ML20213D791
| ML20213D791 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 08/17/1981 |
| From: | Johnston W Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0378, CON-WNP-378 NUDOCS 8109040021 | |
| Download: ML20213D791 (17) | |
Text
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AUG 171981 Docket No. 50-397 MEMORANDUM -FOR:
Robert L. Tedesco, Assistant Director for Licensing Division of Licensing FROM:
William V. Johnston, Assistant Director Materials & Qualifications Engineering Division of Engineering
SUBJECT:
WASHINGTON PUBLIC POWER SUPPLY SYSTEM, WPPSS NUC gg PROJECT NO. 2 g
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Plant Name: WPPSS Nuclear Project No. 2 f
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Suppliers:
General Electric, Burns and Roe, Inc.
Docket Number:
50-397 V oG ff -
-4 Responsible Branch and Project Manager:
LB-2, R. Auluck
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p Reviewer:
B. J. Elliot, INEL Description of Task:
Safety Evaluation Report Input j.
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's Review Status: Additional Information Required The Component Integrity Section, Materials Engineering Branch, Division of Engineering, has reviewed the Final Safety Analysis Report for WPPSS Nuclear Project No. 2.
Based on our review of this information, we have prepared our input to the Safety Evaluation Report (Attachment 1).
In this Safety Evalua-tion Report we have granted exemptions to some paragraphs of Appendices G and H and have identified areas for which sufficient information has not been submitted to determine compliance with or justify an exemption to Appendices G and H, 10 CFR Part 50.
The areas where sufficient information has not been provided will remain open items until Washington Public Power Supply System provides the necessary information.
The specific information required to resolve these open itsms is contained in i and is summarized for convenience in Attachment 2.
The requested information must be received by November 1, 1981 in order to meet the SER schedule dates.
I William V. Johnt. ton l
Materials & Qualifications Engineering Division of Engineering l
Enclosures:
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Contact:
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Robert L. Tedesco 2
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D. G. Eisenhut R. H. Vollmer A. Schwencer S. S. Pawlicki G. Johnson W. S. Hazelton R. Auluck P. K. Nagata (INEL)
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ATTACHMENT 1 Washington Public Power Supply System WPPSS Nuclear Plant Unit 2 Docket No. 50-97 MATERIALS ENGINEERING BRANCH COMPONENT INTEGRITY SECTION
' 5. 3.1 Reactor Vessel Materials General Design Criterion 31, " Fracture Prevention of Reactor Coolant Pressure Boundary," Appendix A, 10 CFR Part 50, requires that the reactor coolant pressure boundary be designed with sufficient margin to assure that when stressed under operating, maintenance, and testing conditions the boundary behaves in a nonbrittle manner and the probability of rapidly propagating fracture is minimized.
General Design Criterion 32, " Inspection of Reactor Coolant Pressure Boundary," Appendix A, 10 CFR Part 50, requires, in part, that the reactor coolant pressure boundary be designed to permit an appropriate material surveillance 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 to the design and fabrication of any reactor vessel is specified in l
Section 50.55a of 10 CFR Part 50.
Based on the Construction Permit date, this 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, l
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 i
i 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 i
which will ensure that material properties are equivalent to these specified in Section 50.55a, 10 CFR Part 50.
5-1
Appendix G, " Fracture Toughness Requirements," Appendix H, " Reactor Vessel Material Surveillance Requiiements," of 10 CFR Part 50, specify the fracture toughness requirements for the ferritic materials of the reactor coolant pressure boundary.
1 EVALUATION OF COMPLIANCE TO APPENDIX G 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.1, III.C.2, IV.A.2.a, IV.A.3, and IV.B. of Appendix G.
Noncompliance to Paragraphs III.C.2 and IV.A.2.a was not identified by the applicant. Our evaluation of the paragraphs in Appendix G, 10 CFR Part 50, in which the applicant has not complied, follows.
