ML19347F534

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Testimony of Ranganath Re Tx Pirg Contention 39,Generic Task A-11,fracture Toughness.Prof Qualifications Encl. Related Correspondence
ML19347F534
Person / Time
Site: Allens Creek File:Houston Lighting and Power Company icon.png
Issue date: 05/11/1981
From: Ranganath S
GENERAL ELECTRIC CO.
To:
Shared Package
ML19347F516 List:
References
NUDOCS 8105190551
Download: ML19347F534 (9)
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Text

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UNITED STATES OF AMERCIA T

1 NUCLEAR REGULATORY CO!D'ISSION c,

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9 BEFORf, THE ATOMIC SAFETY AND LICENSING BOARD C 3

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4 In the Matter of

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5 liOUSTON LIGHTING & POWER COMPANY

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Docket No. 50-466

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6 (Allens Creek Nuclear Generating

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Station, Unit No. 1)

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7 8

TESTIMONY OF SAM PANGANATH ON BEHALF OF HOUSTON LIGHTING & POWER COMPANY ON 9

TEXPIRG CONTENTION 39 GENERIC All -

FRACTURE TOUGHNESS 10 11 Q.

Please state your name and place of employment.

12 A.

My name is Sam Ranganath and I am employed as Manager, 13 Stress and Fracture Analy.As Unit at the General Electric 14 Company.

15 Q.

Would you describe your professional qualifications?

My professional qualifications are set forth in Exhibit 16 A.

17 SR-1 to this testimony.

18 Q.

What is the purpose of your testimony?

19 A.

The purpose of this testimony is to address TexPirg's 1

20 contention that ACNGS should not be licensed until the 21 generic safety issue denoted by the Staff as Task A-ll, 22 which concerns reactor vessel materials toughness, has 23 been resolved.

24 Q.

Would you please describe the problem identified 25 by the Staff in task A-ll?

26 A.

les.

In NUREG-0371, the Staff noted that preventing 27 a failure of the reactor vesse) " depends primarily on 28 maintaining the reactor vessel material fracture toughness 8105190'Es51

1 2 at levels that will resist brittle fracture during plant 3

operation."

The Staff stated that the longer a ralant 4

is in operation, the continued neutron irradiation reduces 5

fracture toughness of the pressure vessel material which 6

in turn could reduce the safety margins.

According to 7'

the Staff, this problem is an issue only in certain older 8

operating PWR plants.

9 Q.

What is the current NRC criteria for fracture toughness 10 which ACNGS must mee't?

11 A.

The Commission first proposed rules for establishing 12 pressure-temperature operating limits in 1971.

The Commis-13 sion's regulations relating to reactor vessel fracture 14 toughness are contained in Appendices G and H to 10 CFR Part 15 50.

Appendix G specifies the appropriate fracture 16 toughness requirements for the carbon and low-alloy 17 ferritic steel pressure vessel components such as the 18 Allens Creek react,r pressure vessel for welds and weld f

19 heat-affected zones (HAZ), and for material used for bolting 20 with specified yield strengths not over 130,000 psi.

21 Appendix H covers the material surveillance program required 22 l

to monitor changes in the fracture toughness properties 28 of ferritic mau rials due to exposure to neutron irradiation.

24 The prediction of the amount of change in material properties 25 is given in Regulatory Guide 1.99.

The Allens Creek vessel 26 meets all the above fracture roughness requirements.

27 l

28

1 <

2' In addition, compliance with Section III, Appendix G 1

3 of the ASME B&PV Code is required under Appendices G and 4

H to Part 50.

The ASME Code gives the details for determining 5

the fracture toughness of a material through testing.

The 6

rules for establishing the pressure-temperature limit 7

curves for reactor operation are prescribed by 10 CFR 50, 8

Appendix G.

9 Q.

What are the fracture toughness requirements of the 10 ASME B&PV Code?

11 A.

The testing methods used to determine fracture toughness 12 and the methods used to determine the pressure-temperature 13 operating limits contained in 10 CFR 50, Appendix G and 14 ASME B&PV Code,Section III, Appendix G, are based on 15 detailed structural and material analytical methods known 16 as fracture mechanics that have been developed over the 17 past 20 or more years.

These methods have been widely 18 used in the aerospece industry and are well proven and 19 accepted.

