ML20213D990
| ML20213D990 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 02/17/1982 |
| From: | Johnston W Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0465, CON-WNP-465 NUDOCS 8203010358 | |
| Download: ML20213D990 (12) | |
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Docket No. 50-937,
Q-wr-y Robert L. Tedesco, Assistant Director for LicI$j'ni--
MEMORANDUM FOR:
Division of Licensing FROM:
William V. Johnston, Assistant Director for Materials and Qualification Engineering, Division of Engineering
SUBJECT:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM NUCLEAR PLANT UNIT 2 Plant Name: WPPSS Nuclear Plant Unit 2 Suppliers: General Electric; Burns & Roe, Inc.
Docket Number:
50-397 Responsible Branch and Project Manager:
LB-2, R. Auluck Reviewer:
B. J. Elliot, INEL Description of Task:
Safety Evaluation Report Input for Sections 5.3.1, 5.3.2, and 5.3.3 Review Status:
Complete The Component Integrity Section, Materials Engineering Branch, Division of Engineering has reviewed the Final Safety Analysis Report (FSAR) for WPPSS Nuclear Plant Unit 2.
Based on our review of the information in FSAR amend-ments through number 20 dated November 1981 and letter from G. O. Bouchey to A. Schwencer dated December 18, 1981, we have prepared our input to Safety Evaluation Report (SER) Sections 5.3.1, 5.3.2, and 5.3.3 which is attached.
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William V. Johnston, Assistant Director for Materials and Qualification Engineering Division of Engineering
Attachment:
As stated cc:
D. G. Eisenhut R. Auluck l
R. H. Vollmer W. S. Hazelton W.-4/. Johnston R. W. Klecker A. Schwencer B. J. Elliot E. Sullivan P. K. Nagata (INEL)
Contact:
B. J. Elliot Distribui.i n MTEB Reading File
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e ATTACHMENT 1 Washington Public Power Supply System WPPSS Nuclear Plant Unit 2 Docket No. 50-397 MATERIALS ENGINEERING BRANCH COMPONENT INTEGRITY SECTION 5.3.1 Reactor Vessel Materials The staff of EG&G, Idaho National Engineering Laboratory has reviewed the frac-ture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, and the materials surveillance program for the reactor vessel belt-line.
The acceptance criteria and references which are the basis for this evaluation are set forth in Paragraph II.3.a of Standard Review Plan (SRP) Sec-tion 5.2.3 and 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 discus-sion of this review follows.
General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," Appendix A,10 CFR Part 50, requires that the reactor coolant pres-sure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and test conditions, the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is mini-mized. General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary " Appendix A,10 CFR Part 50, requires, in part, that the reactor coolant presst're boundary be designed to permit an appropriate material sur-veillance program for the reactor pressure boundary.
The Construction Permit for WPPSS Nuclear Plant Unit 2 (hereafter WNP-2) was issued on March 19, 1973. The Edition and Addenda of the ASME Code applicable t.o the design and fabrication of any reactor vessel is specified in Sec-tion 50.55a of 10 CFR Part 50.
Based on the Construction Permit date, this WNP-2 SER 5-1
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section of the Code of Federal Regulations requires that the WNP-2 reactor vessels meet the requirements of at least the 1971 Edition of the ASME Code, Summer 1971 Addenda.
The WNP-2 FSAR states that the reactor vessels were designed, fabricated, tested, inspected, and stamped according to the 1971 ASME Code.
Therefore, the applicant did not comply with the explicit requirements of Paragraph 50.55a(c)(2), 10 CFR Part 50.
However, we will evaluate the applicant's RCPB materials to Appendices G and H, 10 CFR Part 50, which will ensure that material properties are equivalent to those specified in Sec-tion 50.55a, 10 CFR Part 50.
Appendix G, " Fracture Toughness Requirements," Appendix H, " Reactor Vessel Material Surveillance Requirements," of 10 CFR part 50, specify the fracture toughness requirements for the ferritic materials of the reactor coolant pres-sure boundary.
