ML20054B836
| ML20054B836 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 04/12/1982 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Fiedler P JERSEY CENTRAL POWER & LIGHT CO. |
| Shared Package | |
| ML20054B837 | List: |
| References | |
| TASK-03-01, TASK-3-1, TASK-RR LSO5-82-04-030, LSO5-82-4-30, NUDOCS 8204190241 | |
| Download: ML20054B836 (7) | |
Text
,
s April 12,1982 Docket flo. 50-219 LS05-82-04-030 m
e b y C
'i ifr. P. B. Fiedler p
c sh b Vice President and Director - Oyster Creek qV o
Oyster Creek fluclear Generating Station zR gf^.p.
9 7
Post Office Box 38P tr p
Forked River, Nuw Jersey 03731 fe v*
C 4
- 9
Dear Mr. Fledler:
tu s
SUBJECT:
SEP TOPIC III-1, QUALITY GROUP CLASSIFICATION OF COMPONENTS AND SYSTEMS (0YSTER CREEK HUCLEAR GENERATINGSTATION)
Enclosed is the staff's draft safety evaluation of SEP Topic III-l for the Oyster Creek Nuclear Generating Station. Our evaluation (Enclosure
- 1) is based upon our contractor's final evaluation (Enclosure 2) of this topic. This assessment compares your facility with the criteria currently used for licensing new facilities. You are requested to ex-amine the facts upon which the staff has based its evaluation and res-pondeetther by confirming that the facts are correct, or by identifying errors and supplying the correct information.
The staff was unable to complete this topic due to the lack of infoma-tion of original design requirements for vattous components. We have concluded, for those components where a comparison of codes was possible S{pf that the changes in the codes since the. original design do not signifi-3 cantly affect the safety of the plant. Based on our sampling of code comparisons to Sate we do not expect the remaining items to pose a sig-f /
nificant hazard to safe plant operation.
/
h p f 4 O c L,5 f l Your response is requested within 30 days of receipt of this evaluation. g If no response is received in this time we will assume the evaluation is 6.SN correct.
Sincerely, i3 -
8204190241 820412 Dennis M. Crutchfield, Chief DR ADOCK 05000219 Operating Reactors Branch No. 5 P
PDR Division of Licensing Eni;losures:
As stated
, J) omcrdec..w/ enclosures:...SEP.Bh
..S Ef B,,,,
.iPBM9,Y 0RB#5.
,,,,,0 RB,Y,5,,,,h:DL
, Russel1,.,,
_J Lom,ba,,,
@r,u,t,c h,f,i el d, haba,s,,,,,,,
suamt blSee. next pap
. MBoy,1 e,.: bl,,GCNa t i,na,.
W
,/g,/82 4/I.!82, 4L1/12...
. 4L6./82..,..,4.(,9.,(,8 2,,,,, u it,,]je,p,,,
4 em OFFICIAL RECORD COPY usm im_my,3 Q rORM 318 00 8C) NRCM Cao
e*
Mr. P. B. Fiedler cc G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittnan, Potts and Trowbridge c/o U. S. NRC 1800 M Street, N. W.
Post Office Box 445 Washington, D. C.
20036 Forked River, New Jersey 08731 J. B. Lieberman, Esquire Commissioner Berlack, Israels & Lieberman New Jersey Department of Energy 26 Broadway 101 Commerce Street New York, New York 10004 Newark, New Jersey 07102 Ronald C. Haynes, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 J.. Knubel BWR Licensing Manager GPU Nuclear 100 Interplace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New Jersey Department of Law and Public Safety 36 West State Street - CN ii2 Trenton, New Jersey 08625 Mayor Lacey Township 818 Lacey Road Forked River, New Jersey 08731 U. S. Environmental Protection Agency Region II Office ATTN:
Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Licensing Supervisor Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731 e
SYSTEMATIC EVALUATION PROGRAM TOPIC III-l OYSTER CREEK NUCLEAR GENERATING STATION TOPIC:
III-1, Classification of Structures, Components and Systems (Seismic and Quality)
I.
INTRODUCTION SEP plants were generally designed and constructed during the time span from the 1950's to the late 1960's. The plants were designed to generally recognized codes, standards and criteria in effect at that time; however, the codes, standards and criteria have been periodically revised. Therefore, the SEP plants may have been de-signed and constructed to codes, standards and criteria no longer in effect or acceptable to the NRC.
The purpose of Topic III-l is the review of the classification of structures, systems and components of as-built plants compared to the current classifications required for seismic and quality groups in the codes, standards and criteria. Since the review of seismic classifications is addressed in other SEP topics (See Section III of this evaluation), this topic has been limited to the evaluation of quality group classifications.
II.
REVIEW CRITERIA The review criteria for this topic are presented in Appendix A of Technical Evaluation Report C5257-431, " Quality Group Classification of Components and Systems - Oyster Creek Nuclear Generating Station,"
prepared for the NRC by Franklin Research Center (Attached).
III.
RELATED SAFETY TOPICS The scope of review for this topic was limited to avoid duplication of effort since some aspects of the review are performed in related topics.
As stated previously, the seismic aspect of this topic has been deleted.
The quality aspect for the reactor vessel and steam generators (PWRs only) and the quality assurance have been deleted.
The related safety topics, and the subject matter covered in the topics, that cover the aspects deleted 'in Topic III-l are identified below.
III-6 Seismic Design Considerations III-7.B Design Codes, Design Criteria, Load Combinations and Reactor Cavity Design Criteria V-6 Reactor Vessel Integrity V-8 Steam Generator Integrity XVII Operational Quality Assurance Program The resolution of Topic V-8 is part of Unresolved Safety Issues A-3, A-4 and A-5.
l t
IV.
REVIEW GUIDELINES The review guidelines are presented in Section 3 of Report C5257-431 (Attached).
V.
EVALUATION The basic input for this report is Table 4.1 in Section 4 of Report C5257-431.
Table 4.1 is a compilation of all systems and components which are required to be classified by Regulatory Guide 1.26.and the original codes, standards and criteria used in the plant design.
After comparing the original codes, standards and criteria with those currently used for licensing facilities the following areas were identified where the requirements have changed:
1)
Fracture Toughness
- 2) Quality Group Classification
- 3) Code Stress Limits
- 4) Radiography Requirements
- 5) Fatigue Analysis of Piping Systems An evaluation of each of these areas is presented in Section 5 of Report C5257-431 with a detailed discussion included in the Appendix of the report.
We have determined that changes in the following areas have not signi-ficantly affected the safety functions of the systems and components reviewed in this report:
- 1) Quality Group Classification
- 2) Code Stress Limits
- 3) Fatigue Analysis of Piping Systems In the remaining two areas we have concluded the following:
1)
Fracture Toughness - The current code requires that pressure re-taining materials be impact tested.
For 5 of 40 components re-viewed, sufficient information was available to exempt them from this requirement.
- 2) Radiography Requirements - For pressure vessels and pump casing, we have concluded the following:
a) Vessels built to ASME III (1965) Class A or ASME VIII (1965) satisfy current radiography requirements for Class 1 and Class 3 vessels, respectively.
e
. t b) Vessels built to ASME III (1965) Class C requirements and currently classified as Class 2 or Class 3 satisfy current radiography requirements for Category A or B joints.
c) Category C joints in current Class 2 vessels built to Class C requirements do not satisfy current radiography requirements.
For piping and valves, we have concluded that they meet current radiographic requirements provided Code Case N-7 was invoked in the design and fabrication of the components.
Our review has not identified any significant deviations from past codes.
However, we were unable to complete our evaluation due to insufficient information for the following:
1)
Fracture Toughness - For 35 of 40 components there is insufficient information on materials to complete our review. The licensee should provide the necessary information using the format provided in Tables A4-4 through A4-6 in Appendix A of Report C5257-431.
Table 5-1 of the Report identifies those components for which this information is necessary.
- 2) Radiography Requirements - The licensee should provide the following information:
a)
Indicate those components for which Code Case N-7 was invoked.
b) The radiography requirements imposed on Class 2 vessels, Class 1 and 2 piping and valves, and Class 1 and 2 pumps.
- 3) Valves - Provide, on a sample basis for Class 1, 2 and 3 valves, in-formation regarding the design of the valve in order to evaluate if they meet current body shape and pressure - temperature rating require-ments.
Valves designed only to ASME Code Section I should be checked to determine whether thermal stress and cyclic loading requirements were accounted for.
- 4) Pumps - For the recirculation system pumps, a demonstration of compli-ance with the current fatigue analysis requirements should be provided.
All pumps except the liquid poison system pumps were designed accord-ing to ASA B31.1 and ASME Code Section I.
However, since pumpsdesign is not covered under these codes, it seems more likely that Section III of ASME Code would have been used for pump design.
Correct infor-mation, calculations, and evaluate if they meet current standards.
4
- 5) Storage Tanks - Provide the following:
a) Confirm that the atmospheric storage tanks meet current compressive stress requirements.
b) Confirm that the 0 to 15 PSIG storage tanks meet current tensile allowables for biaxial stress field conditions.
c) The Liquid Poison System and Liquid Waste System tanks were designed to API 650.
The requirements of API 650 are not comparable to present design codes. Therefore, re-evaluate the design and contruction of these tanks against current criteria.
d)
It was indicated that the condensate tank was designed to ASME Code Section I; however tank design is not covered under Section I, therefore, re-evaluate the design and construction of this tank against current cd teria.
- 6) Piping - Piping systems designed only to Section I of the ASME Code should be investigated to determine how the thermal stress and cyclic loading requirements were accounted for. The following piping systems should be investigated:
a) Automatic depressurization system piping, b) Piping from the reactor vessel to the first isolation valve external to the drywell, c) Control rod drive housing, d) Emergency system isolation condenser piping; and e) Service water system piping up to the first isolation valve.
- 7) Missing Information - (1) information missing from Tables 4-2(a),
4-2(b), and 4-2(c) of this report regarding the code, code class, and code cases used in designing 9 of 49 components should be provided; (2) assumptions on code editions that were made in order to complete Table 4-1 should be confirmed; (3) the assumption made on the temper-ature drop (100'F) from 100% power to 0% power for piping should be confirmed; (4) specifications for the main steam isolation valve and the standby gas treatment system should be provided.
- 8) Clarification of Infomation - Table 6-1 identifies (1) discrepancies between information supplied by the licensee and information in appli-cable codes and standards, (2) necessary information that has not been supplied by the licensee, and (3) assumptions made where information was not supplied by the licensee.
The licensee should address all items listed in Table 6-1, clarify all discrepancies, and supply all the mis-sing information.
A more detailed explanation of the information to be provided may be found in Report C5257-431.
d
w* !
o VI.
CONCLUSION We have determined that for the following, changes between current and original code requirements for Oyster Creek will not significantly af-fect the safety functions of the systems and components reviewed:
- 1) Quality Group,
- 2) Code Stress; and
- 3) Fatigue Analysis for Piping Systems.
We were unable to complete our review due to insufficient information regarding various other systems and components. The required infor-mation is discussed in Section V of this evaluation.
.i l
Based on our sampling of code comparisons to date we do not expect the remaining items to pose a significant hazard to safe plant operation and therefore, have determined that the schedule and need for providing the remaining information can be determined during the integrated plant safety assessment.
t Q
i s
f t
e
e-
[NCLoSvRE2 TECHNICAL EVALUATION REPORT i
~~ ~ ~
QUALITY GROUP CLASSIFICATION t
?3 OF COMPONENTS AND SYSTEMS 3*
JERSEY CENTRAL POWER AND LIGHT COMPANY g;
OYSTER. CREEK NUCLEAR. POWER. PLANT.
.,n
-w,mnermwwsmamn-mmmaer car--marameauxam-vma-r-
-+-
g N
NRC DOCKET NO.
50-219 FRC PROJECT M
??
NRC TAC NO. 41595 FRC ASS 1GNMENT 17 I)1 NRC CONTRACT NO. NRC-03-79-118 FRC TASK 431 y
6 g
- N S. Tikoo
.f Franklin Resaarch Center Author:
A. Gonzale:
20th and Raca Street 1 3erkovie:
}
Philadelphia. PA 19103 FRC Group Leader:
A. Gonzales Preparea for Nuclear Regulatory Commissicn Washington. D.C 20555 Lead NRC Engineer: n. Boyle 3
h March 24, 1932 wd4 Kg This report was prepared as an account of work sponsored by an e;ency of
)$
the United States Govemment. Neither the United States Govemment nor any agency theraof, cr any of the;' employees, makes any warranty, ex-lg pressed or implied, er assumes any legal fiatslity or responsibility for any third party's use, or the resu!:s of such use. of any information. apparatus, product er process disclosed in inis re::c't, or represents that its use by y
such third party would not infringe private 4y owited rights, i
M 1 9l i
3
.m
- h D
T lU JJ Franklin Research Center l
'y A DMsien c The Franklin Institute x
Tbt Sen.sein 5 anktr. Parney. Ph.'.a. Pa 9103 (215) 648 : OCO 82 o 3 11 0l @ )lA s)
/
E_
o
=
TER-C5257-431 O
CONTENTS Section Title Page 1
INTRODUCTION.
1 2
SCOPE OF THE EVALUATION.
2 3
METHOD OF REVIEW.
5 4
QUALITY CLASSIFICATION OF SYSTEMS AND COMPONENTS.
6 5
EVALUATION OF SPECIFIC CCMPONENTS 22 5.1 General Paquirements 22 5.2 Pressure vessels 29 5.3 Piping 30 5.4 Pumps 31 5.5 Valves.
34 5.6 Storage Tanks 35 6
CONCLUSIONS AND RECOMMENDATIONS 37 7
REFERENCES 44 APPENDIX A - REVIEW OF CODES AND STANDARDS iii dk n.; Franklin Research Center s w orNvvanm.w.
i
- e.
l I
TER-C5257-431 4
}
=
l FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center i
under a contrW. iitd the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical i
assistance in support of NRC operating reactor licensing actions. The i
technical evaluation was conducted in accordance with criteria established by the NRC.
l Mr. L. Berkowitz contributed to the technical preparation of this report through a subcontract with Innovation Technology, Inc.
1 i
i J
4 e.-2--
J-y C !! Frank!in Research Center a wea en 3. rreon
.a,.
= n-
-,,nn.--m
~
c--
-,w-
e b
TER-C5257-431 l
l.
INTRODUCTION Systems and components in nuclear power plants should be designed, fabricated, installed, and tested to quality standards that reflect the importance of their safety functions. This is the concern addressed by the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide 1.26 [1], " Quality Group Classifications and Standards for Water, Steam, and Radioactive" Waste-Containing Components of Nuclear Power Plants," which classifies components into four Quality Groups, A, B, C, and D, and gives the standards applicable to each group.
The systems and components of plants being reviewed as part of the
~
Systematic Evaluation Program (SEP) were designed, fabricated, installed, and tested to standards different from those applied today. This report is the result of work that addresses the safety margins of these systems and components in light of the changes that have taken place in licensing criteria.