The fracture mechanics requirements define the 20 minimum toughness of the reactor vessel material and assure 21 a nominal safety factor of two against failure with a very 22 large flaw located in the most highly radiated area.

In 23 addition, the portion of the vessel subject to significant 24 radiation is designed to avoid fatigue conditions that 25 could cause formation or gra th of cracks.

26 Q.

Has the ACNGS reactor vessel been fabricated in 27 accordance with the requirements of Appendix G of Part 50, 28 and Section III of the ASME B&PV Code?

l

1,

2 A.

Yes.

The ACNGS reactor vessel meets the require-3' ments of the. Commission's regulations as described above.

4 Q.

How are irradiation effects allowed for in the design 5

and operation of Allens Creek?

6 A.

As the ACNGS vessel is irradiated, the permitted 7

pressure-temperature limits must shift to account for the 8

radiation effects and to maintain the margin of safety.

9 Most of the neutron radiation damage over core life occurs 10 in the " beltline", that part of the cylindrical sheel of 11 the reactor vessel directly opposite the core.

The beltline 12 usually contains at least 2-shell courses and several welds.

13 By design, there are no nozzles, flanges or changes in 14 thickness of the shell of the vessel in the most highly 15 irradiated region.

Thus, the design avoids regions of 16 high stress in the beltline region.

17 Q.

Can you predict the amount of radiation damage which 18 wil occur over a period of time?

19 A.

The amount of radiation damage after a given amount of 1

20 service can be predicted on a conservative basis using 21 Regulatory Guide 1.99.

Regulatory Guide 1.99 gives the i

1 22 change in material toughness in terms of a reference 23 temperature, RT r.nd a decrease in upper shelf energy.

NDT 24 The reference temperature, RT s related in de ASE NDT 25 Code Appendix G to the reference stress intensity K 1

73 26 for ferritic steels such as SA 533 Gr.B which is used in 27 the ACNGS vessel.

This stress intensity is a fracture 1

28 toughness parameter which is a direct measure of the l

l

1,

2 resistance to brittle fracture under the imposed loadings.

3 Therefore, RT can be used as a measure of the change in NDT 4

fracture toughness.

The upper shelf energy level is the 5

point at which a fracture of the material is fully ductile.

6 It is the p'. int of maximum material toughness capability.

7 The value of both of these paremeters change with neutron 8

irradiation and Regulatory Guide 1.99 gives the change as a 9

function of neutron fluence and material composition.

In 10-order to minimize the changes in RT and Upper shelf NDT 11 energy, the Allens Creek vessel specifications have restric-12 tions on the copper and phosphorus content of the core 13 beltline base metal and weld metal.

In addition, minimum 14 upper shelf energy level requirements are placed on all core 15 beltline materials.

16 Q.

How will ACNGS comply with the requirements of 17 Appendix H to Part 50?

18 A.

The prediction of change in RT and upper shelf energy NDT 19 levels is checked by the surveillance testing as required 20 by Appendix H to 10 CFR 50.

The ACNGS reactor vessel 21 materials surveillance specimens are provided in accordance 22 with requirements of ASTM E 185-73 and 10 CFR S0 Appendix H.

23 Materials for the program are selected to represent materials 24 used in the reactor beltline region.

Specimens are menu-25 factured from a plate actually used in the beltline region N

and a weld typical of those in the beltline region and thus N

represent base metal, weld material, and the weld heat 28' affected zone material.

The plate and weid are heat treated il

i 1.

2 in a manner which simulates the actual heat treatment per-3 formed on the core region shell plates of the complete vessel.

4 Each in-reactor surveillance capsule contains 36 5

Charpy-V-Notch specimens.

The capsule loading consists of 6l' 12 specimens each of base metal, weld metal, and heat 7

affected zone material.

A set of out-of-reactor baseline 8

Charpy-V-Notch specimens and archive material are provided 9

with the surveillance test specimens.

10 The surveillance specimen capsules are located at 11 three aximuths at a common elevation in the core beltline 12 region.

The sealed capsules are not attached to the vessel 13 but arc in welded capsule holders.

The capsule holders 14 are mechanically retained by capsule holder brackets welded 15 to the tessel cladding.

The capsule holder brackets allow 16 the removal and reinsertion of capsule holders.