Evaluation of Compliance to Appendix G,10 CFR Part 50 Based on our review of the applicant's submittal that describes the extent of compliance of WNP-2 to Appendix G, 10 CFR Part 50, we have determined that the requirements of Appendix G have been met except for Paragraphs III.B.1, III.B.3, III.B.4, III.C.2, and IV. A.3 of Appendix G.
Paragraph III.B.1 requires that the Charpy V-notch (CVN) impact tests and the drop-weight tests be conducted in accordance with Paragraph N8-2322 of the ASME Code.
Paragraph NB-2322 requires, in part, that CVN tests specimens which represent the pressure vessel base metal, shall be taken perpendicular to the principal rolling direction (transverse direction).
However, in accordance with the earlier ASME Code requirements to which the WNP-2 pressure vessel was built, the CVN tests conducted for the WNP-2 pressure vessel base metal were performed using longitudinally oriented specimens.
To evaluate compliance with the specimen orientation requirements of Appendix G, we have evaluated the data obtained for the WNP-2 vessel base metal and addi-tional data obtained for similar reactor vessel steels having both longitudi-nal and transverse specimen orientations.
The additional data are contained in WRC Bulletin 217 and Electric Power Research Institute EPRI Report NP-933, l
WNP-2 SER 5-2
December 1978. Based on our review and evaluation of these data, we conclude thet adequate correlations can be used to translate the data obtained using longitudinally oriented specimens to an equivalent transverse CVN impact energy for comparison with the requirements of Appendix G.
The application of these correlations between longitudinal and transverse specimen orientations demon-strate that the longitudinal specimens for the WNP-2 vessel base metal meet the transverse requirements of Appendix G and consequently, we conclude that exemp-tions to Paragraph III.8.1 is justified.
The specific orientation correlation factors that are used to demonstrate compliance with the acceptance criteria for material fracture toughness are described in later paragraphs that evaluate the specific fracture toughness requirements.
Paragraph III.B.3 requires that the temperature instruments and Charpy test machines be calibrated in accordance with Paragraph NB-2360 of Section III of the ASME Code.
Verification of this required' calibration was impossible since the testing organization only retained the calibration report until the next calibration.
However, General Electric has stated that the test instruments and machines were routinely calibrated on a periodic basis.
Based on the standard practice of this period and on past experience with Charpy testing, we conclude that it is very unlikely that the test instruments and machines were not adequately calibrated and that an exemption to the requirement for maintaining the calibration report is justified.
Paragraph III.B.4 requires that the testing personnel shall be qualified by training and experience and should be able to perform the tests in accordance with written procedures.
For WNP-2 component testing, no written procedures were in existence as required by the later regulation.
However, the individuals were qualified by on-the-job training and past experience.
Because these tests were reTatively routine in nature and are continually being performed in the laboratory that conducted these tests, it is unlikely that the tests were con-ducted improperly.
Consequently, we conclude that an exemption for not perform-ing the tests in accordance with written procedures is justified.
Paragraph III.C.2 of Appendix G,10 CFR Part 50, requires, in part, that the base materials and weld materials used to prepare test specimens for the reactor vessel beltline region shall be from excess material from the vessel beltline WNP-2 SER 5-3
region.
The applicant has produced data for the reactor beltline region base materials and so has complied with the requirements of Paragraph III.C.2 with respect to base materials.
However, WNP-2 weld test specimens were taken from simulated weldments prepared from excess production plate which were not from the beltline region.
The weld wire and flux materials used in the test speci-mens are the same as those used in the reactor vessel beltline.
Since the weld toughness properties are determined primarily by weld wire, flux, welding proc-ess, and heat treatment, and not by base material, the use of weldment test specimens having the same weld wire, flux, welding process, and heat treatment as the beltline welds is sufficient to satisfy the intent of Paragraph III.C.2 and provides acceptable justification for an exemption to the requirements of Paragraph III.C.2.