Paragraph III.B.1 requires that the Cha'rpy V notch (CVN) impact tests and the drop-weight tests be conducted in accordance with Paragraph NB-2322 of the ASME Code.
Paragraph NV-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 l
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 WFP-2 vessel base metal and additional data obtained for similar reactor vessel steels having both longitudinal and transverse specimen orientations. The additional data are contained in WRC Bulletin 217 and Electric Power Research Institute EPRI Report NP-933, December 1978.
Based on our review and evaluation of these data, we conclude that adequate correlations can be used to translate the data obtained using longitudinally oriented specimens to an equivalent transverse l
CVN impact energy for comparison with the requirements of Appendix G.
The application of these correlations between longitudinal and transverse specimen l
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orientations demonstrate that the longitudinal specimens for the WNP-2 vessel base metal meet the transverse requirements of Appendix G and consequently,'we conclude that exemptions to Paragraph III.B.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 mac'hines 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 indi-viduals were qualified by on-the-job training and past experience.
Because these tests are relatively routine in nature and are continually being performed in the laboratory that conducted these tests, it is unlikely that the tests were conducted improperly.
Consequently, we conclude that an exemption for not performing the tests in accordance with written procedures is justified.
Paragraph III.C.1 requires, in part, that reactor vessel beltline material be, CVN impact tested over an appropriate temperature range sufficient to define the CVN impact test curves, including the upper shelf levels in terms of both fracture energy and lateral expansion.
WNP-2 FSAR shows data at only one temperature which is what was required by the 1971 Code, Summer 1971 Addenda.
Until the applicant supplies CVN impact data to define the upper shelf levels 5-3
for all reactor vessel beltline materials, Paragraph III.C.1 of Appendix G, 10 CFR Part 50 will remain an open item.
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 region.
The applicant has produced data for the reactor beltline region base materials and so has complied with the requirements of Para-graph 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 specimens are the same as those used in the reactor vessel beltline.
Since the weld toughness properties are determined primarily by weld wire, flux, welding process, and heat treatment, and not by differ-ences in similar base materials, 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 requirements of Paragraph III.C.2 and provides acceptable justification for an exemption to the requirements of Paragraph III.C.2.
Paragraph IV.A.2.a requires that a reference temperature, RTNDT, be determined for each ferritic material of the reactor coolant pressure boundary and that this reference temperature be used as a basis for providing adequate margins of safety for reactor operation. The value of RT is defined in the ASME NDT Code as the higher of either (a) the nil ductility temperature, as determined I
by the dropweight test, or (b) a temperature of 60 F less than the temperature at which 50 ft-lb energy and 35 mils lateral expansion is achieved, as determined by the CVN impact test.
The CVN impact test for base metals is to be conducted using specimens oriented in the transverse direction.
l WNP-2 FSAR gives RT data for the reactor beltline base and weld materials.
NDT However, no RT data are given for ferritic reactor coolant pressure NDT boundary base and weld metals outside the beltline region.
In order to demonstrate compliance with Paragraph IV.A.2.a, the applicant must report the RT f r all ferritic plates, pipes, forgings and welds outside the beltline NDT 5-4
region which will be limiting for operation of the reactor vessel.
If the method for determining the RT f r welds and base metals outside the NDT beltline region is different than that required by Paragraph IV.A.2.a, the applicant must identify the method and provide technical justification for its use.
For beltline base metal the RT was determined in accordance with Paragraph NDT IV.A.2.a except that the Charpy V-notch specimens were located in the longitudinal rather than transverse direction.
To compensate for specimen orientation, the temperatures at which the 50 ft-lb energy levels-would have been achieved for transverse specimens were estimated at 30 F higher than the temperature indicated from the longitudinal data.
For the beltline weld metal, the value of RT was not in strict compliance NDT with Paragraph IV. A.2.a because the nil ductility temperature was not ' defined explicitly for all the beltline welds and the CVN specimens were tested only at a single temperature, which, in general, was not sufficient to define the 50 ft-lb energy level.
