ML20136F326
| ML20136F326 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Vogtle |
| Issue date: | 10/09/1984 |
| From: | Johnston W Office of Nuclear Reactor Regulation |
| To: | Novak T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML082840446 | List:
|
| References | |
| FOIA-84-663 NUDOCS 8410170084 | |
| Download: ML20136F326 (11) | |
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jg UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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OCT C 9 1334
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Docket Nos. 50-424/425 MEMORANDUM FOR: Thomas M. Novak, Assistant Director for Licensing Division of Licensing FROM:
William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering
SUBJECT:
GEORGIA POWER COMPANY - V0GTLE ELECTRIC
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GENERATING PLANT, UNITS 1 AND 2 Plant Name: Vogtle Electric Generating Plant, Units 1 and 2 (VEGP)
Supplier: Westinghouse, Bechtel Docket Nos: 50-424/425 L
Responsible Branch: Licensing Branch #4
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Project Manager:
M. Miller Reviewer:
B. J. Elliot (INEL) a'-
Description of Task: Draft Safety Evaluation Report for i
I Sections 5.3.1, 5.3.2, and 5.3.3 Review Status: Additional Information Required The Materials Application Section, Materials Engineering Branch, Division of Engineering, with the assistance of Idaho National Engineering Labora-tory, has reviewed the Final Safety Analysis Report for VEGP. Based on our review of this information, we have prepared our input to the Safety Evaluation Report (Attachment 1).
In this safety evaluation, we have identified the area for which sufficient information has not been sub-mitted to determine ccmpliance with Appendices G and H, 10 CFR 50. The areas, where sufficient information has not been provided, will remain open items until Georgia Power Company provides the necessary information.
i The specific information required to resolve these open items are contained in Attachment 2.
t William V. Johnston, Assistant Director Materials, Chemical & Environmental Technology Division of Engineering l-Attachments: As stated i
cc: See Page 2
Contact:
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Elinor Adensam OCT 0 91984 cc:
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W. Johnston T. Novak E. Sullivan S. Pawlicki
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ATTACHMENT I GEORGIA POWER COMPANY V0GTLE ELECTRIC GENERATING PLANT UNITS 1 and 2 DOCKET NUMBERS40-424 and 50-425 REACTOR COOLANT PRESSURE BOUNDARY FRACTURE TOUGHNESS l
5.3.1 Reactor Vessel Material h
The fracture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, and the materials surveillance p:'ogram for the reactor vessel beltline have been reviewed.
The acceptance criteria and references f
which are the basis for this evaluation are set forth in paragraphs II.5, II.6, and II.7 (Appendices G and H, 10 CFR Part 50) of SRP Section 5.3.1 in NUREG-0800 Rev. 1 dated July 1981.
A discussion of this review follows.
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General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," Appendix A, 10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed with sufficient margin to ensure that, when stressed under operating, maintenance, and test conditions, the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized.
General Design Criterion 31, " Inspection of Reactor Coolant Pressure Boundary," Appendix A, 10 CFR Part 50, requires, in part, j
that the reactor coolant pressure boundary be designed to permit an appropriate
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material surveillance program for the reactor pressure vessel.
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The fracture toughness requirements for the ferritic materials of the reactor
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coolant pressure boundary are defined in Appendix G, " Fracture Toughness I
Requirements," and Appendix H, " Reactor Vessel Material Surveillance Require-
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ments" of 10 CFR Part 50.
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-10/04/84 5-1 V0GTLE SER INPUT SEC 5.3.1 i
Compliance to Section 50.55(a), 10 CFR Part 50 The Edition and Addenda of the ASME Code that are applicable to the design and i
fabrication of the reactor vessel and reactor coolant pressure boundary (RCPB) components are specified in Section 50.55(a) of 10 CFR Part 50.
The ASME Code Edition and Addenda that are required depend upon the date the construction permit was issued. The Vogtle Electric Generation Plants, Units 1 and 2 (VEGP), construction permit was issued on June 28, 1974.
Based upon the construction permit date,10 CFR Part 50 Section 50.55(a) requires that ferritic materials used for the VEGP reactor vessels be designed and constructed to editions that are no earlier than the Winter 1971 Addenda to the 1971 ASME Code (hereafter Code) and that ferritic materials used in piping, pumps, and valves be constructed to editions that are no earlier than the Winter 1972 Addenda to the Code.
The VEGP ferritic materials meet all the above require-l ments.
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Compliance with Appendix G, 10 CFR Part 50 q.
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Our evaluation of the VEGP FSAR to determine the degree of compliance with the fracture toughness requirements of Appendix G, 10 CFR Part 50, indicates that the Applicant has met all the requirements of this Appendix except as discussed below.
Appendix G requires that for the reactor beltline materials the Charpy V-notch (C ) impact tests shall be conducted at appropriate temperatures over a tempera-y ture range sufficient to define the C test curves (including the upper-shelf y
levels) in terms of both fracture energy and lateral expansion of specimens.
The VEGP FSAR (Tables 5.3.1-2 and 5.3.1-3) contain upper-shelf Charpy V-notch (C ) impact test data for the reactor beltline materials, but do not have the y
C curves in terms of fracture energy and lateral expansion.