Paragraph IV. A.3 of Appendix G requires, in part, that materials for piping, pumps, and valves meet the requirements of Paragraph NB-2332 of the ASME Code.
According to WNP-2 FSAR, the MSIVs (main steam isolation valves) were not tested because the ASME Code existing at the' time of the purchase order, April 1971, did not require brittle fracture testing on ferritic pressure boundary components when the system temperature was in excess of 250 F at 20%
of the design pressure.
However, the Edition of %e ASME Code to which the MSIVs should have been bought is the 1971 Edition, Summer 1971 Addenda.
This Code requires CVN impact, testing of the MSIV materials at WNP-2.
The applicant has complied with the requirements of Paragraph IV.A.3 except-that the main steam isolation valve (MSIV) discs, covers, and bodies were not CVN impart tested.
However, the applicant has provided CVN impact data from the literature and other nuclear plants on material equivalent to the WNP-2 MSIV discs, covers, and bodies that demonstrate that if the WNP-2 material had been tested, they would have met the requirements of Paragraph IV.A.3.
There-fore, on the above basis, we consider granting an exemption from the explicit requirements of Paragraph IV. A.3 of Appendix G,10 CFR Part 50 for the MSIV discs, covers, and valves is justified.
WNP-2 SER 5-4
Evaluation of Compliance to Appendix H, 10 CFR Part 50 Based on our review of the applicant's submittal that detailed the extent of compliance of WNP-2 with Appendix H, 10 CFR Part 50, we have determined that the requirements of Appendix H have been met except for Paragraph II.B.
Paragraph II.B requires, in part, that the surveillance program for the fer-ritic materials in the reactor vessel beltline comply with ASTM E 185-73,
" Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessel." ASTM E 185-73 defines the type, number, and selection criteria for the reactor vessel irradiation surveillance program.
The applicant is not in exact compliance with two requirements of Paragraph II.B of Appendix H.
These requirements are that the Charpy specimens from base material must b'e oriented in the transverse direction, and that the surveillance program materials be selected per ASTM E 185-73.
ASTM E 185-73 requires the surveillance materials be the most limiting mate-rials, i.e., t. hose having the highest end of-life (E0L) adjusted RT The NDT.
most limiting material in the WNP-2 beltline area is C1272-1.
The WNP-2 sur-veillance materials are 85301-1 and 3P4966/1214 for the plates and welds, respectively.
Therefore, the WNP-2 surveillance program does not conform to the requirements of Paragraph II.B of Appendix H,10 CFR Part 50.
However, the data from the WNP-2 survaillance specimens will be treated in the following manner so that the CVN impact specimen orientation and limiting reactor vessel material are properly evaluated.
(1) The CVN data from the longitudinal plate specimens will be adjusted using Branch Technical Position - MTEB No. 5-2 to provide an adequate estimate for the transverse limiting material.
These adjustments will be made to the upper shelf CVN impact energies and the temperature at which a specific CVN impact energy is obtained.
(2) The adjustment in RT f r the limiting material (plate: C1272-1, NDT initial RTNDT:28 F, %Cu:.15, %P:.013) will be calculated from Regalatory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage WNP-2 SER 5-5
to Reactor Vessel Materials." This value will be compared with the adjusted RT from the plate material (B5301-1) in the surveillance NDT capsule to determine the predicted shift of the limiting plate in the reactor vessel beltline.
Although the WNP-2 surveillance program does not strictly conform to the explicit requirements of Paragraph II.B of Appendix H,10 CFR Part 50, the use of the present WNP-2 surveillance materials, Branch Technical Position - MTEB No. 5-2, and Regulatory Guide 1.99 will allow us to conservatively estimate the adjusted RT f r the limiting beltline materials.
Based on the above, NDT an exemption from the explicit requirements of Paragraph II.B of, Appendix H, 10 C'FR Part 50 is justified.
