ML20058G323
| ML20058G323 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 11/24/1993 |
| From: | Kenyon T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9312090175 | |
| Download: ML20058G323 (24) | |
Text
._.
- WWQ,
. p uc,e o
w
.,i Ft UNITED STATES l
I
[kemXf,i j NUCLEAR REGULATORY COMMISSION f
f WASHINGTON, D.C. 20566-0001
\\,,
/
November 24, 1993 Docket No.52-003 r
f APPLICANT: Westinghouse Electric Corporation i
FACILITY:
AP600
SUBJECT:
SUMMARY
OF MEETING TO DISCUSS THE MATERIALS USED IN THE AP600 i
DESIGN On November 10, 1993, representatives of the U.S. Nuclear Regulatory Commis-sion (NRC) and Westinghouse met to discuss the materials used in the design of l
l the AP600. is a list of attendees. -is a copy of the
]
slide presentation made by Westinghouse.
The staff opened the meeting stating that the purpose of this meeting was to better understand Westinghouse's philosophy behind its application of materi-i als in the AP600 design.
The staff wants to ensure that problems (such as inspection accessibility and aging) that have been identified on operating l
plants are being addressed on the advanced designs. The staff wants to know what requirements are being implemented, and which are not (and why). The staff also wants to ensure that the AP600 is inspectable and that there will be no need for relief from the current regulations related to inspection of the facility.
Westinghouse began their presentation stating'that the AP600 design meets current NRC regulations, that it is designed to facilitate inspection and-maintenance, that it addresses operating plant issues, that it employs proven components and materials, and that it considers the guidance of the EPRI ALWR Requirements Document. Some of the operating issues the designer focused on included reactor vessel integrity, use of Alloy 600, use of cobalt alloys, steam generator tube corrosion, and inspection access.
In addition, the designer stated that it is addressing inservice inspection requirements for defense-in-depth non-safety-related systems, and issues concerning the canned-i rotor reactor coolant pump.
Westinghouse then addressed the following topics in greater detail:
inservice inspection requirements and provisions for the design 4
=
safety classification of structures, systems, and components (SSCs)
=
material selection for the reactor coolant pressure boundary, the reactor internals, and the engineered safety features j
the de' sign of the reactor vessel
=
the design of the reactor coolant pump, including the depleted uranium a
flywheel, and the design of the steam generator a
NRC iiE CENTER C8PY 300019
(
9312090175 931124 PDR ADOCK 05200003 C
i'
{
i i November 24, 1993 In summary, the staff expressed its concern with the designer's application of 2
inservice inspection require ~ents and safety classification of SSCs to the j
defense-in-depth non-safety-ralated systems. The staff indicated that it was still reviewing Westinghouse's position on the regulatory treatment of non-i safety-related systems. Therefore, the staff has not completed its review of i
the designer's classification of both safety-and non-safety-related SSCs.
l The staff indicated that its evaluation has the potential to affect aspects of i
quality assurance requirements, constructior,, and inservice inspection and j
testing.
l i
The staff also expressed its concern with the use of the depleted uranium flywheel due to uncertainties involving thermal cycling, aging, and casting properties of uranium, and indicated that further evaluation of the issue was necessary.
At the close of the meeting, the staff agreed to develop additional requests for information to followup its concerns discussed during the meeting.
t W h21SinnMy 9
Thomas J. Kenyon, Project Manager Standardization Project Directorate Associate Director for Advanced Reactors and License Renewal l
Office of Nuclear Reactor Regulation
Enclosures:
1.
List of Attendees 2.
