E315
Person / Time
Site:	Allens Creek File:Houston Lighting and Power Company icon.png
Issue date:	07/20/1981
From:	Call R GENERAL ELECTRIC CO.
To:
References
NUDOCS 8107280037
Download: ML20009E315 (9)

Text

July 20, 1981 d \ \ lI / f 1 UNITED STATES OF AMERICA [1;/ -,

JUL- "' NDCLEAR REGULATORY COMMISSION $ $-

2 I Cri'ICL v; ..BEFORE.TbE ATOMIC SAFETY AND LICENSING BOARD

" $' $93I > Q 3 & L . Jit. o :.wIm I,

8 4 In the Matter of S 5 liOUSTON LIGHTING & POWER COMPANY S Docket F K'r . M.,

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(Allens Creek Nuclear Generating -

Station, Unit 1) S p

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DIRECT TESTIMONY OF REY L. CALL ,

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ON DOHERTY CONTENTION 26 - STUD BOLTS 9

M Q. Pleu.se state your name and place of employment.

A. My name is Rey L. Call and I am employed by the General Electric Company as Principal Engineer in the Availability Engineering Subsection. Also, it is important to note that I was the Manager of GE's Reactor 2ressure 14 Vessel Design unit at the time the Allens Creek RPV was 15 being fabricated.

16 Q. Would you describe your professional qualifications?

17 A. A copy of my professional qualifications are set 18 forth as Exhibit RLC-1 to tls testimony.

19 What is the purpose of your testimony?

Q.

20 A. The purpose of my testimony is to address Mr.

21 Doherty's Contention No. 26 which asserts that the reactor 22 vessel and stud bolts are inadequate in the following areas:

23 a) Yield strength is exceeded during anticipated 24 DM3 5

8107280037 PDR ADOCK 05000466810720 / f T

PDR ,

1 transient without scram (ATWS) 2 b) Primary tensile component of the total bolt 3 stress has been inadequately considered.

4 c) The total stress in the bolt as well as the 5 strain energy in the reactor vassel head flange are such 6 that the failure of one stud bolt will cause the failure of 7

all the stud bolts.

g Q. Have any stud bolts ever failed in a BWR?

^' * "Y "" "'* "" ' "** """**'

9 at the La Crosse BWR. That failure was attributable to 10 insufficient fracture toughness resulting from faulty heat treatment in combination with a corrosive attack on the bolts 12 caused by the breakdown of the silver plating. These prob-13 lems have been addressed by the NRC and resulted in 14 Regulatory Guide 1.65. Allens Creek complies with Reg. Guide 15 1.65 as stated in Appendix C of the PSAR.

16 Q. What criteria govern the design of the Reactor 17 Pressure Vessel (RPV) stud bolts?

18 A. The design of the RPV stud bolts is governed by 19 Title 10 of the Code of Federal Regulations, Part 50.55a and 20 Appendix A, Criterion 30, which requires the use of the ASME 21 Boiler and Pressure Vessel Code, Division I,Section III and 22 Section XI. In additior. Regulatory Guide 1.65 issued 23 october 1973 supplements the requirements of the ASME Code, 24

, a

. o 1 Section III and Section XI.

2 Q. Of what material are the bolts made?

3 A. The bolts are a special quality Chromium-Nickel-4 Molybdenum alloy steel which has been specified for use in 5 nuclear applications by the ASME Code. The ASME Code speci-6

~.ication for this material is SA540, elass 3, Grade B23 or 7

B24. This specification places requirements on mechanical g

properties, chemical content, heat treatment, and on many g

other aspects connected with the manufacture of the niaterial.

Regulatory Guide 1.65 supplements this ASME specification.

Q. What mechanical property requirements of the ASME Code and Regulatory Guide 1.65 are pertinent to this contention?

13 A. There are basically four material properties cf 14 importance: (l' yield strength, which is a measure of the 15 stress at which permanent strain deformation wil.' occur; 16 (2) tensile strength which is a measure of the maximum 17 nominal stress a specimen supports in a tension test prior 18 to fracture; (3) fracture toughness, which is a measure of 19 the material's resistance to the extension of a crack; and 20 (4) hardness, which is a measure of the material's resistance 21 to penetration.

