ML20213C812
| ML20213C812 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 11/20/1978 |
| From: | Pawlicki S Office of Nuclear Reactor Regulation |
| To: | Varga S Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0237, CON-WNP-237 NUDOCS 7811280384 | |
| Download: ML20213C812 (7) | |
Text
'
NOV 2 01978 l
i Docket No. 50-397 MS 05-12 HEMORANDLH FOR:
S. A. Varga. Chief 3
1 Light Water Reactors Branch No. 4 Division of Project Management FROM:
S. S. Pawlicki. Chief Materials Engineering Branch
SUBJECT:
WPPSS NUCLEAR PROJECT N0. 2 (BifR-5)
Plant Name: WNP-2 Suppliers: General Electric; Burns and Roe l
Licensing Stage: OL i
Docket Ninber: 50-397 Responsible Branch and Project Manager: LWR 4; M. D. Lynch Reviewer:
M. L. Boyle Requested Completion Date: November 10, 1978 i
l Description of Task: Q-1 l
Review Status: Additional Infomation Required The Materials Integrity Section of the Materials Engineering Branch.
Division of Systems Safety, has reviewed the FSAR for WNP-2 through Amendment No. 1.
The additional information identified in the attachment is required before our evaluation may be completed.
i l
Additional round one inquires concerning preservice and inservice inspection programs for WNP-2 are in preparation. We expect that these questions will be available by December 29, 1978 and we will submit them I
at that time.
The following table was developed to keep WNP-2. Grand Gulf and Susquehanna, which are under a common review, apprised of questions comon to one or i
both of the other plants. Also the table identifies those questions that l
have been asked of previous applicants.
t 0 3 8 5' O l
- /
i l
S. Varga,
i
~
I Comon Questions I
i
_NP-2 Grand Gulf Susquehanna W
121.1 1 21.1 121.1 i
121.2 121.2 121.2 l
121.3 121.3 121.4 l
121.5 121.3 121.4 121.6 121.4 121.5 121.7 121.6
[
121.8(Hatch-2) 1 21.5 121.7 121.9 121.6 121.8 4
After receipt of an acceptable response to any of the comon questions, the Materials Engineering Branch will send the response to the LPM's for theotherplant(s).
Sgtaal sigwl by
[s,S.Pawlicki S. S. Pawlicki, Chief Materials Engineering Branch Division of Systems Safety l
Office of Nuclear Reactor Regulation
Enclosure:
As stated cc w/ enc 1:
cc w/o enc 1:
i R. J. Mattson, DSS R. S. Boyd, DPM l
D. G. Eisenhut, DOR W. Pike. MPA D. M. Crutchfield, NRR J. P. Knight, DSS Distribution:
H. F. Conrad. DSS Docket File (50-397) i R. M. Gamble DSS NRR Reading File l
R. J.-Bosnak. DSS MTEB Reading File S. S. Pawlickt. DSS RE 1.1-1 M. D. Lynch, DPM C. O. Thomas DPM l
S. Miner, DPM t
G. B. Georgiev, DSS R. A. Hemann, DSS M. L. Boyle DSS
_ gSJS: E5h. QSS:57E'Bb DSS:MTEB o,,,c.
- MLBo 16: b iRMGamD
.cki
.. 11/16/78._
..11////78_..
11 78
.m.-
NRC PORM 313 (p.76) NRCM 0240 W us s. sovanmusm? Painvine orrecs sore -one.sa4
121-1 121.0 MATERIALS ENGINEERING BRANCH - MATERIALS INTEGRITY SECTION 121.1 Provide a sketch of. the WNP-2 reactor vessel (including (5.2.3) dimensions) showing all longitudinal and circumferential welds, and all forgings and/or plates. Welds should be identified by a shop) control number (such as a procedure qualification number, the heat of filler metal, type and batch of flux, and welding process. Each forging and/or plate should be identified by a heat number and material specification.
121.2 Supply the following information for each of the ferritic (5.2.3) materials of the pressure retaining components in the reactor-coolant pressure boundary of the WNP-2 plant:
(1) 'The unirradiated mechanigal properties as required by the testing programs in Section III of the ASME Code 'and I
Appendix G of 10 CFR Part 50 (test results to be presented should include Charpy V-notch, dropweight, lateral If any of these properties have not been Nermined h)a.
expansion, tensile, upper shelf energy, T and RTg test method reouired by Appendix G of 10 CFR Part 50, state the actual test procedure used and/or the method used to estimate the test result together with a complete technical justification of the procedure used.
(2)
Identify the material (s) in the reactor coolant pressure boundary that will limit the pressure-temperature operating curves at the beginning-of. life.
i For each reactor vessel beltline weld, plate or forging provide the following additional information:
(3) The chemical composition (particularly the Cu, P and S content) and the maximum end-of-life fluence.
(4)
The' relationship used to predict the shift in RT and percent decrease i.n upper shelf energy as a funcbn of neutron fluence.
(5)
Identify the material (s) in the reactor coolant pressure boundary that will limit the pressure-temperature operating curves at the end-of-life.
1 21.3 The FSAR states that compliance with Appendix G of 10 CFR (5.3.1.5)
Part 50 and Appendix G of Section III of the ASME Code was not possible for components purchased pri'or to the issuance of the Summer 1972 Addenda of the ASME Code without replacement of large amounts of material, reworking of fabricated components and the revision of most all of the design analyses for the components.
e t
E
121-2 The details of the method.of compliance as stated in the FSAR are insufficient to identify the areas of noncompliance with Appendix G of 10 CFR Part 50. The applicant should state specifically those sections in which strict compliance with
' the regulations was not achieved.