The work is part of SEP Topic III-1, " Classification of Structures, Systems, and Components (Seismic and Quality)." NRC has divided this topic into two technical areas:
(1) Seismic review, which will be performed by the NRC, and (2) Quality Group review, which this report addresses for the Oyster Crees Nuclear Power Plant.
i This report was prepared by the Franklin Research Center (FRC) under NRC Contract No. NRC-03-79-ll8.
4 Mi) Franklin Research Center awoeenn. rwa..
a TER-C5257-431 2.
SCOPE OF THE EVALUATION The SEP concerns a review and assessment of the safety of older nuclear plants on the basis of current licensing criteria.
Topic III-l is one of 137 SEP topics. Of the 11 SEP plants, the following 10 are being reviewed:
Plant Name Docket No.
FRC Task No.
Palisades 50-255 17428 Ginna 50-244 17429 Dresden Unit 2 50-237 17430 Oyster Creek 50-219 17431 Millstone Unit 1 50-245 17432 San Cnofre Unit 1 50-206 17433 Big Rock Point 50-155 17434 Haddam Neck 50-213 17435 Yankee Rowe 50-29 17436 Lacrosse 50-409 17437 Specifically, Tbpic III-l entails a review of standards in effect from 1955 to 1965 used in the design of systems and components in older plants, and the 1977 American Society of Mechanical Engineers (ASME) Boiler and Pressure vessel (B&PV) Code as supplemented through the Summer 1978 Addenda (2,3].
The ob]ective of the present evaluation is to assess the ability of systems and components in the Oyster Creek Nuclear Power Plant to perform their safety functions as judged by current standards. This involves two steps:
(1) comparison of current codes and standards with those used in the design, fabrication, installatien, and testing of the plant's systems and components to identify significant differences that might affect structural integrity, and (2) assessment of the effect of these differences on the safety margins of the systems and components.
The scope of this evaluation is limited by or to the following:
1.
Table of Systems and Components (4}, compiled by the NRC, corrected and completed by Jersey Central Power and Light (JCP& L). This table 1.
Plant discussed in this report.
- p....; FranWin Research Center a ww w n Frene m mi.
TER-C5257-431 contains the quality group classification, the current code, and the code used for the listed systems and components when the plant was designed. When the information in the table was incomplete, FRC completed it as well as possible (see Table 4-1) (4).
2.
Information in the Final Safety Analysis Report (FSAR) or a similar document [5].
3.
NRC Regulatory Guide 1.26, Revision 3 [1].
4.
Standard Review Plan 3.2.2 (6].
5.
Major older codes and standards: American Standards Association (ASA) B31.1 (1955), " Code for Pressure Piping" [7]; ASME 1965 Boiler and Pressure Vessel Code,Section I, " Rules for Construction of Power Boilers," (8]; ASME 1965 Boiler and Pressure Vessel Code,Section III, " Rules for Construction of Nuclear Vessels" (9), and Section VIII, "Unfired Pressure vessels" [10]; and applicable Code Cases for ASA B31.1 and ASME VIII.
6.
Current code:
1977 ASME Boiler and Pressure Vescel (B&PV) Code,Section III, Division 1, to include the General Requirements (articles with "NA" subscript), Subsection NB, NC, and ND, and Appendices, supplemented through the 1978 Summer Addenda [2].
7.
Quality Group D components are not considered in this evaluation.
8.
Although discussed in this report, quality assurance for design and construction is outside the scope of the SEP.(1)
Also, the following subjects are explicitly excluded because they have been addressed under other SEP topics:
Tooic Descriction III-5.A Effects of Pipe Break on Structures, Systems and Components Inside Containment III-5.B Pipe Break Outside Containment III-6 Seismic Design Considerations III-7.A Inservice Inspection, Including Prestressed Concrete Containments with Either Grouted or Ungrouted Tendons 1.
Letter f rom S.
Bajwa to S. Carf agno, dated December 10, 1981.
g_%_ 1 Franklin Research Center a tw N n.n.on u.
1 J
l
}
TER-C5257-431 1
1 i
Topic Description 1
III-7.B Design Codes, Design Criteria, Load j
Combination, and Reactor Cavity Design j
Criteria I
j III-7.D Containment Structural Integrity Tests l
III-9 Support Integrity V-3 Overpressurization Protection V-6 Reactor Vessel Integrity V-8 Steam Generator Integrity IX-6 Fire Protection i
1 I
i 4
A J
(
i I
I i
i r
l 9
i l
D i
Od Franklin Research Center a w ea #n,.r, - w a.
I a -. -,,.--. - - -, - -,, - - - -. - - - - - -. ~. - - -. - - -
=
TER-C5257-431 3.
METHOD OF REVIEW i
l To accomplish the objective of this evaluation, FRC performed the review as follows:
1.
Components from the Table of Systems and Components (Table 4-1) referred to in Section 2 were listed in three tables according to Quality Group. For example, all Quality Group A vessels, piping, valves, pumps, and storage tanks are listed in one table. Table 4-2(a) contains Quality Group A components, Table 4-2(b) Quality Group B components, and Table 4-2(c) Quality Group C components.
p l
Within each table, the components are arranged according to type.
2.
Major older codes identified in Table 4-1 were compared against the current code.
Results of the review are given in Appendix A.
3.
The results in Appendix A were used for a comparative analysis which formed the basis for an engineering judgment of the safety margins exhibited by the systems and components by current quality require-ments. Details are given in Section 5.
Appendix A lists all the requirements of the current code (the 1977 ASME B&PV Code,Section III with Addenda [2]) and indicates which requirements are considered applicable and significant for structural integrity (designated as "A"),
which are not considered significant (designated as " *), and which are j
outside the scope of this review (designated as "O").
For each significant l
requirement in the current code, a similar requirement was sought in the older I
codes.
The major older codes for the Oyster Creek plant are ASA B31.1 (1955)
(7) and the 1965 ASME B&PV Code, Sections I, III, and VIII (8, 9, 10].
Differences between significant requirements, such as additions to the older codes, were reviewed, and recommendations were made for assessing their impact on the safety margin of the particular component.
Knowledge of the historical development of the codes and the reasons for the changes was an important element in making effective comparisons. A
. literature survey, supported by consultation witn experts in the field, helped to identify certain changes for special attention, e.g.,
changes in design criteria, analytical methods, load combinations, quality assurance require-ments, fabrication tecnniques, and testing requirements.
l Ah $0 Frankhn Res,earch Center a cme m r.-.eu,
~ _ _.. - - - - - _ _-
a TER-C5257-431
~
4.
QUALITY CLASSIFICATION OF SYSTD4S AND CCMPONENTS Systems and components are Quality Group classified according to the safety functions to be performed.
Table 4-1 contains the systems and components for the Oyster Creek plant, the code required for current licensing criteria, based on NRC Regulatory Guide 1.26 (1) and Section 50.55a of the Code of Federal Regulations (3], and the codes and standards used when the systems and components were originally built. The table also contains information regarding the seismic classification of the systems and components.
r The following systems are listed in Table 4-1 with their respective t
components:
Reactor Coolant System Recirculation System Isolation Condenset System Liquid' Poison System Core Spray System Containment Spray System Shutdown Cooling System Automatic Depressurization System Stanoby Gas Treatment System Reactor Cleanup Demineralizer Syste=
Spent Puel Storage Facility i
Reactor Vessel Head Spray System
{
Concensate/Feedwater System Main Steam System Reactor Shutdown Cooling System Reactor Building Closed Cooling Water System Liquid Waste System i
Service Water System Structures (for information only, not in the scope of this review).
Table 4-2(a) lists all Quality Group A components, Table 4-2(b) lists all Quality Group a components, and Table 4-2(c) lists all Quality Group C compo-nents.
These tables group components as pressure vessels, piping, pumps, valves, 1
and storage tanks and give the major code used when the component was built.
Table 4-2 (d) is an index of systema abbreviations and definitions. More information on the review procedure for System Quality Group Classification is given in tne Standard Review Plan, Section 3.2.2 (6].
5
2 Table 4-1 p
E" gh Classification of Structures, Systems, and components f
f$
Oyster Creek Nuclear Power Plant BE
- e. 3 Quality Classification Ji f, codes azul Codes aux!
Seismic Classification
!.h Structures, Systems, Standards Standards Used Used in In arul Caeponents RG 1.26 (IL in Plant Design ( 2)_
RG 1.29 Plant Design Remarks 4
REAC11m COO!. ANT SYSTtM Reactor vessel ASME III ASME I (1965)
Category I Class I NAI3I Claus 1 Reactor Vessel Supports Category I Class I HA Reactor Vessel Internals ASME III 7
Category I Class I NA Class 1 8
RECIRCU!ATIOtt SYSTtM Pumps ASME III ASA B31.1 (1955)I4I Category I Class I class 1 Valves ASME III ASME I (1965)
Category I Class I Class 1 Pipirej ASME III ASME I (1965)
Category I class I class 1 ASA B31.1 (1955)
H i
to 4
1.
ASME III stands for ' Boiler and Pressure vessel Code,"Section III, Division 1, published by the American Society of Mechanical Erujineers, 1977 blition with Addenda through Summer 1978.
vi 1
2.
When plant design is in accordance with Sections I, 111, azul VIII of ASME Boller and Pressure Vessel Code, 1965 U
tilition is implied.
y 3.
11A irxilcates additional information provided in this table that is outside the scope of this report.
g FHC notes that pump arnt heat exchanger design is not covered under ASME I or ASA B31.1.
e 4.
i t
E t
i I
l j.1 s
g_.,
r
!/
g",3 L Tatile 4-1 (Cont.)
l I
Ic$
a :2 Quality Classificatinn jm Codes and Codes and Seismic Classitfeation y
gg Structures Systems, Standards Standards Used Used in i
s "y ar.d Comtwinent s ItG 1.26 (1)_
in Plant Design (2)
RG 1.29 Plant Design Remarks 5
h0 s3 12tEl(GENCY SYSTEMS 0
Isolation Condenser i
Category I Class I j
C14ss 2 Class A i
4 Shell ASME III ASME VIII (1965)
Category I Class I j
Class 3 r
a 1
ca I
Category I Class I l
Class 2 t.lqisid Poison System Punips ASME III ASME III (1965)
Category I Class I Class 2 Class C Pipler), Fittings, ASME III ASA B31.1 (1955)
Category I Class 1 and Valves Class 2 and Nuclear Code Cases (5)
Tank ASME III API-650 (1964)*
l-Class 2 g
l m
1 W
b vi i
- J
- Inf ormation assumed liecause it is not available at this time.
j j
S mcific nualear code cases were not listed. Tlits information is required.
y 5.
i a
i
%3 1
Y i
e
i e
Table 4-1 (Cont.)
j[ >
Uuality Classification ri f Codes and Codes and Seismic Classification h/
Stsuctures, Systems, Standa r d s Standatds Used Used in 26 aint C<usi1=xiesit s IM; 1.26 ( 1)_
in Plant Design (2)
Ht; 1.29 Plant Design Remarks t3
+
f,3 Core Spray System
- 0 50 (h
Pumps ASE Ill ASA B31.1 (1955)I4I Category I Class I Component jn Class 2 not listed in
[ *,
original y
table, but given in FSAR I
- p. VI-6-1 Pipin9, Fittings, ASME 111 ASA D31.1 (1955)
Category I Class I and Valves Class 1 Piping, Fittings, ASME !!!
ASA B31.1 (1955)
Category I class 1 4
azul valves Class 2 e
Wactor 15a ilding Closed Idx>p Cooling System l' umps ASME III ASME I (1965)
Category I Class I Component not Class 3 ASA 831.1 (1955)III given in original table, but given in FSAR Sec. X, para.
3.4(a) lleet Excitainge rs ASME Ill ASME I (1965)
Cat'egory I Class I Same as above Class 3 ASA B31.1 (1955) g N
Piping, Fittings, ASME III ASME I (1965)
Category 1 Class I g
and Valves Class 3 ASA B31.1 (1955) un and Neclear y
Code cases (5) y a-Automatic Depressuriza-ASME III ASME I (1965)
Category I Class I U
titan System Class 1
i Table 4-1 (Cont.)
p, IE *,
Ouality Classification
{.
Codes and Codes and Seismic Classification a 5-Structures, Systesus, Standards Standards Used Used in g
arut Coauponents hG 1.26 (1) in Plant Design (2)
HG 1.29 Plant Design Remarks
?$
l f(
Service Water System
'S s p Pipirvj, Fittings, ASME III ASME I, IK (1965)
Categor) I Class 1 ASME I, IX
?.
asal Valves Class.1 ASA 831.1 (1955) up to first 3
isolation valve and rest designed to ASA B31.1 (1955)
Qtit ilnment Spray L
Sy s t ens a
P ipirvj, Fittings, ASME III ASA B31.1 (1955)
Category I Class I ASME VIII is i
arul Valves Class 2 ASME VIII (1965) used for design purposes up to first valve outside containment STA!JDBY GAS 1REA'INENT ASME III GE SpecificationIII Category I Class I
'ASME III SYSTEM Class 2 Class 2 not required if post-g accident g
fission g
product vi receval is M
l not the y
I design basis
- w 6.
Specification was not provided for review.
a e
2 Table 4-1 (Cont.)
p b*
Quality Classification
"?
Codes aimi Codes arul Seismic Classification Structures, Systems, Starmlards Standards Used Used in arel Cougeonents HG 1.26 (IL in Plant Design (2)
HG 1.29 Plant Design Remarks
$,, w" COtarHo! HOD DHIVE ASME III ASME I (1965)
Category I class I 1 in IHAIS it4G Class I fQ CONTHOt HOD DHIVE ASME III
?
Category I
?
s2 SYST124 Class 2 3
SPtstrr FUE!. AHD t3EW ASME III
?
Category 1 Class I FUE!. S1UHAGE FACit.ITitS Class 3 MAIN STEAF4 SYSTIM Fipi n'J from Heactor ASME III ASME I, 11 (1965)
Category I Class I Vessel to First Class 1
,p Isolation Valve e
External to Drywell Pipise; f mosa Outermost ASME III ASA B31.1 (1955)
Category I
?
Conta i n wnt Ir.olation Class 2 Valve up to Turtsine
+
Stop Valve itsin Steam ASME III ASME I (1965)
Category I Class I j
Isolation Valve class 1 ASA B31.1 (1955)
GE Spec. 21AS467(6)
Safety Valve ASME III ASME I (1965)
Category I Class I See FSAR Class 1 Code Case 1271N Sec. IV-2-5 ASA B31.1 (1955) g N
Heliet Valve ASME III ASME I (1965)
Category 1 Class I Class 1 w
SlitrrtMUl4 COOLING N
S STIM 1
w H
Iteat ExctiatvJers (Tutie ASME III ASME III (1965)
Category I Class II
\\
Sidel Class 2 Class C
i Table 4-1 (Cont.)
of
{n
[?