These 17 brackets are designed, fabricated and analyzed to the require-18 ments of Section III ASME Code.

A positive spring loaded 19 locking device is provided to retain the capsules in 20 position throughout any anticipated event during the li fe-21 time of the vessel.

The number of capsules and the proposed 22 withdrawal schedule is in accordance with 10 CFR 50 Appendix H.

28 Following withdrawal, the specimens are tested to confirm the 24 predicted change in RT and upper shelf energy.

NDT 25 Q.

Is the generic safety issue A-ll applicable to new 26 plants which meet the current NRC requiremcats on fracture 27 toughness?

28 A.

The material toughness ccncerns stated in Task A-ll

1.

2 are not applicable to new plants which meet the current 3

NRC requirements of Appendices G and H.

The ACNGS vessel 4

meets all the toughness criteria of Appendices G and H F

(including upper shelf requirements), has restrictions in 6

the copper and phosphorus content and conservatively 7

accounts for radiation effects chrough Regulatory Guide 1.99.

8 Further, the low neutron fluence in the vessel beltline 9

region for the BWR assures that the reduction in toughness 10 with radiation is small.

Therefore the A-ll problem is 11 not a relevant issue for the ACNGS.

12 The A-ll problem has been an issue only in certain 13 older operating PWR plants with inadequate control on 14 copper and phosphorus and without upper shelf toughness 15 requiroments in the beltline material.

The Staff notes in 16 NUREG-0371 that even for these PWR reactor vessels, appropriate 17 safety margins can be maintained by shifts of the operating 18 pressure-temperature limits as dictated by the results of 19 the material surveillance programs described in Appendix H 20 and Regulatory Guide 1.99.

In fact, udequate toughness 21 margins can be defined using the current Appendix G methods 22 which are based on the conservative assumption of elastic 23 behavior.

The advanced fracture mechanics methods being 24 developed in the Task A-ll program consider the elastic-25 plastic behavior and will provide more exact predictions 26 indicating additional fracture margin.

It is therefore not 27 surprising that the Staff finds the currer.t licensing 28 criteria sufficient to provide complete assurance that

1 2 reactor vessels for plants now in the licensing process, 3

including ACNGS, will have adequate margins of safety 4

against brittle fracture failure.

5 Q.

What are your conclusions?

6 A.

Large fracture margins have been incorporated in the 7

ACNGS design by meeting the NRC fracture toughness require-8 ments. by restricting the copper and phosphorous content in i the vessel beltline material and by conservatively accounting 9

10-for radiation effects through Regulatory Guide 1.99.

The 11 low neutron fluence in the BWR provides additional assurance 12 of fracture margin.

Therefore, the toughness concerns 13 stated in the Task A-ll problem are not relevant to ANCGS.

14 While the infcrmation derived from the Task A-11 15 advanced fracture mechanics program pertinent to current 16 vessel design may be useful, for ACNGS it will only provide 17 confirmation of even larger fracture toughness margins for 18 ACNGS than predicted by the current 10 CFR 50 Appendix G 19 methods.

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21 l

22 l

23 24 25 26 27 28

e Exhibit SR-1 1

1 EDUCATION AND PROFESSIONAL QUitLIFICATIONS 2

Dr.

S.

Ranganath 3

4 Dr. Ranganath is the Manager of the Stress and Fracture Analysis Unit in the Nuclear Energy Business Group at General 5

6 Electric.

In this capacity, he is responsible for the structural mechanics and fracture mechanics /the fatigue evaluation of BWR 7

g pressure vessel components.

He has been active in the area of fracture mechanics and fatigue behavior of structures and 9

10 has published several papers in his field.

He is a member of the ASME Code Section XI sub-group on evaluation and standards.

11 This group develops the fracture mechanics evaluat'.nn methods 12 and acceptance standards for nuclear pressure vessel components.

13 Dr. Ranganath is also an Adjunct Lecturer at the University

}4 of Santa Clara and teaches graduate coursesin pressure vessel 15 16 design and fracture mechanics.

Ranganath received his Ph.a from Brown University, 17 Dr.

18 Providence, Rhode Island in 1971.

He received his MS from

'ndian Institute of Technology, Bangalore, India in 1967.

19 He is a rc,istered professional engineer in the State of 20 21 California.

22 23 24 25 26 27 28

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