The technical bases for the proposed' alternate methods used to satisfy the requirements of those sections of Appendix G of 10 CFR Part 50 where strict compliance was not achieved should be presented. These bases should include technical justification to demonstrate that the proposed alternatives provide acceptable safety margins relative to the Appendix G requirements.
1 21.4 FSAR 5.3.1 states that paragraph C.2.b of Regulatory Guide 1.65 (5.3.1) was not followed in the ultrasonic examination of the reactor (C.2) vessel closure studs. Paragraph C.2.b indicates that the studs and nuts be ultrasonically examined after final heat treatment according to ASME Specification SA-388. Provide the details of the UT procedure used so that comparability with the requirements of SA-388 may be evaluated.
121.5 Paragraph II.C.2 of Appendix H, 10 CFR Part 50 states:
(5.3)
" Surveillance capsules containing the surveillance specimens shall be located near but not attached to the inside vessel wall in the beltline region,...."
FSAR Section 5.3 indicates that the capsule holder brackets were welded to the reactor pressure vessel innerwall.
Present sufficient design and fabrication detail to demonstrate that the capsule attachments were designed and constructed in accordance with accepted standards, such as the ASME Code Section III rules for attach-ments to vessels.
121.6 In FSAR Sections 1.6 and 5.3 General Electric Report NED0-20631, (1.6)
" Mechanical Property Surveillance of Reactor Pressure Vessels (5.3) for General Electric BWR-6 Plants," dated March 1975, is referenced. At present this report is not in the OL docket file.
In order to make an evaluation of t.ompliance with Appendix H, of 10 CFR Part 50, for this plant, the information referenced by this GE report must be submitted for review.
121.7 To provide assurance that high' energy turbine missiles will not (10.2.3) be produced at operating speed or design overspeed provide documentation (including the results of material property test-ing) to show the degree of conformance of the turbine-generator with the guidelines in SRP 10.2.3, " Turbine Disk Integrity,"
paragraph II, " Acceptance Criteria."
J i
~
1 121-3 121.8 The following infomation is necessary to demonstrate that the
-(Ref.HNP-2 feedwater inlet nozzle thennal sleeve /sparger design has been Response to evaluated with due consideration to nozzle cracking due to 121.15 and thermal cycling and that a program of scheduled augmented 121.18) inservice inspections, with a sensitive method that will assure detection, has been developed:
(1) The technical basis to assure the structural integrity of both the feedwater inlet nozzle and the sparger.
(2) An evaluation of the feasibility of automated ultrasonic testing (UT) fixtures installed on all feedwater inlet nozzles with particular attention on examination of the nozzle bore region.
(3) An evaluation of the feasibility of perfoming the internal surface examination by magnetic particle methods.
Your response should contain:
a description of the nozzle and sparger design including dimensions, materials of construction and weld locations.
description of analyses and test data, referencing if necessary data previously submitted to the staff where directly appropriate for this plant.
projected crack growth rates, stress levels and usage factors for both the nozzle and the sparger should be described in detail.
any plant modifications.that are planned to reduce the feedwater to reactor water temperature differential during lower power operation.
l any instrumentation that will 'be installed in the reactor i
to verify the conclusions of the design analysis should be I
identified.
l l
Several ultrasonic testing concepts and procedures have been l
used to examine the feedwater inlet nozz'le regions in operating plants. Define the specific ultrasonic testing procedure that will be used on this plant. Discuss the influence of local grindouts on crack detection on your ultrasonic testing method.
l In addition, provide a description of the augmented inservice inspection (ISI) program to be implemented including scheduled surface examination, ultrasonic testing and verification of the
.Y
121-4 leak tight integrity of the thermal sleeve to safe end joint on.
all nozzles. The essential elements of an acceptable program are given below:
Augmented Inservice Inspection Plan (1) Preservice Examination - Preservice UT examination should include all nozzle inner radius, bore, and safe end regions.
In addition, a preservice surface examination should be performed on the accessible regions of all nozzle inner radii.
(2)
Inservice Examination - To confirm the continuing structural integrity, the following examinations should be performed:
(a) At each scheduled refueling outage, an external UT examination of all feedwater nozzle inner radii, bore and safe end regions.
(b) After 50 startup/ shutdown cycles but prior to 70 cycles a surface examination of the accessible regions of all nozzle inner radii. The definition of startup/ shutdown cycles and the procedure for liquid penetrant examination is contained in report NUREG-0312,
" Interim Technical Report on BWR Feedwater and Control Rod Drive Return Line Nozzle Cracking."
(c) Subsequent surface examinations of the accessible region of all nozzle inner radii should be performed at the earlier of (i) every other scheduled refueling outage, or (ii) at the scheduled refueling outage after 20 but prior to 40 startup/ shutdown cycles after the last surface examination.
(3) Thermal Sleeve to Safe End Joint - An examination method, such as a leak test should be developed to confirm the continuing structural and leak tight integrity of the thermal sleeve to safe end joint.
Acceptance Standards (1) All UT indications evaluated to be cracks should be verified by appropriate surface examination and removed by local grinding.
(2) All surface indications evaluated to be service induced cracks should be removed by local grinding.
I
~
121-5 (3) The UT inspection personnel should be required to demonstrate supplemental qualifications by either (1) past successful experience in locatin BWR feedwater inlet nozzles or (g and identifying cracks in ii) performing a qualific-ation test on a full size unclad nozzle mockup.
Recording and Reporting Standards Requirements for recording of indications and reporting of inspection results are contained in report NUREG-0312.
121.9 Considering the recent BWR service experience of cracking of the vessel nozzle and wall associated with the control rod drive return line, we require a description of any proposed plant modifications (such as changes in material, location of the CRD return line, deletion of the CRD return line, etc.)
that will preclude such cracking and a complete technical justification for the proposed modifications.
b
- t
--