Conclusions Our technical evaluation has not identified any practical methods by which the existing WNP-2 reactor vessel can comply with the specific requirements of Paragraphs II.B.1, III.B.3, III.B.4, III.C.2, and IV. A.3 of Appendix G and Paragraph II.B of Appendix H,10 CFR Part 50. Alternate methods justify an exemption for Paragraphs III.B.1, III.B.3, III.B.4 and III.C.2 of Appendix G; and Paragraph II.B of Appendix H.
Based on the foregoing, pursuant to 10 CFR, Section 50.12, exemptions from the specific requirements of Appendices G and H of 10 CFR Part 50, as discussed above, are authorized by law and can be granted without endangering life or property or the common defense and security and are otherwise in the public interest.
We conclude that the public is served by not imposing certain oro-visions of Appendices G and H of 10 CFR Part 50 that have been determined to be either TTnpractical or would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
l Furthermore, we have determined that the granting of these exemptions does not l
authorize a change in effluent types or total amounts nor an increase in power j
level and will not result in any significant environmental impact. We have c.oncluded that these exemptions would be significant from the standpoint of environmental impact and pursuant to 10 CFR 51.5(d)(4) that an environmental WNP-2 SER 5-6
impact statement, or negative declaration and environment appraisals, need not be granted in connection with this action.
Appendix G, " Protection Against Nonductile Failure,"Section III of the ASME Boiler and Pressuce Vessel Code, will be used, together with the fracture toughness test results required by Appendices G and H, 10 CFR Part 50, to calculate the reactor coolant pressure boundary pressure-temperature limita-tions for WNP-2.
The fracture toughness tests required by the ASME Code and the Appendix G of 10 CFR Part 50 will provide reasonable assurance that adequate safety margins against the possibility of nonductile 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 toughness tests performed in accordance with the ASME Code and NRC regulations, will provide adequate safety margins during operating, testing, maintenance, and anticipated transient conditions.
Compliance with these Code provisions and NRC r'egulations constitutes an acceptable basis for satisfying the fracture toughness requirements of General Design Criterion 31.
The materials surveillance program, required by Appendix H, 10 CFR Part 50, will provide information on material properties and the effects of irradiation on material properties so that changes in fracture toughness of material in WNP-2 reactor vessel beltline caused by exposure to neutron radiation can be properly assessed, and adequate safety margins against the possibility of vessel failure can be provided.
Complian"ce with ASTM E 185-73 and Appendix H,10 CFR Part 50, assures that the surveillance program constitutes an acceptable basis for monitoring radiation-induced changes in the fracture toughness of the reactor vessel material and satisfies the material surveillance requirements of General Design Criteria 31 and 32.
WNP-2 SER 5-7
5.3.2 Pressure-Temperature Limits The staff of EG&G, Idaho National Engineering Laboratory has reviewed the applicant's pressure-temperature limits for operation of WNP-2 reactor vessels.
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 NUREG-0800 Rev. 1 dated July 1981. A discussion of this review follows.
Appendix G, " Fracture Toughness Requirements," and Appendix H, " Reactor Vessel Material Surveillance Program Requirements," 10 CFR Part 50, describe the con-ditions that require pressure-temperature limits for the reactor coolant pre's-
'sure boundary and provide the general bases of these limits.
These appendices specifically require that pressure-temperature limits must provide safety mar-gins for the reactor coolant pressure boundary at least as great as the safety margins recommended in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G, " Protection Against Nonductile Failure." Appendix G, 10 CFR Part 50, requires additional safety margins whenever the reactor core is critical, except for low-level physics tests.
The following pressure-temperature limits imposed on the reactor coolant pres-sure boundary during operation and tests are reviewed to ensure that they pro-vide adequate safety margins against nonductile behavior or rapidly propagating failure of ferritic components as required by General Design Criterion 31:
(1) Preservice hydrostatic tests, (2) Inservice leak and hydrostatic tests, (3) lleatup and cooldown operations, and (4) Core operation.