Quality Classification Codes arxl Codes arul Seismic Classification Structures, Systems, Starklards Standards Used Used in J{
and C(anponents RG 1.26 H L in Plant Design (2[
RG 1.29 Plant IValqn Remarks 5
IIeat Exchangers (Shell ASME III ASME VIII (1965)
Category I Class II fg Side)
Class 2 X
Pumps ASME III ASA B31.1 (1955) III Category I Class II Pumps Claus 2 ASME I (1965) designed to ASME III Class Cs
>1nformation obtained from FSAR p. X-2-2 ey CO!JDr.raSA1H SluitAG6 SYSTDt Tank ASME III ASME I (1965)
Category II Class II Claus 3 Pump 3 ASM8 III ASH 3 I (1965)I4I Category II Class II Class 3 f.tOUID WASTE SYSTtM j
Tank ASME III API-650 (1964)*
7 Class II Class 3 Filter and Demineralizer ASME III ASME VIII (1965)
?
?
Cbeponents not vessels Class 3 given in H
original N
tables b
information M
obtained from $
l'SAR Sec. IX,
{
para. 3.1.4 w
H k
6
i l
21.
Table 4-1 (Cont.)
W*.
7 S
Quality Classification h
Codes aim!
Codes arul Seismic Classification g:o Structures, Systems, Standards Standards Used Used in 3
anel Comimnents RG 1.26 (IL in Plant Design (2)
RG 1.29 Plant Design Remarks 5 t*
hh REACTOR CLEA!iUP ASME III ASME III (1965)
Non-Seismic Class II j
gp DFMINERAI,1ZEll SYSTEM Class 1 Class C a3 ASA B31.1 (1955)
I O
CONDE!! SATE /FEEIMATER SYSTt21 Pumps ASME III ASA B31.1 (1955) I4I 7
Class II Class 2 ASME 1 (1965)
Feedwater Control Valves ASME III ASME I (1965)
Category I Class I
[
Class 1 8
lleat Excharvjets (Tube ASME III ASA B31.1 (1955) (4)
Category I Class I Side)
Class 2 ASME I (1965) lleat Exchargers (Shell ASME III ASA B31.1 (1955)I4I Category I Class I Side)
Class 3 ASME I (1965)
HEACD)H VESSEL llEAD SPRAY SYSTEM I
Pipirg, Fittings, ASME III ASME I (1965)
?
?
Components aimi Valves Class 2 ASA B31.1 (1955) not listed in original table, but given in FSAR
- p. X-2-2 1
0 0
STittlCnlHES m
.a He.sctor Buildity Category I Class I HAI3I 1
w>*
.El.
yg 16 as Table 4-1 (Cont.)
$N
'5 k
Quality Classification 3
Codes ard Codes and Seismic Classification O
Structures, Systems, Standards Standards Used Used in aimi Coenpone nt s f(G 1. 26 (1) in Plant Design (2)_
RG 1.29 Plant Design Remarks Drywell, Torus, Vents ASMS VIII (1962)
Category I Class I HA arx! Nuclear Code Cases Control Room Panels Category I Class I NA 4
Spent ruel Pool ASMS VIII Category I Class I HA veint Stack Non-Seismic Class II NA Category I (OBE)
Turt>1ne Buildir>J Class II NA Itadw.sste 15uilding Non-Seis=1c Class II NA Category I (ODE)
Intake ard Discharge Category 1 Class II NA Screen House Class II MA H
N
.A Y.
e 9
9
. _ _ =. _ _ _- _
=
I i
I TER-C5257-431
~
,r 1
1 Table 4-2(a) l Quality Group A Components (l)
Code:
ASME III-Class 1(2)
~.
1 Pressure Vessels
."d e None l
f Pipinq
~
j Recirculation System Piping (RCS)
ASME I (1965)
ASA B31.1 (1955)
Core Spray System Piping (CSS)
ASA B31.1 (1955)
~
q t
Automatic Depressurization System Piping (ADS)
ASME I (196,5)
Piping from Reactor Vessel to First ASME I (1965)
Isolation Valve External to Drywell (MSS)
ASME IX (1965) i 1
Control Rod Drive Housing (CRDH)
ASME I.(1965)l, r
ASA B31.1 (1955)' and Piping to Second Isolation Valve (MSS)
Nuclear Code, Cases (3)
~
Pumps
~ :.
Recirculation System Pumps (RCS)
ASA B31.1 (1955) (4) s s
Valves Recirculation System Valves (RCS)
ASME I (1965)'..
Core Spray System Valves (CSS)
ASA B31.1 (1955)
Automatic Depressurization System Valves (ADS)
ASME I (1965) 1.
Refer to Table 4-2(d) for abbreviations.
2.
ASME III-Class 1 stands for " Boiler and Pressure Vessel Code,"Section III, Division 1, Subsection NB, 1977 Edition and Addenda through Summer 1978.
3.
Specific nuclear code cases were not listed. This informstion is required.
4.
FRC notes that pump design is not covered under ASA B31.1 5.
Specification was not provided for review.
- 4J Frank!in Research Center
. won a.v..~.
w.
]
TER-C5257-431 l
Table 4-2 (a) (Cont.)
Valves (Cont.)
Code Main Steam Isolation valve (MSS)
ASME I (1965)
ASA B31.1 (1955) i GE Spec 21AS467(5)
Safety Valve (MSS)
ASMS I (1965)
ASA B31.1 (1955)
Code Case 1271N Relief Valve (MSS)
ASME I (1965)
Feedwater Control Valves (C/7WS)
ASME I (1965)
Storage Tanks (Atmospheric and 0-15 psig)
None 4
1 i
l 1
I i
. O AU Frank!!n Research Center a hea ce n, rwa w.:mn
e O
TER-C5257-4 31 Table 4-2 (b)
Quality Group B Componentdl)
Code:
ASME III-Class 2(2)
Pressure Vessels Code Emergency Systems Isolation Condenser -
ASME III (1965)
Tube Side (IC)
Class A Control Rod Drive System (CRDS)
?
ASME III (1965)
Shutdown Cooling System Heat Excnangers - Tube Side (SCS)
Class C Shutdown Cooling System ASME VIII (1965)
Heat Exchangers - Shell Side (SCS)
Condensate /Feedwater System ASME I (1965)
Heat Exchangers - Shell Side (C/FWS)
ASA 331.1 (1955) (3)
Piping Emergency Systems Isolation Condenser ASME I (1965)
Piping (IC)
Liquid Poison System Piping (LPS)
ASA B31.1 (1955) and Nuclear Code Cases (4)
Core Spray System Piping (Other Than ASA B31.1 (1955)
Class 1 Piping) (CSS)
Containment Spray System Piping (CS)
ASA B31.1 (1955) 1 ASME VIII (1965) l Standby Gas Treatment System Piping (SGTS)
GE Specification (5) l Piping from Outermost Containment Isolation ASA B31.1 (1955)
Valve up to Turbine Stop Valve (MSS) 1.
Ref er to Table 4-2(d) for abbreviations.
2.
ASME III-Class 2 stands for " Boiler and Pressure vessel Code,"Section III, Division 1, Subsection NC, 1977 Edition and Addenda through Summer 1973.
3.
FRC notes that heat exchanger design is not covered under ASME I or ASA B31.1.
4.
Specific nuclear code cases were not listed. This information is required.
5.
Specification was not provided for review. l
..rJ Franklin Research Center
+ a e N re.,*o.no.
l
o 4
TER-CS257-431 Table 4-2 (b) (Cont. )
Piping (Cont.)
Code Reactor Vessel Head Spray System Piping ASA B31.1 (1955)
(RVHSS)
ASME~I (1965) l Pumps Liquid Poison System Pumps (LPS)
ASME III (1965)
Class C Core Spray System Pumps (CSS)
ASA B31.1 (1955) (6) 1 Shutdown Cooling System Pumps (SCS) (7)
ASME I (1965)
ASA B31.1 (1955) (6)
Condensate /Feedwater System Pumps (C/ PAS)
ASME I (1965)
ASA B31.1 (1955) (6)
Valves Liquid Poison System Valves (LPS)
ASA B31.1 (1955) and Nuclear Code Cases (4)
Core Spray System Valves (Other Than ASA B31.1 (1955)
Class 1 Valves) (CSS)
Containment Spray System Valves (CS)
ASA B31.1 (1955)
ASME VIII (1965)
Standby Cas Treatment System Valves (SGTS)
GE Specification (5) l Reactor vessel Head Spray System Valves (RVHSS)
ASA B31.1 (1955)
ASME I (1965) l Storsce Tanks (Atmospheric and 0-15 psig)
I Liquid Poison Sysram Tank (LPS)
API-650 (1964)*
6.
FFC notes that pump design is not covered under ASME I or ASA B31.1.
7.
The plant FSAR, p. X-2-2, states that these pumps.ere designed to ASME III Class C.
Informatien assumed because it is not availar" it this time.
i l
A bbnk!!n Research Center
%
- w r n.m %
l
TER-C5257-4 31
~
Table 4-2 (c)
Quality Group C Componentdl)
Code:
ASHE III-Class 3 (2)
Pressure Vessels Code Emergency Systems Isolation Condenser -
ASME VIII (1965)
Shell Side (IC)
' Reactor Building Closed Loop ASME I (1965)
Cooling System Heat Exchanger (CLCS)
ASA B31.1 (1955) (3)
Filter and Demineralicer Vessels (LRS)
ASME VIII (1965)
Reactor Cleanup Demineraliser (RCDS)
ASME III (1965)
~
Class C Condensate /Feedwater System ASME I (1965)
Heat Exchangers - Tube Side (C/FWS)
ASA B31.1 (1955)(3)
Piping Reactor Building Closed Loop ASME I (1965)
Cooling System Piping (CLCS)
ASA B31.1 (1955)
Nuclear Code Cases (4)
Service Water System Piping up to First Isolation ASME I, IX (1965)
Valve (SWS)
Service Water System Piping (SWS)
ASA B31.1 (1955)
Reactor Cleanup Demineraliser System Piping (RCDS)
ASA B31.1 (1955)
Pumps Reactor Building Closed Loop ASA B31.1 (1955) (5)
Cooling System Pumps (CLCS)
ASME I (1965) 1.
Refer to Table 4-2(d) for abbreviations.
2.
ASME III-Class 3 stands for ' Boiler and Pressure Vessel Code,"Section III, Division 1, Subsection ND, 1977 Edition and Addenda through Summer 1978.
3.
FRC notes that heat exchanger design is not covered under ASME I or ASA B31.1.
4.
Specific nuclear code cases were not listed. This information is required.
5.
FRC notes that pump design is not covered under ASME I or ASA B31.1.
Information assumed because it is not available at this time.
A i) Franklin Research Center ronwws>.vmmesau.
TER-C5257-431 Table 4-2 (c) (Cont. )
Pumps (Cont.)
Code Condensate Pumps (C/PdS)
ASME I (1965)
Valves Reactor Building Closed ASME I (1965)
Loop Cooling System Valves (CLCS)
ASA B31.1 (1955)
Nuclear Code Cases (4)
Service Water System Valves (SWS)
ASME I, IX (1965)
ASA B31.1 (1955)
Reactor Cleanup Dcmineralizer System Valves (RCDS)
ASA B31.1 (1955)
Storage Tanks (Atmospheric and 0-15 psig)
Liquid Waste System Tank (IldS)
API-650 (1964)*
Condensate Storage Tank (C/PdS)
ASME I (1965) i l
l l
)
4.s !
5.'d Frank!in Research Center 4
4 % s N rem
[
=
4 i
TER-C5257-431 Table 4-2 (d)
Index of Abbreviations for Systems Abbrevation Definition ADS Automatic Depressurization System CLCS Reactor Building Closed Loop Cooling System CRDS Control Rod Drive System i
CRDH Control Rod Drive Housing i
CS Containment Spray System Core Spray System CSS C/FWS Condensate /Feedwater System IC Isolation Condenser LPS Liquid Poison System udS Liquid Waste System MSS Main Steam System RCS Recirculation System RCDS Reactor Cleanup Demineralizer System t
RVHSS Reactor Vessel Head Spray System SCS Shutdown Cooling System SGTS Standby Gas Treatment System SWS Service Water System
...cJ Frank!!n Research Center a cre a at Se F ma *.we
o TER-C5257-431 5.
EVALUATION OF SPECIFIC COMPONENTS 5.1 GENERAL REQUIRLMTNTS The purpose of this section is to evaluate, for the specific components of the Oyster Creek Nuclear Power Plant, how the general code requirements of the current code affect the safety margin to which these components were originally designed.
General code requirements are those requirements that apply to all the components discussed in this report (i.e., piping, pressure vessels, valves, pumps, and tanks). The following topics were identified in Section 4.1 of Appendix A to be general requirements that have changed from older codes to the current code:
fracture toughness, quality assurance,III quality group
~
classification, and code stress limits. They will be discussed herein.
5.1.1 Fracture Touchness As indicated in Section 4.1.1 of Appendix A, the current code (2] requires that pressure-retaining material be impact tested, but there are exemptions from this requirement. Tables A4-4 through A4-6, developed in Appendix A, are used as a guideline in evaluating whether it is necessary to impact test the material used for each specific component of the Oyster Creek Nuclear Power Plant. The results of this evaluation are compiled in Table 5-1.
Data on nil ductility transition temperature (TNDT) e feren ae as an be found in References 11,12, and 13.
Of the 40 components reviewed in Table 5-1:
o five components (12%) do not require impact testing o
the type of stainless steel used (:nost probably austenitic) was not specified for 2 components (5%)
o the material used was not specified for 31 components (78%)
o additional data are required to assess 2 components (5%).
1.
Quality assurance is outside the scope of the SEP according to the letter from S. Bajwa to S. Carfagno dated December 10, 1981.
, f 3 Franklin Research Center s cw w % -
l
Table 5-1 i
Review of Fracture Toughness Requirements f,
Oyster Creek Nuclear Power Plant 715
[
Structures, Systems, Quality Group Impact Test Reason for s 5' and Comp nents Classification Matertal Required?
Exempt ion (1)
Remarks i
$$?
1 HECIHCUI.ATION SYSTtM ia CR l
$7 Pumps -
l$2 Casi ng Class A Austenitic No 8e
- h Stainless Steel SA-351-CF8M Valves Class A Austenttic M) 8e Stainless Steel SA-351-CF8M Piping Class A Austenitic No 8e Stainless Steel I$
AS1M A358 I
EMERGENCY SYSTtMS IsoIation Condenser Tutie Side Class B Not given Not discussed in FSAR Stiell Side Class C tbt given Not discussed in FSAR Piping Class B Not given Not discussed in FSAR H
Liquid Poison Systera
- A Pumps Class B Stainless Steel Insufficient Probably m
Data Austenitic u
Stainless Steel Y
sw W
l.
Heter to Tables A4-4 through A4-6 of Appendix A for explanation of exemptions.
i
4 Table 5-1 (Cont.)
> 71 h
Structures, Systems, Quality Group Impact Test Reason for
.gm and Cornponents Classification
&terial itequi red?