To define RT f r the weld metal in accordance with NOT Paragraph IV.A.2.a, the applicant employed the following alternative procedures:
b (1) The nil ductility temperature (NDT) was assumed to be -50 y, (2) If the CVN impact energy obtained at the single test temperature was less than 50 ft-lb, then the temperature at which 50 ft-lb energy would be 1
achieved was estimated from the available data by using a temperature-l impact energy correlation of 2 F per ft-lb to extrapolate the energy i
level obtained at the test temperature to the temperature corresponding to the 50 ft-lb energy level.
(3) If the CVN impact energy obtained at the single test temperature was greater than 50 ft-lb, then the test temperature itself was used at the 50 ft-lb temperature.
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We have reviewed the data in WNP-2 FSAR, WRC Bulletin 217, Electric Power Research Institute Reports,'EPRI NP-121, Volume II (April 1976) and EPRI NP-933 (December 1978).
Our review of these data indicate that the correlations used by the applicant to determine the effect of specimen orientation and the temperature at which 50 ft-lb would be achieved is conservative and results in values of RT f r the plate and weld materials that are equivalent to those NDT that would be determined if the tests were conducted in strict compliance to Appendix G.
However, while an assumed NDT of -50 F in Section 5.3.1.5.2.2 in WNP-2 FSAR appears conservative for the weld metal, the applicant has not provided sufficient justification to confirm this value.
The data submitted to support this value did not identify the thermal treatment used in preparing the welds that exhibited the -50 F NDT.
To confirm that this value is conservative, the applicant must identify the thermal treatment used in fabricating the welds that exhibited a -50 F NDT and show that these are representative of the welds in the WNP-2 reactor vessel beltline.
Until this information is supplied we cannot complete our evaluation of compliance with Paragraph IV.A.2.a.
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 the ASME Code to which the MSIVs should have been bought is the 1971 Edition, Summer 1971 Addenda.
This Code requires testing of valve materials which are greater than 0.5 in, thick and have inlet connections greater than 6 inches. The applicant has provided no data for MSIV materials, and the WNP-2 FSAR does not indicate whether the MSIVs fall into either of the above categories. To determine compliance with Para-graph IV. A.3 of Appendix G,10 CFR Part 50, the applicant must indicate whether the MSIVs are exempt from CVN impact testing using the 1971 ASME Code, 5-6
s Summer 1971 Addenda.
If the MSIVs are not exempt from CVN impact testing, the applicant must supply data to show they conform to the Edition and Addenda of the ASME Code mentioned above.
Any data submitted shall include the following:
(1) ASME material designation (e.g., SA-216 Grade WCB),
(2) Heat treatment history, (3) Heat or lot number, (4) Energy absorbed (ft-lb), and (5) Test temperature.
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If the above-mentioned data does not exist, then the applicant must demonstrate the adequacy of the MSIV materials by an alternate method.
This method can include data from the literature, further tests, and/or analyses.
Any data submitted shall include the above-mentioned information.
Until the applicant demonstrates the adequacy of the MSIV materials, we cannot complete our evaluation of compliance with Paragraph IV.A.3.
Paragraph IV.B of Appendix G requires that the reactor vessel beltline materials have a minimum upper-shelf energy, as determined from Charpy V-notch impact tests on unirradiated specimens in accordance with Paragraph NB-2322.2(a) of the ASME Code, of 75 ft-lb, unless it can be demonstrated to the Commission by appropriate data and analyses that lower values of upper-shelf energy still provide adequate margin for deterioration from irradiation.
In accordance with 10 CFR 50.55a, the fracture toughness tests were conducted to an ASME Code Edition that preceded the effective date of Appendix G to 10 CFR Part 50.
This Edition of the ASME Code did not require that the upper-shelf energy be established but only required that the tests be conducted at a single temperature equal to 60 F below the lowest service temperature. The test temperature determined in this manner typically was 10 F.
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The reactor beltline materials are listed in the response to Question 121.1.