Appendix G of the ASME Code specifies the procedures that are to be used to construct the pressure-temperature limits for the ferritic components in the r,eactor coolant pressure boundary. These procedures include definition of the WNP-2 SER 5-8
initial reference temperature, RTNOT, f r the ferritic materials and considera-tion of the change in initial RT due to neutron irradiation.
NOT We have reviewed the WNP-2 pressure-temperature limits for initial operation as found in Figure 5.3-4 of the applicant's FSAR.
Based on methodology in Regula-tory Guide 1.99, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," and Standard Review Plan Section 5.3.2, " Pressure-Teverature Limits," we have found that the pressure-temperature limits for initial operation are acceptable for 3 effective full power years (EFPY).
Therefore, these pressure-temperature curves may be used for only 3 EFPY. But, at the refueling cycle prior to 3 EFPY, the applicant must submit, as part of their technical specifications, curves which are acceptable for at least one additional fuel cycle.
These curves must be based on an adjusted reference temperature for the limiting material, plate C1272-1 (initial RTNOT: +28 F,
%Cu:0.15 and %P:.013) which is calculated in accordance with Regulatory Guide 1.99.
When the surveillance material is tested in approximately 8 EFPY, the pressure-temperature limits must be calculated based on the method identi-fied in SER Section 5.3.1.
The pressure-temperature limits to be imposed on the reactor coolant system for all operating and testing conditions to ensure adequate safety margins against nonductile or rapidly propagating failure must be in conformance with established criteria, codes, and standards acceptable to the staff.
The use of operating limits based on these criteria, as defined by applicable regula-tions, codes, and standards, will provide reasonable assurance that nonductile or rapidly propagating failure will not occur and will constitute an accept-able basis for satisfying the applicable requirements of General Design Criterien 31.
5.3.3 Reactor Vessel Integrity We have reviewed the FSAR sections related to the reactor vessel integrity of WNP-2.
Although most areas are reviewed separately,' reactor vessel integrity is of such importance that a special summary review of all factors relating to reactor vessel integrity is warranted.
WNP-2 SER 5-9
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The staff of EG&G, Idaho National Engineering Laboratory has reviewed the frac-ture toughness of ferritic reactor vessel and reactor coolant pressure boundary materials, the pressure temperature limits for operation of the reactor vessels, 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, 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.
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 integ-rity.
The applicant has complied with Appendices G and H, 10 CFR Part 50, except for the following items:
(1) Paragraph III.B.1, Appendix G:
The applicant did not take the CVN speci-mens representing the reactor pressure vessel perpendicular to the major rolling direction.
(2) Paragraph III.B.3, Appendix G: The temperature instruments and the CVN test machines were not calibrated in accordance with Paragraph NB-2360 of Section III of the ASME Code.
(3) Paragraph III.B.4, Appendix G: The testing personnel were not tested in accordance with written procedures as required.
(4) Paragraph III.C.2, Appendix G:
The WNP-2 weld material test specimens were not taken from excess beltline material.
(5) Paregraph IV.A.3, Appendix G:
The WNP-2 MSIV discs, covers, and bodies were not tested per Paragraph NB-2330 of the ASME Code.
(6) Paragraph II.8, Appendix H:
The most limiting beltline material per ASTM E 185-73 has not been included in the WNP-2 surveillance program.
- Also, the plate surveillance CVN specimens were taken from the longitudinal direction and not the transverse direction as required by ASTM E 185-73.
WNP-2 SER 5-10
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A We have granted exemptions to the explicit requirements of Paragraphs III.B.1, III.B.3, III.B.4, III.C.2, and IV. A.3 of Appendix G and Paragraph II.8 of Appendix H,10 CFR Part 50. We have reviewed all factors contributing to the structural integrity of the reactor vessel and conclude there were no special I
considerations that make it necessary to consider potential reactor vessel j
failure for WNP-2.
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WNP-2 SER 5-11
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