Exenotionil)
Remarks
,0 fk PipiryJ, Fittihgs, Cla8s B Not given Not discussed in
'S and Valves A
<e
?,
E Tank Class D Not given Not discussed in l'SAR Core Spray System Pumps Class B Not given Not discussed in FSAR I
Pipirvj, Fittings, Class A Stainless Steel Insufficient Probably 8
and Valves Data Austenitic Stainless Steel PipirrJ, Fittings, Class B Carbon Steel Insufficient Sizes and steel an1 Valves Data type not,given Heactor Dullding Closed Imp Cailleri Syst era Pu.npa Claus C tiot given Not discussed in FSAR Heat Exchaskjers Class C Not given Not discussed in FSAR H
N Pipirry, Fi t t irvja, Class C Not given Not discussed in b
and Valves ESAR us Autornatic Depressur-Class A Not given Not discussed in
{
4 ization System FSAR w
V e
4
e Table 5-1 (Cont.)
l-E/
7 Structures, Systems, Quality Group Impact Test Reason for g$
asui Conqwsnents Cl a ssi t ica t.i.on.
Material Required?
Exempt ion (l)
Remarks 1L 3 po Service Water System J
5h Piptwj, Fittirvjs, arul Class C tiot given Not discussed in Va1es 9
FSAR ln Containment Spray System 4
Pipirvj, Fittinja, Class B Not given Not discussed in azul Valves FSAR STAT 3Ditt GAS TitEA'INEITF SYSTEM Class B Not given G8 Specification smt available for i
review N
un I
cot 3THOI. IM)D Di4IVE llOUSlHG Class A Not given Not discussed in FSAR COtmtOI ROD DRIVE SYSTtM Velocity I.imiter Class B 304 Stainless No 8e Steel Castitsj Isalex Tube Class B Stainless Steel No 8e Type 304 SPENT FUEI. At4D NtM Fut:I, S'It)l4 AGE FACII,ITIES Class C Not given Not discussed in g
FSAR Q
A STtW1 SYSTEM vi N
un Pipirmj from Reactor Class A Not given Not discussed in y
Vessel to First FSAR A
It. 'lation Valve External to Drywell
3 Table 5-1 (Cont.)
= nf/
Ui s
E Structures, Systems, Quality Group Impact Test Reason for and Cotaponents Classification Material Required?
Exemption (Q Remarks
/nh Piping f rosa Outeramaat Class B Not given Not discussed in h
Containment Isolation FSAR gg Valve up to Tuttaine 1g Stop Valve
]
Main Steara Class A Carbon Steel Insufficient Size not specified luolation V4lve SA216 Gr WCB data Safety valve Class A Not given Ibt discussed in FSAR Relief Valve Class A tut given Not discussed in Io FSAR m
l SlitlTlxMt1 COO!,ItJG SYST121 steat Exchangers -
Tulee Side class B Not. given Not discussed in FSAR Ileat Exchangers -
Shell Side Class B Not given Not discussed in
,s cot 4DEt4 SATE S'lOllAGE SYSTtM
-A Tank Class C Not given Not discussed in U
FSAR 1
6 Pumps Class C Not given Not discussed in y
FSAR D
I
I
!5',
,2 Table 5-1 (Cont'. )
VAh
,a SE a5 Structures, Systems, Quality Group Impact Test Reason for M
assi Omumie nt s Classification Matertal Required?_
Exemption (l[
Remarks
?$
LIQUID WASTE SYSTtM h
Tank Class C Not given Not discussed in 3
2 FSAR Filter asal Demineralizer Class C Not given Not discussed in i
l VesseIs FSAR REAC10R CIEANUP DOMINERALIZER SYSTtM e
Pipie r), Pittir>Ja, and Class C Not given Not discussed in U
Valves FSAR 4
CONDEttiATE/FEEIMATtR S YSTE}i Pumps Class B Not given Not discussed in FSAR Feedwater Control Valves Class A Not given Not discussed in FSAR Heat Exchar> jets (Tube and Classes B and C Not given Not discussed in Shell Sides)
FSAR 3
HEAC10R VESSEL q
HEAD SPRAY SYSTEM A
Pipti>J, Fit tings, and Class B Hot given Not discussed in u
Valves g
FSAR m
Y.
LJ V
a b
TER-C5257-431 5.1.2 Quality Assurance (
The quality assurance requirements for the design and construction of Class 1, Class 2, and Class 3 components as per the current code. [2] are outlined in Section 4.1.2 of Appendix A.
Most of these requirements were not considered in past codes [7, 9, 12).
Nevertheless, quality assurance was a concern in the Oyster Creek Nuclear Power Plant, as illustrated in the Final Safety Analysis Report (5).
5.1.3 Quality Grouc Classification As indicated in Section 4.1.3 of Appendix A under the title " Quality Group Classification," classification of components was not considered in the old piping code [6] or in the ASME B&PV Code,Section VIII, 1965 Edition (10).
The ASME B&PV Code,Section III, 1965 Edition (9) classified pressure vessels as Class A, B, or C.
Class A is equivalent to Class 1 of the current code [2].
Class B is concerned with containment vessels, which are outside the scope of this report. Class C may currently be classified as Class 2 or 3 of the current code.
Note in Table 4-2(b) that the shutdown cooling system heat exchangers -
tube side were designed according to Class C [9] requirements, whereas the shell side was designed to ASME B&PV Code,Section VIII (10).
The emergency systems isolation condenser - tube side was designed to Class A (10]
requirements.
In Table 4-2 (c), current Class 3 pressure vessels were constructed to Class C [9] or ALME B&PV Code,Section VIII (10) requirements, except for tne reactor building closed loop cooling system heat exchanger, which was designed to ASME I and ASA B31.1 requirements.
FRC notes that heat exchanger design is not covered under ASME I or ASA B31.1.
Class 2 pressure vessels constructed to Class C or Section VIEI requirements should be evaluated against current Class 2 requirements, especially for radiography requirements.
4 For pressure vessels designed to codes other than Section III or VIII (i.e.,
i 1.
Although discussed in this report, quality assurance is outside the scope of the SEP according to the letter from S. Bajwa to S. Carfagno dated December 10, 1981.
M
-IS-W;ij Franklin Research Center A Osamoa d N Framma moon,te
-m--+
---rr--
"A
- - - ^ -- -
2.
TER-C5257-431 reactor building closed loop cooling heat exchangers and control rod drive system), the Licensee should provide calculations and specifications in order to evaluate if they meet current requirements.
See discussion on full radiography requirements in Section 5.2 of this report.
5.1.4 Code Stress Limits Methods of calculating stress limits have changed in two major respects:
the use of different strength theories and the additional consideration of service levels C and D as possible loading conditions with different stress limits.
No mention of service levels C and D were found in the Oyster Creek FSAR [ 5 ]'. '
Although discussed in the previous paragraph, the seismic portion of this topic is outside the scope of this report. The seismic review of systems and components is performed by the NRC.
Design basec on the old piping code (7] and ASME B&PV Code,Section VIII (10] is more conservative, but less exact, than design based on the maximum shear stress theory of failure and stress limits given in the current code (2]
T for Class 1 components. The theory of failure used in ASME B&PV Code,Section III, 1965 Edition (9] for Class A pressure vessels is similar to that of the current code.
The current code for Class 2 and Class 3 components uses the same theory of f ailure as past codes.
l l
5.2 PRESSURE VESSELS As discussed in Appendix A, Section 4.3, major differences between current requirements (2) and old requirements (9, 10] for the construction of pressure vessels appear in four areas:
fracture toughness, quality group classifica-tion, design, and full radiography requirements.
Fracture toughness is discussed in Section 5.1.1 of this report. Quality group classification is discussed in Section 5.1.3.
The basic difference in design requirements concerns strees limits and consideration of service level I
- h M.' Franklin Research Center a w ce vven vnu.
l TER-C5257-431 C and D loading conditions. This topic is addressed in Section 5.1.4 of this report.
Full radiography requirements for pressure vessels are discussed in Section 4.3 of Appendix A.
The conclusion to be drawn frem this discussion is tha t, in general, past full radiography requirements for vessels were more conservative than current requirements, with the exception of Category C welds of vessels currently classified Class 2 which were designed to Class C (9] or Section VIII (10] requirements.
For this exception, the current full radiography requirements are more restrictive than past requirements.
Information regarding the radiography requirements imposed on the welds of the shutdown cooling system heat exenangers - tube side should be provided. This information should be compared with the current requirements given in Section 4.3 of Appendix A.
Class 1 vessels designed to any code other than ASME III (9] Class A requirements should be evaluated to determine if they comply with current fatigue analysis requirements. No Class 1 vessel was reviewed in this report.
5.3 PIPING In aadition to the general requirements previously discussed, the following items are considered when designing Class 1 piping for fatigue stresses based on the current code (2) that were not considered or were considered differently in the past ccde (7]:
o gross discontinuities in the piping systems are accounted for l
o loading due to the thermal gradient through the thickness of the pipe o indices used in calculating secondary stresses are equal to or less than twice the corresponding stress intensification factors in the past code.
The last two items pose no problem as far as the structural integrity of the system and are discussed in detail in Section 4.2 of Appendix A.
l When considering gross discontinuities of piping systems, two loading I
cases can prove to be potentially unconservative designs when evaluated to 4 W A Frankin Research Center i
s cm m avme,w,
TER-C5257-431 current code requirements.
Two examples are given in Section 4.2 of Appendix A in order to assess the potential problems of temperature loading for a large number of cycles and temperature loading for a medium range number of cycles.
These examples are based on Palisades (14] specifications. Stresses for both examples indicate that no problem exists.
From Table 4.2.1 of the FSAR [4], it can be seen that the thermal and loading cycles given for the Oyster Creek plant are similar to those given in the examples of Appendix A.
Data concerning the drop in temperature from 100%
power to 0% power were not given in the FSAR.
Assuming that the drop in temperature is 100*F (see Section 3.4.1.4 of NUREG-0123 (15]), the previous conclusion given in examples in Appendix A also applies to the Oyster Creek plant.
Piping designed according to Section I of the ASME B&PV Code (8] only should be evaluated in view of thermal stress and cyclic loading requirements as discussed in Section 4.2 of Appendix A.
Information regarding the thermal stress and cyclic loading imposed on the following piping systems should be provided: automatic depressuriration system piping, piping from the reactor vessel to the first isolation valve external to the drywell, control rod drive housing, emergency systems isolation condenser piping, and service water system piping up to the first isolation valve.
For Class 2 and Class 3 piping systems, the requirements of past and l
current codes are very similar.
Full radiography requirements for piping, valves, and pumps are discussed in Section 4.2 of Appendix A.
The conclusion to be drawn from this discussion is that, currently, full radiography is required for Class 1 and Class 2 welded Joints, whereas it was not required in the past code (7].
However, Provisions 2 and 3 of Code Case N-7(lI to Reference 6 required full radiography for circumferential and longitudinal welds.
If these provisions of the code case were applied, then current requirements are met.
Using Table 4-1, the Licensee should provide information indicating if provisions 2 and 3 of Code 1.
Mechanical Engineerir.g, August 1962.
4 MU Franklin Research Center
% atn.re.a e,
TER-C5257-431 i
Case N-7 were invoked, bearing in mind that this code case is only applicable to austenitic stainless steel. This information would indicate that current radiography requirements for Class 1 and 2 piping are met.
5.4 PUMPS The Licensee stated that Class 1 recirculation system pumps were designed to ASA B31.1 (7) as indicated in Table 4-2 (a).
However, it seems more likely that Section III of the ASME B&PV Code (9) would have been used for pump design. Table 4-2 (b) indicates that the liquid poison system pumps were designed according to Section III of the 1965 ASME Code (9]. The core spray system pumps, the shutdown cooling system pumps, and the condensate feedwater system pumps were designed according to ASA B31.1 [7] and/or ASME B&PV Code
~
Section I (8].
However, since pump design is not covered under these codes, 1
it seems more likely that Section III of ASME B&PV Code (9) would have been 4
used for pump design. This same comment applies to the reactor building closed loop cooling system pump and condensate pumps (Table 4-2(c)) because they were designed to ASME D&PV Code Section I (8] and/or ASA 331.1 (7].
4 Pumps designed to Section III should be checked against the requirements outlined in the Pressure vessel Section (Section 5.2).
Items to be reviewed regarding pumps are general requirements and full radiography requirements, discussed in Sections 5.1 and 5.2, respectively, of this report, and fatigue analysis discussed herein, i
Information on the radiography requirements imposed on the welds of the Class 1 and Class 2 pumps listed in Tables 4-2(a) and (b) should be provided
{
and ccmpared with the current requirements given in Section 4.2 of Appendix A.
The recirculation system pumps, currently classified as Class 1 pumps, were designed to the piping code ASA 331.1 [7); it is more likely that ASME j
S&PV Code Section III (9] was used for pump design. Class 1 requirements specify fatigue analysis if a set of conditions is not met (see MB-3222. 4 (d) l of Ref erence 2).
If any of the following conditions are not met, the recirculation system pumps should be analy:ed for cyclic loads:
1 i
.<fg33 %) Franklin Research Center A Dres,on of De FrerMn em
TER-C5257-431 (1)
Pressure Fluctuations:
the specified full range of pressure fluctuations during normal service does not exceeds (1/3) (Design Pressure) (S /S )
a m wheres.
S = alternating stress from f atigue curves corresponding to the a
number of pressure fluctuations S = allowable stress intensity at the service temperature m
(2)
Atmospheric to Service Pressure Cycle N2 j[ N(3S )
3 where:
N2 = the maximum number of atmospheric to service pressure cycles N(3Sm) = nummer of cycles from design fatigue curve for Sa " 3S m (3)
Temperature differences between adjacent points, i.e.,
two points along the meridian of a vessel, nozzle, or flange closer than 2 (Rt) (1/2) where R is the mean radius and t is the mean thickness between the two points:
i j[S /(2E3) (i = 1,2)
AT 3
where:
ATi = temperature differences between two adjacent points i = 1: Startup and shutdown i = 2: Normal service E = modulus of elasticity at mean temperature between points t
a = instantaneous coefficient of expansion, mean value (see Table I-5.0 of Reference 2)
S = alte:aating stress from design fatigue curve corresponding to 3
the number of startups and shutdowns, N, and the number of 1
significant temperature difference fluctuations during normal service, N. A significant number of temperature fluctuations 2
are greater than S/(2Ea) where S is the endurance limit, i.e.,
the value of S from the fatigue curve at 106 cycles.
a A Jd.) Frank lin Research Center a o-um w w. r.
a +=w,
TER-C5257-431 t
(4)
Temperature difference - dissimilar materials - see paragraph NS-3222. 4 (d) (4) of Reference 2 (5)
Mechanical loads - Stresses due to mechanical load fluctuations (excluding pressure) such as pipe loads on nozzles less than the value of S frem the design fatigue curve corresponding to the a
number of load fluctuations The Licensee should provide the following:
a.
proof that the five conditions previously outlined were met; therefore, analysis for cyclic loading is not required, or b.
if the five conditions were not met, calculations showing compliance with the current requirements for analysis for cyclic loading as described in Section NB-3222.4 of Reference 2.
Of the seven pumps reviewed in this report only the liquid poison system pumps were designed to ASM3 B&PV Code Section III, 1965 Edition (9].