22 Q. What tests are performed to guarantee that the 23 stud bolts will have adequate material properties?

24 A. For tensile and fracture toughness testing one

1 test is made for each " lot" of material. A lot is one batch 2

of material heat treated in the same charge or as one 3

continuous operation. Hardness tests are performed on each 4 utud after heat treatment but before final machining. A 5 chemical analysis of the stud material is also done to check 6 that the chemical composition of the material meets the 7 material specification requirements.

8 Q. Besides mechanical properties, what other properties 9 or processes are considered in fabricating stud bolts in 10 order to prevent failure in service?

11 A. In order to resist corrosion, cach stud bolt has a 12 maganese phosphate coating which is applied as a measure to 13 Prevent oxidation. Regulatory Guide 1.65 specifically en-d rses this material for use as a coating. The thread 14 lubricant, which is a colloidal suspension of graphite in 15 isopropyl alcohol, is also in compliance with Regulatory 6

Guide 1.65.

17 To provide resistance to stress corrosion cracking, the ACNGS stud bolts are tempered to a maximum t- sile strength of 170 KSI. The 170 KSI tensile limit is acluded 20 in the bolt design and ensured by prescribed testing.

21 The bolts will undergo fabrication inspections, 22 pre-service inspections, and inservice inspections. The 23 fabrication inspection is done in accordance with ASME 24 1 .

1 Section III, Subarti:le NB-2580, as supplemented by Regulatory 2 Guide 1.65. Included are a visual examination, and either 3 a magnetic particle test (MT) or a liquid penetrant test 4 (PT), after final heat treatment. The pre-service inspection is done in accordance with ASME Section XI, which requires a 5

MT surface examination and a volumetric, ultrasonic test (UT) 6 x min ti n. The pre-service inspection serves as a baseline 7

f r future inservice inspection. The inservice inspections 8

g are also governed by ASME Section XI and involve MT and UT examination.

10 Finally the reactor pressure vessel is required to be statically pressurized with water to an overload condition of not less than 1.25 times design pressure. The ACNGS 13 vessel was tested at 1600 psig and no defects were found.

14 This was done in the fabrication shop and will be done again 15 as a pre-operational test at the site. This demonstrates 16 that the stud bolts as well as the vessel can withstand a 17 severe static overload condition.

18 Besides ensuring that the stud bolts meet or exceed Q.

19 minimum design standards for mechanical, structural and 20 chemical properties, what else must be done to assure that 21 the bolts are properly designed?

22 A. A stress analysis of the bolts must be done to 23 ensure that the bolts will be able to carry the imposed load-24 ings. Subarticle NB-3230 of Section III of the ASME Code 1 ,

governs the stress analysis for the bolts.

11 2 Q. Referring to subpart (a) of the contention, has 3 ATWS been considered in the design of the RPV stud bolts?

4 A. General Electric has included an event equivalent 5

in all p rtinent respects to the :ATWS event in the reactor design. The equivalent event is the " Reactor Overpressure 6

with Delayed Scram" and is documented in the Reactor Cycle l 7

Drawing, GE Document #762E458. For ATWS and delayed scram g

the peak pressure in the upper region is 1460 psig.

Q. For ATWS, what is the maximum bolt stress at the 10 periphery and the average service stress and how does this 11 compare to the Code allowables for an ATWS event?

12 A. For ATWS, the maximum bolt stress is 64 KSI, while 13 the average service stress is 41.4 KSI. These stresses 14 occur at a temperature of 575 F. The respective ASME Code 15 limits are 98 KSI and 72.6 KSI at 575'F. The stresses due 16 to ATWS are well within the ASME Code allowables and are 17 considerably below minimum yield strength.

18 Referring to subpart (b) of the contention, does Q.

19 ATWS impose the highest primary tensile stress on the stud 20 bolts.

21 A. Yes.

22 Q. Referring to subpart (c) of the contention, will 23 the failure of one stud bolt cause the failure of all the 24 stud bolts?

1 A. No. ACNGS has 72 stud bolts each with a cross-2 2 sectional area 27.97 in . The total bolt crosssectional 3 area is 2014 in2 This is the area which was used in the

' tress analysis. For the actual loads which are imposed on 4 -

3 the bolts, the minimum crosssectional area needed to carry 2 This equals approximately 58 bolts.