Slide Presentation
)
cc w/ enclosures:
See next page DISTRIBUTION w/ enclosures:
Docket File PDST R/F TKenyon PShea PDR DCrutchfield DISTRIBUTION w/o enclosures:
TMurley/FMiraglia WTravers RBorchardt FHasselberg KShembarger BDLiaw, 7D23 DSmith, 704 DTerao, 7H15 l
GGeorgiev, 7H15 KParczewski, 704 HBrammer, 7E16 TPolich, 10A19 JMoore, 15B18 BDean, EDO ACRS (11) 0FC:
LA:PDST:ADAR PM:PDST!NDARSC:PDST;ADAR-NAME:
PShea fy/)( TKeh sg RArchitbel DATE:
ll/8/b 11/th9_3 11k/93 0FFICIAL RECORD COPY:
DOCUMENT NAME: MTRL_MTG. SUM
i I'
i i
~
I Westinghouse Electric Corporation Docket No.52-003 cc:
Mr. Nicholas J. Liparulo Mr. Victor G. Snell, Director Nuclear Safety and Regulatory Analysis Safety and Licensing Nuclear and Advanced Technology Division AECL Technologies Westinghouse Electric Corporation 9210 Corporate Boulevard P.O. Box 355 Suite 410 Pittsburgh, Pennsylvania 15230 Rockville, Maryland 20650 Mr. B. A. McIntyre Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit Box 355 Pittsburgh, Pennsylvania 15230 Mr. John C. Butler Advanced Plant Safety & Licensing Westinghouse Electric Corporation Energy Systems Business Unit l
Box 355 Pittsburgh, Pennsylvania 15230 Mr. M. D. Beaumont Nuclear and Advanced Technology Division Westinghouse Electric Corporation One Montrose Metro 11921 Rockville Pike
{
Suite 350 l
Rockville, Maryland 20852 l
Mr. Sterling Franks U.S. Department of Energy NE-42 Washington, D.C.
20585 Mr. S. M. Modro EG&G Idaho Inc.
Post Office Box 1625 Idaho Falls, Idaho 83415 Mr. Steve Goldberg i
Budget Examiner 725 17th Street, N.W.
t Room 8002 Washington, D.C.
20503 i
Mr. Frank A. Ross l
U.S. Department of Energy, NE-42 Office of LWR Safety and Technology 19901 Germantown Road Germantown, Maryland 20874 I
i MEETING ATTENDEES AP600 RATERIALS APPLICATIONS l
NOVEMBER 10, 1993 i
MANE AFFILIATION T. Kenyon NRC/PDST I
D. Ekeroth Westinghouse D. Lindgren Westinghouse i
K. Shembarger NRC/PDST M. Jambusara Westinghouse B. D. Liaw NRC/NRR/DE:
D. Smith NRC/NRR/DE/ECGB D. Terao NRC/NRR/DE/ECGB G. Georgiev NRC/NRR/DE/ECGB K. Parczewski NRC/NRR/DE/EMCB H. L. Brammer NRC/NRR/DE/ECGB T. Polich NRR/DRIL/RPEB i
[
i 1
I
=
I AP600 Material and Inspection Regulatory issues i
Meeting with the NRC November 10,1993 i
C 1993 Westinghouse Electric Corp.
p I
Meeting with the Nuclear Regulatory Commission on
]
[.
AP600 Materials and inspection Regulatory Issues November 10,1993 i
l Agenda introduction Lindgren 8:30 Material Selection and Inspection Approach Lindgren 8:40 Operating Plant issues Passive Plant issues inservice Inspection Regulatory Basis Lindgren 8:50 Safety Classification Requirements for Nonsafety-Related Systems inservice inspection Requirements and Provisions Ekeroth 9:15 Requirements for Safety-Related Systems Relief from Section XI Requirements Design for inspectability - Module Design Primary Containment Vessel Material Selection Ekeroth 9:45 General Material Selection Considerations Reactor Coolant Pressure Boundary Reactor internals Engineered Safety Features Reactor Vessel Material Selection and Design Ekeroth 10:15 Long Term Vessel Integrity inspection Access CRDM Penetrations Reactor Coolant Purnp Ekeroth 10:45 Canned Motor Design Flywheel Direct Coupled to Steam Generator Steam Generator Lindgren 11:30 Design Maturity Material Selection Inservice inspection Feedwater Line Cracking Lunch 12:00 Summary Lindgren 1:00 NRC Comments 1:15
t
.m:
an j
- g1 g
1 i
Material Selection and Inspection Approach 1
l The approach of the AP600 in the area of material selection and inservice inspection includes:
l Use Proven Components and Materials j
l Design for inspection and Maintenance Satisfy Regulatory Requirements j
Address Operating Plant issues i
l Consider Utility Requirements j
The differences in inservice inspection requirements and material selection between the AP600 and operating l
plants is a result largely of applying design and operating
]
experience to addressing operating issues.