All the materials listed there plates and welds - do not meet the 75 ft-lh minimum CVN impact test energy with the exception of Girthweld AB/E8018NM/
492L4871/ Lot A422827AF-(location / type / heat / lot).
For the other plates and welds to meet the upper-shelf requirement of Paragraph IV.B of Appendix G, the applicant would either have to show that they possess, for the welds, 75 ft-lb, or, for the plates 115 ft-lb or 75 ft-lb in the longitudinal and transverse directions, respectively, of CVN impact energy at some test temperature; or by using Regulatory Guide 1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials," show that a lower value of initial upper-shelf energy will still provide an end-of-life upper-shelf energy of 50 ft-lb at the quarter thickness vessel wall location (77 ft-lb for the longitudinal direction in the plates).
To provide an acceptable basis for an exemption from the requirements of Paragraph IV.B, the applicant must provide additional cata, information from the literature, and/or analyses to demonstrate that a margin of safety equivalent to that of the requirement has been attained.
The additional plate data may be from tests of similar plate material; i.e., plates of similar steelmaking practice, with a similar thermomechanical history.
The additional weld data may be from tests of similar welds; that is, welds made by using the same weld wire, flux, welding process, and heat treatments as those used in the weld metals.
If test data from similar welds are used, they must include the following:
(1) Heat treatment of test weld (2) Chemical composition of test weld (3) Type / Heat / Lot Numbers (4) Test Temperature (5) Energy absorbed (ft-lb) 5-8
i 2 EVALUATION OF COMPLIANCE OF APPENDIX H 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 ferritic 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 must be oriented in the transverse direction, and that the surveillance program materials be identified per ASTM E 185-73.
The WNP-2 surveillance program does not comply with Paragraph II.B of Appendix H, 10 CFR Part 50 in that the following data, required by ASTM E 185-73, are not given.
(1) Actual surveillance material specification; (2) Origin of each surveillance specimen (base metal:
heat number, plate identification number; weld metal: weld wire, heat of filler material, production welding conditions, and plate material used to make weld specimens);
l (3) Test specimen, and type; (4) Heat treatment history of each test specimen; (5) Chemical composition of each test specimen.
To demonstrate compliance with Paragraphs II.B of Appendix H, 10 CFR Part 50, the applicant shall supply the data detailed above.
Unless the applicant supplies these data, we cannot evaluate WNP-2's compliance to Appendix H.
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i 3 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 III.B.1, III.B.3, III.B.4, III.C.1, III.C.2, IV.A.2.a, IV.A.3, and IV.B 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, and III.B.4 of Appendix G.
Paragraphs III.B.1, III.C.1, III.C.2, IV.A.2.a, IV.A.3 and IV.B of Appendix G and Paragraph II.B of Appendix H will remain open items until the applicant submits data to demonstrate compliance.
Based on the foregoirg, 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 provisions of Appendices G and H of 10 CFR Part 50 that have been determined to be either impractical or would result in hardship or unusual difficulties without a compensating increase in the level of quality and safety.
Furthermore, we have determined that the granting of these exemptions does not 4
authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact.
We have concluded that these exemptions would be insignificant from the standpoint of environmental impact and pursuant to 10 CFR 51.5 (d)(4) that an environmental impact statement, or negative declaration and environments appraisals, need not be granted in connection with this action.
5.3.2 Pressure-Temperature Limits 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 for the reactor coolant pressure boundary and provide the general bases for these limits.
These appendices specifically require that pressure-temperature limits must provide safety margins for the reactor coolant pressure boundary at least as great as 5-10
T the safety margins recommended in the ASME Boiler end 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 pressure boundary during operation and tests are reviewed to ensure that they provide 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) Heatup 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 reactor coolant pressure boundary.
These procedures include definition of the initial reference temperature, RTNDT, f r the ferritic materials and con-sideration of the change in initial RT due to neutron irradiation.