The recirculation system pumps, the core spray system pu=ps, the shutdown cooling system pumps, the condensate /feedwater system pumps, and the reactor building closed loop cooling system pumps were designed to ASA B31.1 (7].
The Licensee indicated that the condensate pumps, the shutdcwn cooling system pumps, the condensate /feedwater system pumps, and the reactor building closed loop cooling system pumps were designed to ASME B&PV Section I [8].
However, since pump design is not covered under ASA B31.1 (7] or ASME Section I (8], it is more likely that ASME B&PV Code Section III [9] was used for pump design.
Correct information and calculaticns should be provided for the previously mentioned pumps in order to evaluate if they meet current requirements.
5.5 VALVES Major differences between current requirements (2) and past requirements (6) for valves are discussed in Section 4.5 of Appendix A.
Class 1 valves designed in accordance with past requirements should be adequate when judged by current standards except for:
1.
fracture tougnness requirements 2.
stress limits might not be satisfied for valves that differ significantly from the body shapes described in the current code
-eTQg3 '
DJ Frank!in Research Center A cnma a w. r- %.
TER-CS257-431 3.
stress limits for service level C might not be satisfied 4.
full radiography requirements (Class 1 and Class 2).
The following recommendations should be followed in order to evaluate the adequacy of Class 1 valves (see Table 4-2(a)) in the Oyster Creek plant:
1.
See Table 5-1 for the fracture toughness requirements evaluation.
2.
Compare actual body shape of valves with the body shape rules of Section NB-3544 (2].
If the body shapes are significantly different,the Licensee should provide calculations based on alternative rules in order to prove the adequacy of the valve.
3.
Show that the valve has been subjected to service level C conditions and no replacement is necessary.
If this is true, item 2 need not be investigated.
The following recommendation should be followed in order to evaluate Class 2 and 3 valves:
The pressure-temperature rating of Class 2 and 3 valves in the Oyster Creek plant (see Tables 4-2 (b) and 4-2 (c)) should be compared with current pressure-temperature ratings [16].
Full radiography requirements for piping, valves, and pumps are discussed in Section 4.2 of Appendix A.
The conclusion to be drawn from this discussion is that, currently, f'ull radiography is required for Class 1 and Class 2 welded Joints, wnereas it was not required in the past code (7].
However, Provisions l
2 and 3 of Code Case N-7 (1) to Reference 6 required full radiography for i
circumferential and longitedinal welds.
Using Table 4-1, the Licensee should l
provice information indicsting wr.cther Provisions 2 and 3 of Code Case N-7 l
were invoked, bearing in mind that this code case applies only to austenitic l
stainless steel.
If these provisions of the code case were applied, then current requirements are met.
The safety valves were designed to Section I (8] with Code Case 1271N invoked. Code Case 1271N considers the operational requirements only, and Section I (8] requires full radiography for circumferen-tial and longitudinal welds.
The Licensee should confirm that full radiography requirements were enforced for the safety valves, i
l 1.
Mechanical Eng ineering, August 1962.
r l
.eggg3 l JCL2 Franklin Research Center s % d % nanon neue
i TER-C5257-431 5.6 STORAGE TANKS As discussed in Section 4.7 of Appendix A, atmospheric storage tanks designed to the 1965 Edition of ASHE B&PV Code,Section III Class C or Section VIII, should be checked to determine whether the current compressive stress requirements are met.
Class C atmospheric storage tanks currently classified as Class 2 should be checked against current quality assurance requirements.
As also discussed in Section 4.7 of Appendix A, O to 15 psig storage tanks designed to Class C requirements may not satisfy current tensile allowables for the biaxial stress field.
Zero to 15 psig Class C storage tanks currently classified as Class 2 may not satisfy current quality assurance requirements.I Storage tanks designed to the American Petroleum Institute API-650,1964 Edition (17] should be investigated to determine if they meet current requirements.
The liquid poison system tank, liquid waste system tank, and condensate storage tank are reviewed in this report.
The liquid poison system tank and the liquid waste system tank were designed to API-650.
Stress allowables for the tank walls were lowered in API-650 (1964) when compared to current 1
standards. Stress allowables for the roof satisfy current standards. The use of A-7 plate material permitted by API-650 is no longer accepted by the current code.
The Licensee indicated that the condensate storage tank was designed to ASME B&PV Code Section I (8]; however, tank design is not covered in ASME B&PV Code Section I.
Calculations and specifications on the three previously mentioned tanks should be provided in order to see if they meet current standards.
1.
Although discussed in this report, quality assurance is outside the scope of the SEP according to the letter f rem S. Bajwa to S. Carf agno dated Decemoer 10, 1981.
44 ! ped Frank!!n Research Center A Dwaan of The Frarwan ism;n4e
TER-CS257-431 6.
CONCLUSIONS AND RECCMMENDATIONS A comparison of the standards in effect during the design and construction of the Oyster Creek Nuclear Power Plant against current standards indicates differences in the following areas:
fracture toughness requirements, quality assurance requirements,
' quality group classification, code stress limits, full radiography requirements, and fatigue analysis of Class 1 piping systems, pressure vessels, and pumps.
Although the requirements for code stress liraits and fatigue analysis of piping systems have changed throughout the historical development of the current code, the changes in these areas have not significantly affected the
~
safety functions of the systems and components reviewed in this report (see Item 8 in summary of conclusions).
In many instances, information provided by the Licensee in the table of quality group classification of systems and components and the FSAR is not sufficient or clear enough to enable FRC to make a proper judgment on the safety functions of the systems and components reviewed in this report.
Recommendations are given in Section 5 of this report with regard to the necessity for additional information to permit an adequate assessment of the impact of the new or changed requirements of the current code (2] on the safety functions of the systems and components reviewed in this report.
A summary of conclusions and reccmmendations is as follows:
i 1.
Fracture toughness - a total of 40 components were reviewed in Table 5-1 to determine if impact testing was required. From the information in this table, it was found that 12% of the components do not require impact testing, 5% of the components require confirmation i
that austenitic stainless steel was the material used, 78% of the l
components did not specify the material used, and 5% of the components require more data in order to be assessed. The missing l
inf ormation should be provided by the Licensee and, using Tables A4-4 l
througn A4-6 in Appendix A, an evaluation should be made for each l
component to determine whether impact testing is required.
I l
1.
Although discussed in tnis report, quality assurance is outside the scope I
of the SEP according to the letter f rom S. Ba]wa to S. Carfagno dated l
December 10, 1981.
1
.<TQg3 M Franidin Research Center Ac-e.% =
TER-C5257-431 2.
Full radiography requirements - information should be provided regarding the radiography requirements implemented for (i) Class 2 vessels, (ii) Class 1 and 2 piping and valves, and (iii) Class 1 and 2 pumps.
Indicate if Code Case N-7 of B31.1 was invoked for piping and valves.
Piping design to Section I of ASMS B&PV Code (8) complies with current full radiography requirements. Vessels and pumps designed to Class A requirements (9) and current Class 3 piping, pumps, and valves meet current full radiography requirements.
Tables 4-2(a), 4-2(b), and 4-2(c) chould be used in providing the required information.
3.
Quality group classification - Class A (9) vessels are equivalent to current Class 1 vessels. Class C vessels may currently be classified as Class 2 or 3.
In the Oyster Creek Nuclear Power Plant, the shutdown cooling system heat exchangers - tube side currently classified as Class 2 vessels were designed to Class C requirements and the shell side currently classified as Class 2 vessels were designed to Section VIII requirements.
Radiography requirements imposed on the welds of the heat exchangers should be provided and compared to current Class 2 requirements.
4.
Piping - in addition to impact testing and full radiography requirements, piping systems designed only to Section I of the ASME
~
B&PV Code (8] requirements should be investigated to determine how the thermal stress and cyclic loading requirements were accounted for.
The following piping systems should be investigated as discussed in Section 4.2 of Appendix A:
automatic depressurization system piping, piping from the reactor vessel to the first isolation valve external to the drywell, control rod drive housing, emergency system isolation condenser piping, and service water system piping up to the first isolation valve.
5.
Valves - in addition to the impact testing and full radiography requirements previously discussed, information should be provided by the Licensee, on a sample basis, regarding the design of valves in order to determine if they meet current body shape and pressure-temperature rating requirements as discussed in Section 5.5 of this report. Valves designed only to Section I (8) should be checked to determine whether thermal stress and cyclic loading requirements were l
accounted for.
l 6.
Pumps - seven pumps were reviewed in this report.
The Licensee has indicated that all pumps except the liquid poison system pumps were designed according to ASA B31.1 (7] and ASME 3&PV Code Section I (8).
However, since pump design is not covered under these codes, it seems more likely that Section III of ASME B&PV Code [9] would have been used for pump design. Correct information, calculations, and specifications should be provided for these pumps in order to evaluate if they meet current standards.
For the recirculation f
- db Franklin Research Center A cm a n rean mwe
TER-C5257-431 system pumps (Class 1 pumps), proof of compliance with current fatigue analysis requirements should be provided (see discussion in Section 5.4 of this report).
7.
Storage tanks - (1) atmospheric storage tanks should be checked to verify that they meet current compressive stress requirements; (ii)
O to 15 psig storage tanks should be checked to verify that they meet current tensile allowables for biaxial stress field condition; (iii) storage tanks designed to API-650 (1964) (17] (the liquid poison system tank and the liquid waste system tank), should be investigated to determine whether they meet current stress allowables and materials standards. The Licensee indicated that the condensate tank is designed according to ASME B&PV Code Section I (7) ; however, tank design is not covered under Section I.
Calculations and specifica-tions for the three storage tanks discussed in this report should be provided in order to see if they meet current standards.
8.
Missing information - (1) information missing from Tables 4-2(a),
4-2(b), and 4-2 (c) of this report regarding the code, code class, and code cases used in designing 9 of 49 components should be provided; (ii) assumptions on code editions that were made in order to complete Table 4-1 should be confirmed; (iii) the assumption made on the temperature drop (100*F) from 100% power to 0% power should be confirmed; (iv) specifications for the main steam isolation valve and the standby gas treatment system should be provided.
9.
Clarification of information - Table 6-1 identifies (1) discrepancies between information supplied by the Licensee and information in applicable codes and standards, (2) necessary information that has not yet been supplied by the Licensee, and (3) assumptions made by FRC wnere information was not supplied by the Licensee. The Licensee should address all items listed in Table 6-1, clarify all discrepancies, and supply all the missing information.
Zh..J FranMn Research Center ao~
o.r= %,
1 jfg Table 6-1
<=.
~
Items Requirinj Clarification for g
Oyster Creek 2:oclear Power Plant a 5' 2.7 q
Quality Clasolfication fy Stated by I.t censee
=_ :r Codes and Co.les and Codes and fQ Structures, Systen.a, Standardu Staridards Used Standards Commonly
= ?.
and Componento HG 1.26 (l[
in Plant Design (2)
Used in Plant Design (2)
Remarks REctRcu!Arrota SYSTrial Puups ASME III ASA B31.1 (1955)
ASME III Pu.ap design not Claus 1 Class A or C covered in piping code. I.icensee to indicate the appil-I cable code class used I
for plant design.
e EHEltGEtJCY SYSTDt1 I,lould l'oissan Syst em Pigal twj, Fi t tirej a, and ASME III ASA B31.1 (1955)
ASA B31.1 (1955)
Licensee to indicate Valves Class 2 and N.2 clear Code and Nuclear Code appropriate code Cases Cases cases used for plant design.
Tank ASME III
?
API-650 (1964)
Information assumed Class 2 because code for plant i
g design was not g
specified, un 1.
DJ ASME III stands for "Boller and Pressure Vessel Cole,"Section III, Division 1, published by the American un Society of Mechanical Engineers,
-4 1977 Edition with Addenda through Susmaer 1978.
1, 1965 Edition was assumed when plant was designed in accordance with Sections I, 2.
III, and VIII of ASME Doller and y
Preuaure Vessel Code.
Provide confArmation for this assuinption.
l l
e
e g
C. *. 3 Table 6-1 (Cont.)
In as m
Giality Classification g8 Stated by Licensee f$
Codes avst Codes aski Codes and 39 Structures, Systems, Standards Starklards Used Standards Commonly p
and Components RG 1.26 (R in Plant Design (2)
Used in Plant Design (2)
Remarks
-a4 Core Spray System Pumps ASME III ASA B31.1 (1955)
ASME III Pump design not Class 2 Class A or C covered in piping code.
Licensee to indicate the app 11-cable code class used for plant design.
4 Y
Heactor Buildlin Closed to p Cooling System Pumps ASME III ASME I (1965)
ASME III (1965)
Pump design not Class 3 ASA B31.1 (1955)
Class A or C covered under ASME I or piping code.
Licensee to indicate the applicable code class used for plant design.
Ileat ExchanJers ASME III ASME I (1965)
ASME III (1965)
Same as above.
Class 3 ASA B31.1 (1955)
Class A or C P ip t ivj, Fittings, and ASME III ASME I (1965)
ASME I (1965)
Nuclear code cases used tg 8
Valves Class 3 ASA B31.1 (1955)
ASA B31.1 (1955) in plant design should and Nuclear Code be specified.
m cases y
4 B
6
Table 6-1 (Cont.)
p ti
- Quality Classification D ~n Stated by Licensee h#
Codes and Codes anx!
Codes and 5E Structures, Systems, Standards Standards used Standards Commonly N
an 1 Compone nt a ItG 1.26 (11 in Plant Design (2)_
Used in Plant Design (2)
Remarks U
Q STAtml1Y GAS THEA'It4Ettr ASME III GE Specification
?
Furnish GE specifi-
{ f}
SYSTtti Class 2 cation for design a
fn review and indicate
}
other design codes it used.
Cot 4THOI. Iwo DRIVE ASME III
?
?
Indicate the codes SYSTtM Class 2 used in plant design.
St Ettr FUEL. AfdD NEW FUEL ASME III
?
?
Indicate the codes 4-S1UHAGE FACII.ITIES Class 3 used in plant design.
tJ l
MA!!4 STEAtt SYSTtH Hain Steora Isolation ASME III ASME I (1965)
ASME I (1965)
Furnish GE specifi-valve Class 1 ASA B31.1 (1955)
ASA B31.1 (1955) cation for design GE Spec 21A5467 review and indicate other design codes used.
S11UTtXMt1 COOI.Attr SYSTEM Pumps ASME III ASA B31.1 (1955)
ASitE III Pump design is not Class 2 ASME I (1965)
Claus C covered under ASME I g
or piping code. The g
plant FSAR, p. X-2-2, g
states that these vi pumps were designed y
to ASME III Class C.
y Licensee should indicate the appli-N cable code class used for plant design.
e
!?
li' dq a
E as Table 6-1 (cont.)
$N I:
j$
Qua'11ty Classification 7
Stated by Iicensee h
Codes and Codes and Codes and k
Structures, Systems, Standards Standards Used Standards Commonly awl Comtw>nents RG 1.26 (1) in Plant Destyn (2)
Used in Plant Design (2)_
Remarks CONDE!JSATE STURAGE SYSTEM Pumps ASME III ASME I (1965)
ASP ~ YII Pump design is not Class 3 Cl covered under ASME I.
l b
1.lOUID WASTE SYSTEM i
Tank ASME III
?