6 the load is 1611.5 in .

In short, it is extremely improbable that the failure of one 7

bolt will cause a failure of all the other bolts, as alleged 8

9 Q. Can you summarize your testimony?

A. In summary, the bolt design meets all the require-ments of the ASME Code and Reg. Guide 1.65. The bolting material is a high strength alloy material with excellent 13 fracture toughness properties and is inherently resistant to 14 stress corrosion cracking. No metallic platings are used.

15 j The mechanical properties are extensively tested to ensure 16 The the minimum acceptable properties of the .aterial.

17 bolts are also extensively inspected during fabrication and' 18 in service. All stresses in the bolts are within the 19 conservative ASME Code allowables.

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i 23 ,

24 l

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.. Attachment RLC-1 RESUEE NAME: Rey L. Call ADDRESS: 4640 Royal' Grove Court, San Jose, CA 95135 BIRTHDATE: 11 December 1921 BIRTHPLACE: Bountiful, Utah, USA PROFFSSIONAL REGISTRATION: Mechanical, California, since 1965 SOCIETY MEMBERSHIP: Member, American Society of Mechanical Engineers (ASME)

EDUCATION: BSME, University of Utah, 1950 l MSME, University of Idaho, 1961 E,FTLOYER: General Electric Co.

175 Curtner Avenue San Jose, CA 95125 TITLE: Principal Engineer Availability Engineering WORK EXPERIENCE:

I 4/80 to General Electric Co.

Present San Jose, CA l l

Principal Engineer - Availability Engineering Program Manager for Failure Modes and Ef fects Analysis preparation for a nuclear l reactor project. Prepare reliability and availability studies of reactor components and systems.

1/80 to General Electric Co.

l 4/80 San Jose, CA Principal Engineer - RPV & Internals Design i Certified RPV design specifications in accordance with ASME B&PV Code,.Section III I Reviewed and approved vendor design drawings and stress analyses. Member, Work Group - Vessels,Section III of ASME B&PV Code Committee, i

7/75 to General Electric Co.

1/80 San Jose; CA

! Manager - RPV Design Unit j Responsible for mechanical design of reactor pressure vessels. Planned and i performed reactor pressure vessel design to the level of detail necessary to control the design. Utilized design components and tools provided by others.

Prepared design drawings and specifications required for manufacture and procurement of reactor pressure vessels. Maintained the standard designs.

Provided interfaces for construction, erection, testing and operation of the j reactor pressure vesseln.

7/72 to CBI Nuclear Co.

7/75 Memphis, TN Supervisor-Development Group Responsible for development work within Engineering in CBI Nuclear. Supervised two engineers in their work on development contracts. Reviewed proposals for development work, estimated costs and prepared Application & Authorization forms for approval of development contracts.

10/66 to General Electric Co.

7/72 San Jose, CA Senior Engineer Prepared reactor pressure vessel design specifications. Certified these specifica-tions in accordance with Section III, ASME B6PV Code. Developed reactor vessel thermal cycle drawings for the design specs. Reviewed vendor drawings and stress analyses. Designed reactor core structures.

8/62 to General Electric Co.

10/66 San Jose, CA Engineer I - Irradiation Processing Operation, Vallecitos, CA)

Designed irradiation capsules for testing in GETR and MTR in Idaho. Designed structures and mechanical equipment for sampling stations for an experimental power reactor. Was lead engineer in design hardware (including pressure vessel) for a plutonium subcritical experiment. Supervised work of three engineers.

9/52 to General Electric Co.

8/62 Richland, WA Engineer II & Engineer I Designed and developed mechanical equipment for nuclear reactors. Monitored the reactor operations at two reactor sites to audit compliance with engineering standards and specifications. Wrote engineering standards and specifications for reactor operation. Performed design and development work on reactor fuel elements.

Designed and conducted irradiation testing of reactor fuel elements.

9/45 to U. S. Air Force 9/46 (Ogden Air Materiel Con =and, Ogden, UT)

Aircraft Mechanic

>HLITARY SERVICE:

9/50 to U. S. Army 9/52 Assistant Battalion Survey Officer in a field artillery observation battalion.

7/42 to U. S. Navy 9/45 AMM2/C - Aircraft Mechanic J