As NDT Section 5.3.1 of this safety evaluation discusses, the applicant has provided acceptable methods to define initial RTNDT, except for the weld material where additional information is required to justify a nil ductility temperature of
-50 F.
Until the applicant has supplied this information, we cannot complete our evaluation of the pressure-temperature limits.
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 regulations, codes, and standards, will provide reasonable assurance that 5-11
nonductile or rapidly propagating failure will not occur, and will constitute an acceptable basis for satisfying the applicable requirements of General Design Criterion 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.
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), and (4) Operating conditions (SER 5.3.2).
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 and H, 10 CFR Part 50, except for the following items:
(1) Paragraph IV.A.2.a. Appendix G:
The applicant has not provided sufficient information to define the reference temperature, RTNDT' I "
l all ferritic pressure-retaining materials in the reactor coolant pressure boundary.
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(2) Paragraph IV.A.3, Appendix G:
The applicant has not provided sufficient data to determine if the main steam isolation valves, meet the requirements of Appendix G.
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(3) Paragraphs III.C.1, and IV.B Appendix G:
The CVN impact test d&ta for all base and weld mate ~ rials in the core beltline region except welds E8018NM/492L4871/ Lot A422827AF, RAC01NM/SP6756/ Lot 0342, and E8018/04T931/ Lot A423827AG are not sufficient to demonstrate that the minimum initial upper-shelf requirement of 75 ft-lb has been met.
(4) Paragraph II.B, Appendix H:
The surveillance capsule identification data per ASTM E 185-73 have not been included in the WNP-2 FSAR.
Until the applicant has supplied the information necessary to complete our evaluation of compliance with Appendices G and H, 10 CFR Part 50, we cannot complete our evaluation of the structural integrity of the reactor vessel of WNP-2.
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ATTACHMENT 2 Request for Additional Information 251.0 Materials Engineering Branch - Component Integrity Section 251.1 Provide data to justify that the value of -50 F used to estimate the nil ductility temperature for the beltline welds was obtained from test samples that represent the beltline welds in the WNP-2 reactor pressure vessel.
This information should include a comparison of the significant weld parameters (e.g., weld wire, flux, thermal treatment) and mechanical properties from both the sample and beltline welds.
251.2 In order to demonstrate that all reactor coolant pressure boundary (RCPB) materials comply with the requirements of Paragraph IV.A.2.a, the applicant must report the RT f r all ferritic (RCPB) plates, NOT pipes, forgings and welds outside the reactor vessel beltline region which will be limiting for operation of the reactor vessel.
If the method for determining RT f r these materials is different than NDT that required by Paragraph IV.A.2.a, the applicant must identify the method and provide technical justification for its use.
251.3 Provide CVN impact data to demonstrate that all the reactor beltline materials except Girthweld A8/E8018NM/492L4871/ Lot A422827AF1, Girthweld A8/RAC01NM/5P6756/ Lot 0342, and Girthweld AB/E8018/04T931/
l Lot A423827AG have a minimum CVN impact test upper shelf energy of 75 ft-lb as required by Paragraph IV.B of Appendix G.
The applicant j
must provide additional data, information from the literature, and/or analyses to demonstrate that the materials' upper-shelf will be assured for normal operation. The additional data should be from tests of plates having similar steelmaking practices and thermo-I mechanical history.
For the welds, the additional data should be from similar welds, i.e., those having the same weld wire, flux, and i
1 5-16
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thermal treatments as the beltline welds and prepared by the same
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fabricator as the original beltline welds.
251.4 To demonstrate compliance with Appendix H, 10 CFR Part 50, provide a table that includes the following information for each surveillance material specimen:
(1) Actual material specification; (2) Origin of each surveillance specimen (base metal:
heat number, plate identification number; weld metal: weld wire, heat of filler material, production welding process, and plate material used to make weld specimens);
(3) Test specimen, and type; (4) Heat treatment of each test specimen; (5) Chemical composition of each test specimen.
Provide the asmuthal location, lead factor and withdrawal time for each specimen capsule.