To fully comply y
with Appendix G, the Applicant must supply the impact test data and the C y curves for the beltline materials.
Until the Applicant provides the data and curves we cannot complete our review of this item.
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j Compliance With Appendix H, 10 CFR Part 50 1
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The materials surveillance program at VEGP will be used to mcnitor changes in j
the fracture toughness properties of ferritic materials in the reactor vessel
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beltline region, resulting from exposure to neutron irradiation and the thermal environment as required by General Design Criterion 32, " Inspection of reactor i
Coolant Pressure Boundary." The VEGP surveillance program, which must be
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compliance with Appendix H, 10 CFR Part 50 and ASTM E 185, " Standard Recommended 2
Practices for Surveillance Tests for Nuclear Reactor Vessels," requires i
j fracture toughness data be obtained from material specimens that are representa-
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tive of the limiting base, weld, and heat-affected zone materials in the belt-
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line region.
These data will permit the determination of the conditions under 1
which the vessel can be operated with adequate margins of safety against l
fracture throughout its service life.
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Based on our review of the Applicant's submittal that detailed the extent of the compliance of VEGP with Appendix H, 10 CFR Part 50 and ASTM E 185, we have determined that these requirements have been met except as follows.
,(Q To determine the effect that neutron irradiation has on the reactor vessel, samples from the limiting material should be placed into the surveillance capsules.
The limiting materials for the VEGP Unit 2 reactor vessel beltline j
is plate material from heat 88628-1 and material from wald G-1.60.
The materials in VEGP Unit 2 surveillance capsules are from plate 88628-1 and weld E-3.23.
Because the VEGP Unit 2 surveillance weld material is not the most limiting material, the Applicant's surveillance program will not completely monitor the 3
extent of radiation damage to the weld material in VEGP Unit 2.
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Although the limiting weld metal is not contained in the Applicant's surveil-h lance program for VEGP Unit 2, the Applicant will be required to determine the effect of neutron irradiation damage on the limiting weld metal using Regulatory
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Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to
.hy Reactor Vessel Materials." We have found that the methods of predicting l
neutron irradiation damage that are documented in Regulatory Guide 1.99 are l
conservative.
Hence, the use of this guide to predict neutron irradiation 3
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damage is an acceptable alternative to testing the limiting weld metal as part 1
of the surveillance program.
1 ij ASTM E 185 also requires that the withdrawal schedules for the surveillance j
specimens be defined.
Table 2 of Amendment 5 of the Applicant's FSAR states j
that the withdrawal schedules will be in accordance with the VEGP Technical j'
Specification.
Until the Applicant defines the withdrawal schedules we will not be able to complete our review of this part of the FSAR.
Until the Applicant provides the information previously discussed, we cannot I
conclude there is reasonable assurance that the surveillance program will g
monitor the change in the beltline region material properties to the extent l
required for establishing pressure-temperature limits and to preserve the I
integrity of the vessel.
This program must generate sufficient information to permit the determination of conditions under which the reactor vessel will be operated with an adequate margin of safety against rapidly propagating fracture throughout its service lifetime.
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Conclusions for Compliance with Appendices G and H, 10 CFR Part 50 Appendix G, " Protection Against Non-Ductile Failures,"Section III of the ASME Code, was used, together with the fracture toughness test results required by Appendices G and H, 10 CFR Part 50, to calculate the pressure-temperature limitations for the VEGP reactor vessels.
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The fracture toughness tests required by the ASME Code and by Appendix G of 10 CFR Part 50 provide reasonable assurance that adequate safety margins against the possibility of non-ductile behavior or rapidly propagating fracture can be established for all pressure-retaining components of the reactor coolant boundary.
The use of Appendix G,Section III of the ASME Code, as a guide in establishing safe operating procedures, and use of the results of the fracture l
j toughness tests performed in accordance with the ASME Code and NRC regulations, will provide adequate safety margins during operating, testing, maintenance, j
and anticipated transient conditions.
Compliance with these Code provisions s
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10/04/84 5-4 V0GTLE SER INPUT SEC 5.3.1
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~m and NRC regulations constitutes an acceptable basis for satisfying the require-ments for General Design Criterion 31.
The materials surveillance program, required by Appendix H, 10 CFR Part 50, will provide information on the effects of irradiation on material properties so that changes in the fracture toughness of the material in the VEGP reactor vessels beltline can be properly assessed, and adequate safety margins against the possibility of vessel failure can be provided.
Compliance with Appendix H, 10 CFR Part 50 and ASTM E 185 assures that the surveillance program will be capable of monitoring radiation induced changes in the fracture toughness of the reactor vessel material and satisfies the requirements of General Design Criterion 32.
There is reasonable assurance that the surveillance program will monitor the change in the beltline region material properties to the extent required for
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establishing pressure-temperature limits and to preserve the integrity of the vessel.
The surveillance program will generate sufficient information to t
permit the determination of conditions under which the reactor vessel will be operated with an adequate margin of safety against rapidly propagating fracture throughout its service lifetime.
5.3.2 Pressure Temperature Limits The Applicant's pressure temperature limits for operation of the reactor vessel have been reviewed.