API-650 (1964)
Information assumed Class 3 because code for plant design was not specifled.
CONDENSATE /FEEIHATER SYSTIM
,r Ptunpa ASME III ASA B31.1 (1955)
ASME III Pump design is not Class 3 ASME I (1965)
Class C covered under ASilE I e
or piping code.
fleat exchanger design Heat Exchanger ASME III ASA B31.1 (1955)
ASME III g
f Classes 2 & 3 ASME I (1965)
Class C not coseretl under
+
AEME I cr piping code.
f w
i R
y-1
~.
La 3
't
/
4 f
j 1
a. >
./
4:
e
TER-C5257-431 7.
REFERENCES 1.
" Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waate-Containing Components of Nuclear Power Plants,"
Rev. 3 NRC, February 1976 Regulatory Guide 1.26 l
2.
ASME Boiler and Pressure Vessel Code,Section III, Division 1 1977 Edition and Addenda through Summer 1973 3.
Title 10 of the Code of Federal Regulations, Section 50.55a
" Codes and Standards" Revised January 1,1981
- 4., Codes and Standards for Category 1 Structures (provided by NRC in original package); Letter from Yoshito Nagai (GPU) to Alan Wang (NRC) dated November 30, 1981; Telephone Memorandum 5. Tikoo (FRC) to A. Wang (NRC) on January 12, 1982; Telephone Memorandum, A. Gonzalez (FRC) to A. Wang (NRC) dated January 21, 1982 5.
Final Safety Analysis Report for Cyster Creek Nuclear Power Plant (2 volumes)
Jersey Central Power and Light Company Docketed USAEC, Docket No. 50-219 6.
Standard Review Plan, Section 3.2.2, " System Quality Group Classification" NRC, Office of Nuclear Reactor Regulation NUREG-75/087 7.
" Code for Pressure Piping" American Society of Mechanical Engineers,1955 ASA B31.1-1955 S.
Boiler and Pressure Vessel Code Section I,
" Rules for Construction of Power Boilers" American Society of Mechanical Engineers, 1965 9.
Boiler and Pressure Vessel Code Section III, " Rules for Construction of Nuclear Vessels" American Society of Mechanical Engineers, 1965 4 !dd') Franklin Research Center
% e n. % %
w
TER-C5257-431 10.
Boiler and Pressure Vessel Code Section VIII, "Unfired Pressure Vessels" American Society of Mechanical Engineers, 1965 11.
R. P. Snaider, J. M. Hodge, H. A. Levin, and J. J. Zudans
" Potential for Low Fracture Toughness and Lamellar Tearing on PRR Steam Generator and Reactor Coolant Pump Supports" Published for Comment, October 1979 NUREG-0577 12.
" Nuclear Pressure Vessel Steel Data Base" Electric Power Research Institute Prepared by Fracture Control Corporation, December 1978 NP-933, Research Project 886-1 13.
" Metals Handbook" Metals Park, Ohio: American Society for Metals, Ninth Edition,1979 14.
Final Safety Analysis Report for Consumers Power Company Palisades Plant (3 Volumes)
November 5, 1968, Docketed USAEC, Docket No. 50-255 15.
" Standard Technical Specifications for General Electric Boiling Water Reactors," Rev. 3 NRC, Of fice of Nuclear Reactor Regulation,1980 NUREG-0123 16.
" Steel valves" American Society of Mechanical Engineers,1977 ANSI B16.34-1977 17.
" Welded Steel Tanks for 011 Storage" American Petroleum Institute, Second Edition, April 1964 API-650 h MJ FranMn Research Center 4%onn.r- %
l J
I I
I i
APPENDIX A REVIEW OF CODES M1D STANDARDS APPLICABLE % OYSTER CREEK, MILLSZNE, GINNA, AND DRESDOi PLMITS t
I I
I h
A n
JL.. Franklin Research Center A Division of The Fran'4fn insttute n s.nen re.r.e.n p.n...yw.
p t sica a m ua.ioce
CONTENTS 4
Section Title Page 1
INTRODUCTION A-1 2
SUMMARY
OF RESULTS OF CODE COMPARISON A-3 2.1 General A-3 2.2 Piping.
A-3 2.3 Pressure Vessels A-3 J
2.4 Pumps.
A-28 2.5 Valves.
A-28 2.6 Heat Exchangers A-28 2.7 Storage Tanks.
A-28 3
CONCLUSIONS AND RECOMMENDATIONS.
A-29 4
CCMPARISON OF SIGNIFICANT CURRENT CODE REQUIREMENTS AND PAST REQUIREMENTS A-30 t
4.1 General Requirements A-30 4.2 Piping.
A-61 4.3 Pressure vessels A-80 4.4 Pumps.
A-84 4.5 Valves.
A-84 4.6 Heat Exchangers A-90 4.7 Storage Tanks.
A-92 5
BASIS FOR SELECTING REQUIREMENTS MCST SIGNIFICANT TO COMPCNENT INTEGRITY.
A-97 6
REFERENCES.
A-98 l
l iii
- '.L Frank
- in Research Center A Cows.on of N Frarwen we
"I 1.
INTRODUCTION The purpose of this appendix is to compare the code currently used in the design, f aorication, erection, and testing of systems and components for nuclear power plants against the codes and standards used in the design of plants being reviewed under the Systematic Evaluation Program (SEP). The current code is the American Society of Mechanical Engineers' Boiler and Pressure Vessel Code (B&PV),Section III,1977 Edition as supplemented by the Summer 1978 Addenda [1, 2].
The three major older codes being compared again st the current code are the B&PV Code,Section III, 1965 Edition [3]; the
" Code for Pressure Piping," American Standard Association B31.1,1955 Edition
[4]; and B&PV Code,Section VIII,1965 Edition [5].
Taole Al-1 groups the SEP plants according to the ma]or codes used to
~
design them.
In order to take advantage of the similarities in each group, this appendix applies only to the Group I plants:
Palisade s, Ginna, Millstone Unit 1, Dre sden Unit 2, and Oyster Creek.
The B&PV Code,Section I, 1965 Edition [5] is also discussed in this appendix at it applies to Oyster Creek, Millstene Unit 1, and Dresden Unit 2.
t The older requirements are evaluated to identify differences from the current code requirements and to assess the impact of these differences on the structural integrity of the systems and components. The current code require-ments are discussed in Section 2.
The mujor identified differences are discussed in Section 4.
The acope of this comparison is limited to quality classification of systems and components as discussed in Regulatory Guide 1.26 [6] and Section 3.2.2 of the Standard Review Plan [7].
The reactor vessel, steam ger.erator s, and supports are outside the scope of this appendix, as is the seismic classification of systems and components. All these subjects are addressed in other SEP topics. Quality assurance has also been determined to be outside the scope of this comparison, but has been included for informational purposes only.
1.
Letter from S. Ba]wa to S. Carf agno dated Decem er 10, 1981.
A-1 g
...J Franklin Research Center swmum,w,
Table Al-1 Major Codes and Standards Used in Design of Systems and Components of SEP Plants Commercial Plant Operation Major Codes Group I (1969-1971)
Palisades Dec. 1971 1.
ASME III (1955)
)
Millstone 1 March 1971 2.
ASA B31.1 (1955) and Code Cases Ginna July 1970 3.
ASME VIII (1965) and Code Cases Dresden 2 July 1970 4.
ASME I (1965)
(Oyster Creek, Millstone 1, Dresden 2)
Oyster Creek Dec. 1969 Grouc II (1968) 1 Lacrosse Nov. 1969 1.
ASME I & VIII (1962) i l
and Code Cases 1
l San Cnofre Jan. 1960 2.
ASA B31.1 (1955) and Code Cases Hacdam Neck Jan. 1968 l
Group III (1961-1963) i Big Rock Point March 1963 1.
ASME I & VIII (1959) f and Code Cases 2.
ASA B31.1 (1955) and Code Cases Yankee Rowe July 1961 1.
ASME I & VIII (1956) and Code Cases 2.
ASA B31.1 (1955) and Code Cases L
1 A'
i;Ja Franklin Research Center 4 cum.ca w n. Fwa irsist.
2.
SUMMARY
OF RESULTS OF CODE CCMPARISON 2.1 GENERAL The current coce requirements for the construction of nuclear power plant components [1] are outlined in Table A2-1.
For each article or subarticle, the applicaollity to Code Class 1, 2, or 3, corresponding to Quality Class A, B, or C, respectively, is noted.
Requirements considered especially signifi-cant from the viewpoint of pressure boundary integrity are indicated by an "A" in the "Significant" column. The basis for selecting significant items is discussed in Section 5 of this appendix.
2.2 PIPING Table A2-2 presents a comparison of the current and past code require-ments for the materials, design, fabrication, examination, and testing of piping systems and components for nuclear power plants. The past code for piping is the B31.1 (1955) power piping code. The ASME I (1965) (5) power boiler code may have been invoked for piping between the BWR vessel and the first set of shutoff and check valves in the line. A comparison of signifi-cant past and current piping requirements may be found in Sections 4.1 and 4.2 of this appendix.
2.3 PRESSURE VESSELS Taoles A2-3 and A2-4 compare tne current and past code requirements for the materials, design, fabrication, examination, and testing of pressure vessels for nuclear power plants. Taole A2-3 compares the current code against ASME III (1965).
Taole A2-4 compares the current code against ASME VIII (1965).
Note that past Class A vessels were built in accordance with ASME III (1965), which would be equivalent to the current Class 1 classification.
Past Class B vessels were defined as containment vessels, which are outside the scope of this review.
l l
l
^~
A d Franklin Research Center a w o. r..waman
Table A2-1 Current code Requirements [l[
LEl 3 (~ ) /
Article
- 3' or Class Class Class Sign!-
Sut>a r t icle Description 1
2 3
ficant Rossark s a5 g
NA-1000 SCOPE OF St.CTIOta III A
A A
g R
NA-2000 Ct.hSSIFICATION OF COMPONENTS A
A A
A kQ NA-3000 RES&ONSIBILITIES AND DUTIES A
A A
Ea 4
NA-4000 QUALITY ASSURANCE NA-4100 Quality Assurance Requirements A
A NA A
I NA-5000 INSPECTION t4A-5100 General Requirements for Authorized A
A A
A Inacection Agencies and Inupectors NA-5200 Duties of Ing>cctor s A
A A
A y
IIA-6000 QUALITY CONT 1uE SYSTEMS FOR C1. ASS 3 A
OUNSTRUCTIOtt NA-6100 General kequirements NA NA A
A 13A-6200 0 ganization and Req >onsibilities HA NA A
A NA-6300 Control of Operations HA NA A
A HA-6400 Records and Forms t;A NA A
A i
NA-8000 CERTIFICATES OF AU1110RIZATION, A
A A
I NAMEPI.ATES, STAMPING, AND REPORTS 1000 INTRODUCTION 1100 Scope A
A A
A 2000 MATERIAL 2100 General A
A A
A 2200 Material Teet Coupons and Specimens A A
A for Ferritic Steel Materials A Addressed in the Code for the q>ecified class or considered significant for this review.
l
- Not considered significant for this review.
l 0 Outside the scope of this review.
NA Not applicable to this review or not addressed in the Code for the specified class.
Article nuuler in current Code will be preceded by Ha for Class I cos.ponent, NC for Class 2 component, and ND for Class 3 comp >nent.
e 4
m/
Table A2-1 (Cont.)
{E Article
- O or
/
Class Class Class Signi-Subarticle De wr ipt ion 1
2 3
ficant Henarks o
g Ji 2300 Fracture 1bughness Hequirement A
A A
A for m terial g
2400 Welding and BrazirvJ A
A A
A g
2500 Examination arxl Hepair of Pressure A
A A
Hetaininj m terials 2600 mterial Nnuf acturers' Quality A
A A
A System Program 2700 Dimensional Starxiard A
A A
3000 Dt: SIGN 3100 General A
A A
A p
3200 Design by Analysis (C1. 1); Alternate A A
NA A
t Design kulea for Veusela (C1. 2) 3300 Vesa,el Design A
A A
A 3400 Pump Design A
A A
A 3500 Valve Design A
A A
A 3600 Pipir>J Design A
A A
A 3700 Electrical and Mechanical Penetration NA A
A A
Assemblie s 3800 Design ot Atmoq>heric Storage Tanks NA A
A A
3900 0-15 pai (0-103 kPa) Sturage Tank NA A
A A
De sign 4000 FABRICATION AND INSTALLATION 4100 General A
A A
4200 Forming, Fitting, armj Aligning A
A A
4300 Welding Qualitications A
A A
A 4400 Hules Governing mking, Exas,inirvj, A
A A
and Hepairing Welds 4500 brazing A
A A
4b00 Heat Treatment A
A A
l
. ~..-
. =..-
i c2:lS c:.
Dy h Table A2-1 (Cont.)
fB G
Article
- E" or Class Class Class Signi-Suharticle De ucr ipt ion 1
2 3
ficant Remarks 4700 Mechanical Joints A
A A
4800 Expansion Joints NA A
A ag 5000 EXAMINATION 5100 General Requirennents A
A A
A 4
5200 Requirel Exauf nation of Welds A
A A
A (C1. 1) Examination of Welds (Cl. 2 and C1. 3) 5300 Acceptance Standard A
A A
A 5400 Final Examination of items (Cl. 113 A
NA A
A Sg>L Examination of Welded Joints (C1. 3) e
$500 Qualifications of Ikandestructive A
A A
A Emant,.ation Per sonnel 5600 NA HA NA 5700 Examination Requirement of NA A
A E4t>ansion Jointo 6000 TESTING 6100 General A
A A
6200 Hydrostatic A
A A
6300 eneumatic A
A A
6400 Psensage Test Gages A
A A
6',00 Atm.2wheric and 0-15 palg NA A
A Stosage Tanks 6600 laydtomtatic Testing of Vessels NA A
NA Designed to NC-3200 6700 Psieumatic Testing of Vesuels HA A
NA Designed to NC-3200 6800 6900 Psoof Tests to EstabliaJa NA A
A Dusign Pressure N
e 5
t
- r, IN:'h Mh/
Table A2-1 (Cont.)