251.5 Provide the following data to demonstrate the main steam isolation valve (MSIV) RCPB materials will meet the fracture toughness requirements of NB-2332 of the Summer 1972 Addenda to the 1971 ASME Code.
(a) Indicate the diameter and nominal wall thickness of the connecting pipes.
(b)
Indicate the material specification, material type, material supplier, heat treatment reccived by the material, minimum design wall thickness and CVN impact test results for all ferritic RCPB materials used in the MSIV bodies, covers, discs, stems and bolting.
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(c)
If CVN impact data can not be provided from each of these materials, provide CVN impact data from other materials whichr were fabricated from the same material specification and type, by same material' supplier and heat treated to an equivalent metallographic condition as the MSIV materials.
251.6 There are six weld heats in response to Q 121.2 and 121.1 whose composition cannot be found in the list of weld metal compositions.
They are (Location / Type / Heat / Lot):
(a) Ring 21BA/RAC01NMM/3P49U/1214 or 3PA966/1214*
(b) Ring 21BB/E8018NM/C3L46C/J020827A (c) Ring 21BC/E8018NM/19L853/A111A27A (d) Ring 21BC/E8018NM/C3L46C/J020A27A (e) Ring 21BD/E8018NM/C3L46C/J020A17A (f) Ring 21BD/E8018NM/C91046/0217A27A or C4P046/0217A27A*
Indicate the chemical composition (particularly the copper, phosphorus and sulfur content) of these weld metals.
251.7 Indicate which reactor vessel beltline weld metals were utilized for root passes and are within the first quarter of the weld thickness.
251.8 If the chemical co;nposition of the weld metals in Q 251.6 are not known and the weld metals are not located within the first quarter of weld thickness, ut'ilize the upper limit curves in Regulatory l
Guide 1.99 to determine the predicted adjusted reference temperature l
(ARTNDT). Indicate the end of life ART f r these materials and NDT recalculate the pressure temperature limit curves if these weld metals become limiting during reactor vessel life.
251.9 Indicate the inside diameter and minimum wall thickness of the reactor vessel beltline.
3 One of these are typographical errors.
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Licensing Branch No. 2. DL 7
FR&4:
R. Auluck, Pro.1ect Manager Licensino Branch No. 2, DL SitnJECT:
FORTHCOMING MANAGEMENT MEETING WITH WASHINGT0!!
PUPLIC POWER SUPPLY SYSTEM (WPPSS)
DATE A TIME:
Thursday,' September 24, 1981 l
8:30 AM LOCATION:
Roorg P-422 i
Phillips 9uildino Rethesda,' ffD P!!RPOSE:
Presentation bv HPPSS manacement of its organizational structure and technical resources for operation of
'#iP-2 PARTICIDAilTS:
'IRC
- 0. B. Vassallo, O. M. Beckham, C. U. Rivenbark, and R. Auluck NPPSS G. D. Pouchey, G. C. Sorensen and P. Nelson s
P. Auluck, Project !"anaaer Licensino Aranch No. 2 Division of Licensing cc: See next nace Distribution:
" Service Docket File I&E (3) i LBd2 bcc:
l nFLn NRC PnR l
OEisenhut/R Purple PISIC QTedesco TERA ASchwencer TIC l
pa.uluck ACRS (16) d.
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Mr. R. L. Ferguson Managing Director Washington Public Power Supply System P. O. Box 968 3000 George Washington Way Richland, Washington 99352 ccs: Nicholas Reynolds, Esq.
~ Debevoise & Liberman 1200 Seventeenth Street, N. W.
Washington, D. C.
20036 Richard Q. Quigley, Esq.
Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympis, Washington 98504 Mr. Albert D. Toth Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 Richland, Washington 99352 Roger Nelson, Licensing Manager Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Mr. O. K. Earle, Project Licensing Supervisor Burns and Roe, Incorporated 601 Williams Boulevard Richland, Washington 99352 i
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