There are a i
small number of differences due to the passive design.
i i
i i
l
i.~
Operating Plant issues Operating plant issues addressed include:
Reactor vessel integrity Use of Alloy 600 including CRDM penetration cracking Use of cobalt alloys Steam generator tube corrosion Inspection access including use of robotic inspection and maintenance tools l
i
4 j
l: O vs i
j l
Passive Plant issues The passive plant design related issues include the l
following:
]
Inservice inspection requirements for defense-in-depth nonsafety-related system.
Material selection and inservice inspection issues j
associated with, the canned motor reactor coolant Pump 2
1 l
4 6
. ~..
O
(=
g Inservice inspection Regulatory Basis 1
i i
i The inservice inspection requirements are consistent with the safety classification of the equipment and systems.
3 Class A, B, and C are safety-related i
- the requi.ements are the same as in current plants.
Class D systems and components are nonsafety-related defense-in-depth systems and radioactive containing systems.
I 2
C[
i i
j Class D systems and components have a number of controls that provide for quality design, fabrication and operation.
The design and inspection requirements for non-
- 1 safety-related systems are consistent with the practices of high quality industrial quality assurance codes and standards such as ANSI B31.1 and ASME.
Code Section Vill Class D systems are included in ITAACs.
Additional regulatory requirements on selected Class D systems and components are based on the recom-i mendations for the regulatory treatment of non-l safety-related systems (RTNSS).
I For a few Class D systems, additional regulatory oversight will include short term availability controls Additional inspection requirements or construction to more rigorous codes or standards is not recommend-ed.
The use of nonsafety-related systems to terminate or.
mitigate transients and abnormal events is found in l
operating plants. The AP600 is not fundamentally different in this respect.
,,v.
,,,,_..ma._,.,.s
..,__,__.-..,..m.,.....
.........,.......~m-,.....
l'
=n r
Inspection Requirements and Provisions Requirements for Safety-Related Systems ASME Section XI for safety-related components and system Evaluation of major component welds for inservice inspection methods and compliance with Section XI completed.
)
Relief from Section XI Requirements is not expected to be required.
4 5
- .- w Design for inspectability - Module Design Program requirements for inspectability are documented Design will be assessed for inspectability per procedure Plant and module design has accessibility as a design requirement Primary Containment Vessel Access provisions incorporated in design for all ASME XI IWE Inspections i
a
. ~. -
,, ~........ -,
' r' (s -
=
m wa y
i l
l Material Selection i
i General Material Selection considerations l
Control of stress corrosion cracking i
Material Selection limited by design Manufacturing controls on materials Chemistry controls of fluids
]
Control of erosion-corrosion i
Material selection Line parameters controlled Consideration of environmental effects on fatigue W involved with industry initiatives to resolve issue l
i l
l I
i I
1 J
J t
1 i
i i
i
r
/
nu a n'
=
x 1
Reactor Coolant Pressure Boundary Proven materials with proven processes Materials are SA 508 CL3 and 316LN Weld material controlled Reactor internals improved stress corrosion cracking resistance with 304L and weldments Engineered Safety Features Same materials as reactor coolant pressure boundary Cobalt Reduction Identify main contributors Identify replacement materials or limit cobalt 4
1
i Reactor Vessel Material Selection and Design i
Long Term Vessel Integrity Material requirements Fracture toughness in high fluence regions Controlled. chemistry in high fluence regions Weld design j
Design for welds in low fluence areas i
Controlled chemistry in high fluence regions 1
i Core design requirements Low power density core with reflector Inspection Access inspection fully from within Insulation standoff allows access from outside Removable insulation for ease of access l
CRDM Penetrations Material of penetration inconel 690 Penetration design based on Standard W design
Reactor Coolant Pump There are three primary differences between the reactor coolant pump in the AP600 and those in operating plants.