E Atticle*
or class class Class Signi-J j@
Sut> article De nicr it>t ion 1
2 3
ficant Remarks
{R
)
7000 PROTECTIDH AGAINST OVEhPRESSURE g
1100 General A
A A
r}
7200 Definitions Applicable to A
A A
Overpressure Protection Devices 7300 Overpressure Protection Report A
A NA (C1. 1): Analysis (C1. 2) 7400 Relieving Capacity Requirements A
A A
and Acceptable Types of Overpressure Protection Devices 7500 Set Pressures of Pressure Relief A
A A
Device s p
I 7600 Operating Design Requirements for A
A A
Pressure Relief Valves 7700 Requirements for Nonteclosing A
A A
Pressure Relief Devices 780C Certitication Requirements A
A A
7900 H.st king, Starspisvj, and Reports A
A A
8000 NAMEPLATES, STAMPING, AND REPORTS 8100 General A
A A
MANDA10RY APPENDICES I
Design Stress Intensity Values, A
A A
A Allowable Stresses, Material Properties, and Design Yatigue Curves II Experimental Stress Analysis A
A A
III Basis for EstablirJaisuj De sign A
A A
A Stress Intensity Values and Allowable Stress Values 4
o
!s f>
Table A2-1 (Cont.)
e,3 Article
- f}k or Class Class Class Signi-Subarticle De a:r ipt t on 1
2 3
fIcant Remarka n
IV Approval of New Haterlata Under A
A A
the ASt4Y. Doller and Pressure Vessel Code for Section III A!> plication V
Cer tificate Holder's Data Ruport A
A A
Forsas and Application Forms for
]
Certificates of Autt.orization for Uans of Code Symbol Staintss VI Wusmied Indications Char ts A
A A
l VII Charts for Determinizaj Shell A
A A
Thickneus of Cylindrical and 1
Sphatical Cosaponents Under External Pressure Rules for blted Flanje NA A
A XI Connections for Clasu 2 and 3 Cougonents and Class HC Vassels XII Design Considerations for Dulted A
A A
A Flaaje Connections XIII Design Baa.ed on Stress Analysis NA A
NA for Vesacis Designed in Accosdance wit.h IC-3200 XIV Domign Based on Fatigue Analysis HA A
NA for Vouselm Designed in Accordance with NC-3200 XVI Nudentructive Examination A
A O
Hethods Applicable to Core l
Sunort Structures XVII Design of Linear Type Supports by A
A A
O Linear Elastic and Plastic Analysis e
TJ lL'
"? Y (E
I a3 2a
~n j an g3:.r 0
Table A2-1 (Cont.)
Article
- or Class Class Class Sign!-
S:staa r t ic'a e Deseription 1
2 3
fIcant Remarke NONMANDATORY APPENDICES A
A NA NA B
Nner's Design Specification A
A A
C Certificate Holder's Stress Report A
NA NA D
Nonmarxiatory Preheat Procedures A
A A
8 Minimum bolt Cross-Sectional Area A
NA NA F
kules for Evaluation of level D A
A A
A Service Limits G
Protection Against Nosuluctile Failure A A
A A
H Capacity Conver mions for Class 3 NA NA A
Safety Valves i
J wner's Design Specifications for A
NA NA O
Core Support. Structure K
Hecommended Maulmum Deviations and A
A A
O Tolerances f or Component Supports 9
h.
Li '.
Table A2-2 Dfy Cow.parison of H31.1 11955) {4) Against ASHE Section III (1977) [l[
GE$
Article
- Corresponding h
or Class Class Class Signi-Article in i
a subarticle De scr ipt ion 1
2 3
ficant B31.1 (1955)
Remarks k'S N
'r a
NA-1000 tiCOPE OF SECTION III A
A A
g R$g NA-2000 CIASSIFICATION OF 00tlIONENTS A
A A
A Not Mdressed NA-3000 REStotultlILITIES AND DUTIES A
A A
11A-4000 QUALITY ASSURA!CE NA-4100 Quality Asaparance Require:ments A
A NA A
bt)t Addressed NA-5000 It4SPECTIO!4 y
NA-5100 General Requires >ents for Authorized A
A A
A Not Moressed e
Ingiection Ajencies and Ing>=ctors NA-5200 Duties of Ing>ectors A
A A
A Not Mdressed NA-6000 QUALITY CONTIOL SYSTDIS FOR CLASS 3 couSTRUCTIOtt NA-6100 Genural Requirements NA NA A
A Not Addressed NA-6200 Organization and Regeonsib111 Lies HA NA A
A Not AdJtessed t1A-6300 Control of Operations NA NA A
A Not Mdressed NA-6400 Recosde arm! Fotos NA NA A
A Not Addressed NA-8000 CERTIFICATES OF AUT110RISATION, A
A A
NAMEPIATES, STAMPING, AND REPOkTS A Mdressed in the Code for the 42ecified class or considered significant for this review.
- Not considered significant for this review.
O Outside the scope of this review.
NA Not applicable to this review or not addresaed in the Code for the specified class.
Article number in current Code will 12o preceded 1,y 105 for Class I component, NC for Class 2 com.ponent, and HD for Class 3 component.
e 9
Ib LT m n hj
~
Table A2-2 (Cont.)
U J 3
g; Article
- Cor re sponding E3 or Class Class Class Signi-Article in N
Subarticle De wr idion 1
2 3
ticant 931.1 (1955)
Remarks N
1000 INTid) DUCTION ih 1100 Scope A
A A
A 101, Table 2a, Note 2 il to
~
2000 MATEHIAL 2100 General A
A A
A 105, Table 1, See Sect. 6 Sect. 7 2200 Material Test Coupons and Specimens A
A A
for Ferritic Steel Materials 2300 Fracture Toughness Hequirement A
A A
A Not Addressed for N terial 2400 Welding and Brazing A
A A
A Sect. 6: Chapter p
s 4 and Appendices
[
2500 Examination arul Repair of Presbure A
A A
kutainitrJ N terials 2600 Nterial Manuf acturers' Quality A
A A
A Not Addressed System Program 2700 Dimensional Starxiard A
A A
3000 DESIGN 3100 General A
A A
A Not Addressed 3200 Design by Analysis (C1. 1)3 Alternate A
A NA A
NA Design Rules for Vesbels (Cl. 2) 3300 Vessel Design A
A A
A NA 3400 Pump Design A
A A
A NA 3500 Valve Design A
A A
A 107,100,124, 129,134,139 3600 Piping Design A
A A
A Sect. 1 3700 Electrical athl Nechanical Penetration NA A
A A
NA A sbemblie s 3800 Demign of Atmospheric Storage Tanks NA A
A A
NA 3900 0-15 psi (0-103 kPa) Storage Tank NA A
A A
NA De sign
g,&
- m f/
Table A2-2 (Cont.)
VB l[a[
Article
- Cor re sponding E'
or Cla na Class Class Signi-Article in N
Sabarticle De u:r init ion 1
2 3
fIcant B11.1 (1955)
Remarka i
3'$
o G,7 4000 FABRICATION AND INSTALLATION Sect. 6 4100 General A
A A
Mt AJdressed
( <n 4200 Forialsyj, Fittisq, and Alignirn A
A A
i
- k 4300 Welding Qualifications A
A A
Appendix A to Sect. 6 4400 kules Governisw) Makits, Exaialning, A
A A
and liepairing Welds 4500 Brasisq A
A A
l 4600 Heat Treatment A
A A
4700 Hechanical Jointa A
A A
Chapter 2 of Sect. 6 p
4800 Expanalon Joints NA A
A Not Addressed i
U
$000 EAAMINATION 5100 General kequirements A
A A
A Not Addressed 5200 Required Examination of Walds A
A A
A N t Addressed i
(C1. 1) Emanination of Welds (C1. 2 and Cl. 3) 5300 Acceptance Standard A
A A
A Not AJJtessed 5400 Final Examination of Items (Cl.1):
A NA A
A Hot Addressed n ot Esamination of Welded Joints 101. 3) 5500 Qua11tications of Nondestructive A
A A
A Not AJdressed Esamination Fernonnel 5600 NA NA NA 5700 Examination Requirements of NA A
A ~
Expansion Joints 6000 TESTING 6100 General A
A A
6200 Hydrostatic A
A A
6300 Pneumatic A
A A
t J
Table A2-2 (Cont.)
g'h Article
- Cor re sponding or Class Class Class Signi-Article in ym I4 Subarticle De nicr i pt ion 1
2 3
ficant R31.1 (1955)
Hemarks 3
le Q.
6400 Pressure Test Gages A
A A
n 6500 Atmospheric and 0-15 psig NA A
A 3
Storage Tanks k
6600 Hydrostatic Tusting of Vessels NA A
NA Designed to NC-3200 6700 Pneumatic TestiswJ of Vesbels NA A
NA Designed to NC-3200 6800 6900 Proof Tests to EstabliaJe NA A
A NA Design Pressure y
7000 PHOTtCTION AGAINST OVERPRESSURE e
7100 General A
A A
NA 7200 Definitions Applicable to A
A A
NA overpressure Protection Devices 7300 Overpressure Protection Report A
A NA NA (C1. 1): Analysis (C1. 2) 7400 Helleving Capacity Requirements A
A A
NA and Acceptable Tylms of Overpressure Protection Devices 17500 Set Preuuures et Presaare Helief A
A A
NA Devices 7600 Operatirmj Design Requirements for A
A A
NA Pressure Relief Valves 7700 He<3uirements f or Nonreclosing A
A A
NA Press.ure Helief Devices 7800 Certification Requirements A
A A
NA 7900 Harking, Stamplasj, and Reports A
A A
NA 8000 NAMEPLATES, STAMPING, AND REPORTS U100 General A
A A
t i
th, E
> m {/
Table A2-2 (Cont.)
p$
[
Atticle*
Cor re spondir>J RI or Class Class Class Sign!-
Art 1cle in M
Sut>a r t icle Description 1
2 3
ficant H31.1 (1955)
Remarks l
'N
'k Halloa'10RY APPENDICES
- r h
I Design Stress Intensity Values, A
A A
A Tables 1 and 2,
- k Allowable Stresses, Material Sect. 1 Properties, and Design Fatigue Curves II Experimental Stress Analysia A
A A
1 III Damis for Entab11 Jaiwj Design A
A A
A NL Addressed Stress Intensity Values and Allowable Stress values IV Approval of New M.sterials Under A
A A
p the ASME tioller and Pressure Vesbel Code for Section III Application
.s V
Certificate lloider's Data Iteport A
A A
Forms aid A;p11 cation Yotas for Certificates of Authorization for Uee of Code Symbol Staa.ps VI Nunded Irriications Charts A
A A
VII Charts for Datermining Shell A
A A
122 Thickness of cylindrical and Spherical Components Under l
External Prestate XI Rules for Nlted Flawje NA A
A 106,111,138, Connections f or Class 2 and 3 143 Compunanta ard C1aem HC Veasels
+
XII Design Considerations for M1ted A
A A
A Flawje Connections XIII Deutgn Based on Stress Analysis NA A
NA Mt Addressed for vessels Designed in Accordaire with NC-3200 XIV Design Dased on Fatigua Analysis NA A
NA NA for Vessels Designed in Accordaire with ta:-3200 W
e 4
g,
- ,iL}
jSS;.
s :2 2M lW j
Table A2-2 (Cont.)
- r Q
Article
- Corresponding 3
or Class Class Class Sign!-
Article in 4
Subarticle De m:r ipt ion 1
2 3
ficant B31.1 11955)
Remarks XVI Nondestructive Emanination A
A A
O MA Nethoda Applicable to Core Support Structureu XVI I Design et 1.inear Wpe Support s by A
A A
O NA Linear Elan tic and Plastic Analysis NOletANDA10RY APPENUICES
[
A A
M M
D Owner's Design Specification A
A A
C Cer tif icate Holder's Stress Report A
NA NA D
Noeumandatory Preheat Procedures A
A A
E Ninimum Dolt Cross-Sectional Area A
NA NA Y
Nules for Evaluation of Level D A
A A
A Service Limits G
Protection Ajainst Nonductile Failure A
A A
A il Capacity Conversloens for Class 3 NA NA A
Safety Valves J
Owner 's Design Specifications for A
NA NA O
HA Core Support Structure K
Recommended N.sminum Deviations and A
A A
O NA Tulerances for Component Supports I
yi Table A2-3 VU [/
Cos. g fesan of ASME B&PV Code Section III lf 1965 Edition [3] Against t rae 1977 Edition [1]
st a 2 48 Correa00nding
'M Article
- Article la or Class Class Class 81gul-ASME BEFV Sect.
- I Sut>a r t icle De as:r ipt ion 1
2 3
ficant III (1965)
Remarks kO en 43 NA-1000 SCOPE OF SECTION III A
A A
N-110 NA-2000 CLASSIFICATION OY (XM4not4ENTS A
A A
A H-130 tlA-3000 RESPoteSitsII.lTIES A!3D 1)UTits A
A A
N-140 NA-4000 QUALITY ASSURAtCE NA-4100 Quality Asuurance Require:aents A
A NA A
Appendix VII e
HA-5000 I!4SPECT10tl Anticles 6,14
'o',
14A-5100 Cor eral IWyulsements for Authorized A
A A
A N-610 Isagsection Agenclus and Inapectors 14A-5200 Dutle:s of inacectors A
A A
A N-610 NA-6000 QUALITY CONTROL SYSTEMS FOR CLASS 3 OL)t4STHUCTION HA-6100 Genegal Requirements NA NA A
A Not AdJtemaed tiA-6200 Organization and Reagonalbilities NA NA A
A Not Addressed NA-6300 Control of Operations HA IJA A
A Not AJdressed 14A-6 400 Recondu and Forms 14A NA A
A Not AdJter.ed 14A-tiU00 CERTIFICATES OF AU'1HORIZATION, A
A A
NAMEPIATES, STAMPIt4G, Allu REPORTS A Addressed in the Code for the avecified class or considered significant for this review.
- Ikat conalddred asignificant for this review.
- O Outside the sa: ope of this review.
HA Not applicable to Liais review or not addressed la the Code for the specified class.
Article number in current Code will 12e lireceded 1sy NB for Class I consponent, NC for Class 2 component, and N3 for Clsas 3 component.
6 4
g, Table A2-3 (Cont.)
U.',
- 71 Correstonding h
Article
- Arkicle in 96 or Class Class Class Sign!-
ASME B6PV Sect.
E3 Sui > article De scr iist ion 1
2 3
ficant III (1965)
Hemarks
$N 2h 1000 INTRODUCTION Art. 2, 11, 21 See Note 1
$ 75
!!00 Scope A
A A
A N-210, M-2110
? :r h
2000 MATERIAL Articles 3,12
,k 2100 General A
A A
A N-310 2200 Nterial Test Coupons and Specimens A
A A
A for Ferritic Stsel Materials 2300 Fracture Toughness Heguirement A
A A
A N-330 for Material 2400 WeldiwJ and BrazirpJ A
A A
A Not Addressed 2500 Examination and Hepair of Pressure A
A A
N-320 Metaining Materials 2600 Material Manufacturers' Quality A
A A
A N-614 p
4 System Program a
2700 Dimensional Standard A
A A
3000 DESIGN Articles 4, 13 3100 General A
A A
A N-440 Only Class 1 3200 Design by Analysis (C1.1) g Alternate A A
NA A
N-430 Only C1. 1, See Design Rules for vessels (C1. 2)
Note 2 3300 Vessel Design A
A A
A Articles 4, 13 3400 Pump Design A
A A
A Not Addressed 3500 Valve Design A
A A
A Not Addressed 3600 Piping Design A
A A
A N-150/Mostly Not Addressed 3700 Electrical and Nechanical Penetration NA A
A A
Not Addressed Assemblie n 3800 Dusign of Atmospheric Storage Tanks NA A
A A
Not Addressed 3900 0-15 pal (0-103 kPa) Storage Tank NA A
A A
Not Addressed De sign i
Notess 1.