The acceptance criteria and list of references which are the basis for this evaluation are set forth in the Standard Review Plan (SRP) Section 5.3.2 of NUREG-0800 Rev. I dated July 1981.
A discuss' ion of this review follows.
Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," 10 CFR Part 50, describe the conditions that require pressure-temperature limits and provide the general bases-for these limits.
These appendices specificallly require that pressure-temperature limits must provide safety margins at least as great as those 10/04/84 5-5 V0GTLE SER INPUT SEC 5.3.1
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recommended in the ASME Code,Section III, Appendix G, " Protection Against s
Non-Ductile Failures." Appendix G, 10 CFR Part 50, requires additional safety t
j margins for the closure flange region materials and beltline materials whenever the reactor core' is critical, except for low-level physics tests.
f The following pressure-temperature limits imposed on the reactor coolant pressure boundary during operation and tests are reviewed to ensure that they provide adequate safety margins against non-ductile behavior or rapidly propa-
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gating failure of ferritic components as required by General Design Criterion 31:
a.
Preservice hydrostatic tests i
b.
Inservice leak and hydrostatic tests
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Heatup and cooldown operations f
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Core operation.
^ gith En 1 Appendices G and H, 10 CFR Part 50, require the Applicant to predict the amount of increase in reference temperature, RT due to neutron irradiation.
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The shift in RT due to neutron irradiation is then added to the initial NOT RT to establish the adjusted reference temperature.
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l The pressure-temperature curves for the reactor vessel closure flange areas were not included.
1 The Applicant must revise his pressure-temperature curves to include the i
safety margins specified in Appendix G, 10 CFR Part 50, for the flanga closure region. We cannot complete our evaluation of the pressure-temperature limits for VEGP until the Applicant revises the pressure-temperature curves to include the closure flange regions.
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The pressure-temperature limits to be imposed on the reactor coolant system f
for all operating and testing conditions must have adequate safety margins against non-ductile or rapidly propagating failure, and must be in conformance with established criteria, codes, and standards.
The use of oprating limits 10/04/84 5-6 V0GTLE SER ItiPUT SEC 5.3.1 t
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based on these criteria, as defined by applicable regulations, codes, and standards, will provide reasonable assurance that non-ductile or rapidly propagating failure will not occur, and will constitute an acceptable basis for satisfying the applicable requirements of General Design Criterion 31.
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5.3.3 Reactor Vessel Integrity b
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Although most areas are reviewed separately in accordance with other review
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plans, reactor vessel integrity is of such importance that 4 special summary
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h review of all factors relating to reactor vessel integrity is warranted.
In this section, we have reviewed the fracture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, the pressure-temperature
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limits for operation of the reactor vessel, and the materials surveillance program for the reactor vessel beltline. The acceptance criteria and references which are the basis for the evaluation are set forth in paragraphs II.2, II.6, I
and II.7 (Appendices G and H, 10 CFR Part 50) of Standard Review Plan ( SRP)
Section 5.3.3 in NUREG 0800 Rev. 1 dated July 1981. A discussion of this review follows.
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'We have reviewed the information in each area to ensure that it is complete and that no inconsistencies exist that would reduce the certainty of vessel integrity. The areas reviewed are:
1.
Design (SER 5.3.1) 2.
Materials of construction (SER 5.3.1) 3.
Fabrication methods (SER 5.3.1) 4.
Operating conditions (SER 5.3.2).
I We have reviewed the above factors contributing to the structural integrity of the reactor vessel and conclude that the Applicant has complied with Appendices G
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and H, of 10 CFR Part 50, except for the following items:
10/04/84 5-7 V0GTLE SER INPUT SEC 5.3.1
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a.
The Applicant has not reported the Charpy V-notch energy and mils
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i lateral expansion data versus temperature for each reactor vessel beltline material.
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b.
The Applicant has not reported the withdrawal schedule for the surveillance specimens.
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c.
The Applicant has not supplied pressure-temperature curves for the f
reactor vessel pressure closure flange regions.
Until the Applicant supplies the information necessary to complete our evalua-tion of compliance to Appendices G and H, 10 CFR Part 50, we cannot complete our evaluation of the structural integrity of the VEGP reactor vessels.
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+a,.- s m ATTACHMENT 2 REQUEST FOR ADDITIONAL INFORMATION i.
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V0GTLE ELECTRICAL GENERATING PLANT UNITS 1 and 2 f
MATERIALS APPLICATION SECTION MATERIALS ENGINEERING BRANCH 251.1 To demonstrate compliance with Appendices G and H, 10 CFR Part 50, s
whic.h were revised in the Federal Register on May 27, 1983 and became effective on July 26, 1983:
l Report the complete Charpy V notch, impact test data versus temperature a.
and curves f'or each beltline material.
The data should include the
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energy absorbed and lateral expansion versus test temperature for j
each material tested.
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b.
Identify the withdrawal schedule for the surveillance capsules and specimens.
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Submit a fracture mechanics analysis or pressure-temperature limit curves which comply with the reactor pressure vessel closure flange region requirements of Appendix G, 10 CFR Part 50.
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