The driver is a canned motor design. Hydraulics are scaled existing designs.
The flywheel is fabricated of depleted uranium and is within the pressure housing.
The pumps are supported by the steam generator.
i 1
~ -,,
a
- n...,
.n--.
1
-r i
The use of a canned motor pump design has a number of i
advantages There are no controlled leakage' seals and thus no i
safety related seal injection water.
i There is no requirement for the cooling water to be safety-related.
Postulated station blackout does not result in a leak I
or small LOCA as may be the case for shaft seal pump seals.
Flywheel is safely housed inside the pressure boundary and flange 1
.i An anti-reverse rotation devise is not necessary to l
assist motor startup.
]
t P
1 d
)
d
.w-.
,o w--_.,
-..<.w,.
...~,.m..
e_,..
,_,....,,._,-,_,.,..,,w._
.-n w..--
. = - -
4.
l
- O a!
l j
The uranium flywheel satisfies design and regulatory 1
l requirements i
i i
j A flywheel is necessary to increase pump coast j
down to extend short term flow through the reactor i
coolant system.
l The uranium alloy is dense and allows for a smaller j
flywheel as compared to steel.
)
j Postulated failure of flywheel is contained within the pressure housing and flange.
The uranium alloy has established material properties.
Westinghouse has manufactured a
prototypical uranium P wheel for bearing testing.
7 i
l 4
i l
1
s w
.3 i
i i
Two canned motor reactor coolant pumps are directly coupled to each steam generator i
4 l
There is no crossover pipe; this is an advantage for some accident conditions.
The pump is mounted motor down to the bottom of l
the steam generator so that the pump and motor are i
self venting j
The hot leg and cold leg are offset in elevation.
l Thus the hot leg is full at "mid-loop" of the cold leg.
i.
)
The location of the pump results in no interference i
with steam generator tube inspection and maintenance.
i i
4 l
i i
i i
L
f Steam Generator The AP600 steam generator is an established design.
The steam generator is an evolution of the Westinghouse Model F design with years of suc-cessful operating experience Surry replacement steam generators Turkey Point replacement steam generators Wolf Creek steam generators l
From the tubesheet up, the AP600 steam generator is the same as the Delta 75 replacement model.
The V. C. Summer replacement steam genera-tors are Delta 75 design.
The Delta 75 design is the current offering by Westinghouse for replacement steam genera-tors.
I i
'Nt g' a AP600 Steam Generator Material Selection The materials and recommended water chemistry in the AP600 steam generator are selected to address tube degradation mechanisms that'have been seen in steam l
generator operation.
The tube material is thermal treated nickel-chromium-iron Alloy 690.
Alloy 690 is considered-"the material of choice for steam generator tubing applications because of its corrosion resistance in a variety of environments."
Selection of Alloy 690 tubing and adherence-to the AVT water chemistry guidelines provide reasonable assurance for maintaining the long-term integrity of the steam generator tubes under current technology assumptions.
The tube support plate material is Type 405 stainiess steel.
These materials have been proven in operation and by testing to be resistant to corrosion mechanisms.
i in addition to the tubes and tube support plates selected elements of the steam generator internals are fabricated of erosion resistant materials
L Steam Generator Inservice inspection The steam generator design provides access to all tubes for inspection and maintenance The steam generator channel head has equal or better access than existing Westinghouse designed steam generators.
The steam generator channel head is designed for robotic inspection and maintenance.
The nozzle dams can also be installed robotically or remotely.
The objective is zero manned entry.
a 4
n,
,a
e 1
l Feedwater Line Cracking The AP600 addresses the issue of cracking on the feedwater line at the steam generator nozzle.
The configuration of the steam generator feedwater nozzle, including a welded feedring thermal sleeve to prevent leaking minimizes temperature stratification.
The geometry of the feedwater line minimizes conditions resulting in thermal stratification.
The feedwater line material is resistant to erosion and corrosion.
ig Summary The AP600 in the area of material selection and inservice inspection :
Uses proven components and materials Is designed for inspection and maintenance access Satisfies regulatory requirements Addresses operating plant issues j
)
.. -.