For requirements of Class 3 vessel, reference is made to Section VIII of the Code.
2.
Une Table 1-1.0 (1977 Edition) shen designing pressure vessels by Alternative Design Analysis (NC-3200).
Ib ga o n h' Table A2-3 (Cont.)
ya lh Corresponding a3 Article
- Article in N
oc Class Class Class Signl-ASME B&PV Sect.
Suhartic).
De u:r ipt ion 1
2 3
ficant III (1965)
Rema rk s g3#
4000 FADHICATION AND INSTAL!.hTION Article 5 Only Class 1 4100 General A
A A
A N-510 Only Class 1 Oh 4200 FosalseJ, Fitting, and AllgratswJ A
A A
4100 Weldisej Qualifications A
A A
A N-520, N-540 Only Class 1 4400 Rules Covernirs MakirwJ, ExaminityJ, A
A A
i and HeparlswJ Welds 4500 Braxisej A
A A
l 4600 heat Treatment A
A A
N-530 Only Class 1 4700 Hechanical Joints A
A A
Not Addressed 4000 Expansion Joints NA A
A Not Addressed 8,
5000 ELV4INATION Articles 6, 14 Only Class 1, 2 ca 5100 Ceneral Requirements A
A A
A N-610 Only Class 1
$200 Eequired Exasisinationi of Welds A
A A
A N-620 (C1. 11 : Exassination of Welds (C1.2 and Cl. 3) 5300 Acceptance Stasulard A
A A
A N-626.5, N-627.7 Only Class 1 5400 Final Examination of Items (C1.1):
A NA A
A N-620 Only Class 1 Spot L amination of Welded Joints I
(Cl. 3) 5500 Qualifications of Hondestructive A
A A
A Not Addressed Examination Personsteal 5600 NA tlA hA 5700 Examination Requirements of NA A
A Expansion Joints 6000 TESTING Article 7 6100 General A
A A
6200
!!ydrostatic A
A A
6300 Pneumatic A
A A
6400 Pressure Test Gages A
A A
6500 Atmospherin and 0-15 psig NA A
A Storage Tanks e
l7
>m I&
E3 Table A2-3 (Cont.)
2M
{$
Corre @1nding gk Article
- Article in
'T or Class Class Class Signi-ASME B&PV Sect.
Q Sutaa r t icle De se r i pt t on 1
2 3
ficant iII (1965)
Remarke sa 6600 Hydrostatic Testing of Vessels NA A
NA Designed to NC-3200 6700 Pneumatic Testing of Vesbels NA A
NA Designed to NC-3200 6800 6900 Proof Tests to Establish NA A
A A
Not Addressed See Table A2-4 Design Prestaure for Class A or B for Class C 3*
4 7000 PROTECTION AGAINST OVERPRESSURE 5
7100 General A
A A
NA 7200 Definitions Applicable to A
A A
NA Overpresuure Protection Devices 7300 Overprest.ute Protection Heport A
A NA NA (C1. 1); Analysis (C1. 2) 7400 Helleving Capacity Requirements A
A A
NA and Acceptable types of Overpressure Protection Devices 1500 Set Pressures of Pressure Relief A
A A
NA Devices 7600 Operating Design Requirements for A
A A
NA Pressure Relief Valves 7700 Requirements for Nonreclosing A
A A
NA Pressure Relief Devices 7800 Certitication Requirements A
A A
NA 7900 Marking, Stamining, and Heports A
A A
NA 8000 NAMEPLATES, STAMPING, At4D REPORTS Articles 8, 15 8100 General A
A A
Table A2-3 (Cont.)
h.;
Correhoonding
- ~=
Article
- Article in 3
or Class Class Class Sign!-
p:-
Subarticle De ucr itat ion 1
2 3
ficant IIT (1965)
Rema r k s a3g MANDA'IOM APPENDICES f'$
R 1
Dealyn Stress Intenasity Values, A
A A
A Article 4 Table N-421 Allowable Stresses, Nterial
- fp Progwrties, and Design Yatigue b3 Curves 4
II Experimental Stress Analysis A
A A
III Basis for Establishir>J Design A
A A
A Appendix II Stress Intensity Values aikt Allowable Stress values IV Approval of New Nterials Under A
A A
the ASME Boiler ard Pressure Vessel Code for Section III Application V
Cer tif icate Holder's Data keport A
A A
Y Fossa asal Application Fogan for Certificates of Authorization for
)
Unna of Code Syndel Stampe VI Nunded Irmlications Ch.:s ts A
A A
VII Charts f or Determinir>J bhell A
A A
Article I-1 of Tiilcknesa of Cyltraf rical and Anvendix I, N-431 Splierical Components Unaler Extetnal Pressure XI Eules f or Nlted Flange NA A
A Article I-12 of Connections f or Clama 2 anal 3 Appendix I, N-471, C4.mponents and Cla ss MC Vessels Table N-422 XII Design Considerations for Isolted A
A A
A Art.icle I-12 of Flas,Je Cormections A perklix I, N-471 l
XIII Design Baied on Stress Analysis NA A
NA Article 1-10 of for Vennels Designed in Accordance A permlix I, N-430 with NC-3200 XIV Design Dased on Fatigue Analyasis NA A
NA Article 1-10 of f or Vessels Designed in Accord.ince Appendix Z, N-430 witte NC-3200 XVI Noshlestructive Examination A
A A
O NA Nthods Applicable to Core Support Structures XVII Design of Linear Type Supports by A
A A
O NA Linear Elastic and Plastic Analysis 45 e
t
I
[qe{
Va ]V' in; a3 Ic7 E
f'3 a
g or Table A2-3 (Cont.)
O n
Cor,re sponding Article
- Article in or Class Class Class Signi-ASME B6PV Sect.
Sut>a r t icle Demeription 1
2 3
ficant III 11965)
Rema r k s N0letANDA10RY APPENDICES A
A NA NA p
B Owner's Design Specification A
A A
[3 C
Certificate Holder's Stress Report A
NA NA 6"
D Nonaandatory Preheat Procedures A
A A
Appendix III E
Minimum bolt Cross-Sectional Area A
NA NA F
Rules for Evaluation of Level D A
A A
A Not Addressed Service Limits G
Protection Against Nanductile Faillre A A
A A
Article N-330 Only Class 1 11 Capacity Conver sions f or Class 3 NA NA A
Satety Valves J
Owner's Design Specifications for A
NA NA O
NA Cure Support Structure K
Hecommended Maulmum Deviations and A
A A
O NA Tulerances for Component Supports
Table A2-4 Crmaparia.an of ASME VIII (1965) [5] with ASME III (1977) [1]
c_Wi (3
/
Article
- Corresponding or Class Class Class Signi-Article in y
Sut>a r t icle Dem:ription 1
2 3
ficant ASME VIII (1965)
Rema r k s s 5' j[j3 NA-1000 SCUPE OF SECTION III A
A A
'5 k
NA-2000 CIASSIFICATION OF 0)HPONENTS A
A A
A NA
- r Q
NA-3000 RESK)uSIBILITIES AND DUTit:S A
A A
la4 14A-4000 QUALITY ASSURANCE NA-4100 Quality Amuurance Requirements A
A liA A
NA NA-5000 INSPECTION 84A-5100 General Requirements for Authorized A
A A
A UG-90 Inq=ction A,Jencies and Inq>ectors NA-5200 Duties of Innspectors A
A A
A UG-91 I
HA-6000 QUALITY CONT 1CL SYSTu43 FOk CIASS 3
[
QXGTl*UCTION NA-6100 General kequirements tA NA A
A NA NA-6200 Oaganization and Responalbilities NA I4A A
A NA 14A-6300 Control of Operations HA NA A
A NA NA-6400 Recoads and Forms NA NA A
A NA NA-t:000 CERTIFICATES OF AU113ORIZATION, A
A A
UG-116 NAetEP1ATES, STAMPit4G, AND REPORTS 1000 INTutX)(X' TION 1100 Scope A
A A
A U-1 i
2000 MATERIAL 2100 General A
A A
A UG-5 2200 Material Test Coupons and Specimens A
A A
for Ferritic Steel Materials A
Addressed in the Code for the a4>ecified class or considered significant for this review.
- Not considered algnificant for this review.
O outside the acope of this review.
NA Nut applicatale to this review or not addressed in the Code for the specified class.
Article nusher in current Code will be g> receded by NB for class I component, NC for Class 2 component, and ND for Class 3 component.
b e
~
b fi*
Table A2-4 (Cont.)
- ?
SE Cor'regending Article
- Class Class Class Signi-Article in J
or subarticle Description 1
2 3
ficant ASME VIII (1965)
Remarks g
9 2300 Fracture Toughness Requirement A
A A
A UG-04 n
for Material
,g 5
2400 Wolding armi Brazisej A
A A
A iM 2500 Examination and Repair of Pressure A
A A
Retaining H.aterlatu 2600 Material Manufacturers' Quality A
A A
A UG-93 I
System Program 2703 Dimensional Standard A
A A
3000 DESIGN
- p 3100 General A
A A
A NA E
3200 Design by Analyula (CI.1) 3 Alternate A A
NA A
Design Hules for Vessels (C1. 2)
W 3300 Vessel Dea!9n A
A A
A iM-8, UF-12 3400 Pua4> Design A
A A
A NA 3500 Valve Design A
A A
A NA 3b00 Piping Design A
A A
A NA 3700 Electrical and Mcchanical Penetration NA A
A A
NA A ssemblie s 3t100 Design of Ata0hpheric Storage Tanks NA A
A A
NA 3900 0-15 pai (0-103 kPa) Stor.sge Tank NA A
A A
NA Design 4000 FABRICATION AND INSTAI.LATION 4100 General A
A A
A UG-75 4200 Forming, Fitting, and Aligning A
A A
4300 Welding Qualifications A
A A
A iM-20, UW-29 4400 Mules Governinj Making, Examining, A
A A
and Repairing Welds 4500 Brazing A
A A
4600 Heat Treatment A
A A
l?.
Table A2-4 (Cont.)
=
r,c.
gK Atticle*
corredponding E3 or Class Class Class Signi-Article in N
Sutarticle Description 1
2 3
ficant ASME VIII (1965) kema r k s k'I, 4700 Mechanical Joints A
A A
UR-19 4800 Expanalon Joints HA A
A HA
[.
5000 EAAMINATION 5100 General Requirements A
A A
A UG-90
$200 Hequired Examination of Welds A
A A
A tM-46 (C1. 1): Examination of Walda (C1. 2 anal C1. 3) 5300 Acceptance Standasd A
A A
A tM-51 (1) 5400 Final Examination of Items A
NA A
A IIA UG-99 (g) require s (C1. 1)s Spot Examination of in maction after Welded Joints (C1. 3) hydrostatic but does a
not specify 11guld
'i' penetrant or sagnetic particle inspections tM-50 requires LPE or magnetic particle in wection before pneumatic testing.
5500 Qualifications of Nondesttuctive A
A A
A NA UG-91 gives Examination reraunnel requirements foe qualificaticr. of Anapector s, but nor.
7 HDE pera>nnel 5600 NA NA HA
$700 Examination Requirements of NA A
A Expanulon Joints s
6000 TESTING 6100 Ceneral A
A A
6200 Uydrontatic A
A A
6300 Pneumatic A
A A
e O
W
l 14 p>
P:ii; [}/
> vi]
Table A2-4 (Cont.)
3 h:$
- e. 3 Cor rd sponding pg Article
- Class Class Class Signi-Article in q
or Subarticle Deecription 1
2 3
ficant ASME VIII (1965)
Remarks fR l
6400 Pressure Test Gages A
A A
5gp 3
6500 Atmo6pheric and 0-15 psig NA A
A 4
Storage Tanks 6600 tlydrostatic Testing of Vessels NA A
NA I
Designed to NC-3200 6600 Pneumatic Testing of Vessels NA A
NA Designed to NC-3200 6800 6900 Proof Tests to Estabil6h NA A
A A
UG-101 Design Pressure P
ey 7000 PHOTECTION AGAINST OVERPRESSURE 7100 General A
A A
7200 Definitions Applicable to A
A A
Overpressure Protection Devices 7300 Overpressure Protection Helort A
A NA (C1.1) g Analysis (Cl. 2) 7400 Helleving Capacity Hequirements A
A A
and Acceptable Types of Overpressure Protection Devices 7500 Set Pres 6ures of Psessure Helief A
A A
Devices 7600 Operating Design Requirements for A
A A
l'ressure Helief Valves 7700 Hequirements for Nonreclosing A
A A
Pressure Halief Devices 7800 Certification Hequirements A
A A
190n Marking, Stampisuj, and Reports A
A A
8000 NAMEPLATES, STAMPING, AND HEPORTS 8100 General A
A A
l 1
L----
LE F*
Table A2-4 (Cont.)
4 !$
Atticle*
CorreadonJing Class Class Class Signi-Article in or 2M Sut>a r t icle De uct lpt ion 1
2 3
ficant ASitM VIII (1965)
Remark s
$ ce g3 itAt4DA' LORY AFFEt4 DICES fn g
I Design Stress Intensity Values, A
A A
A Subsection C Fatigue Curves Allowable Staesses, Mater ial not included in Proper ties, and Design Fatigue Sect VIII Curves II Emperiamental St ress Analysis A
A A
III
- Basis for EstabliaJil N Design A
A A
A Appendices P&Q Stress Intensity Values and Allowable Stress Values IV Approval of New Nterials Unider A
A A
p the ASHd Niler arkt Pressure Vessel h
Cale for Section III Application m
V Certificate lloider's IMta Reg =2rt A
A A
Forum and Application Fotma for Certiflenteu of Autlaorization for Une of Cale Symbol Stataps VI Nonded Isklicationas Claar ta A
A A
VII Ctaarts for Determinity Shell A
A A
UG-28 & Appendix V n icknema of Cylindrical and Spherical Components Under External Pressiste XI Itules for Nited Flange NA A
A Appendix II Connections for Class 2 and 3 Comp 2nents arul Class HC Veneels XII Design Considerations for Bolted A
A A
A NA Flange Connections XIII Design based on Stress Analysis HA A
NA for Vem.els Designed in Accordance with IJC-3200 XIV Design baa.ed on Fatigue Analysia 14A A
tiA for Veubels Deutgued in Accordance wilia NC-3200 4
8 W
4
5.
.p.
D,il}
j'y I/
9K t3 Y$
IE h
Table A2-4 (Cont.)
n S
k Article
- Correa@onding or Class Class Class Signi-Article in Sut>ar t icle Description 1
2 3
ficant ASNE VIII (1965)
Itemark s XVI Hondestructive Examination A
A A
O Hethods Applicable to Core Support Structures XVII Design of Linear Type Support s by A
A A
O P
Linear Elastic and Plastic Analysis I
90 4
NotetANDA'!URY APPENDICES A
A NA NA 11 Owner's Design Specification A
A A
C Cettificate Itolder
- a Streas Report A
NA NA 0
Nnmarmlatory Prelaeat Procedures A
A A
E Ninimum Isolt Cross-Section Area A
NA NA F
Rules for Evaluation of Level D A
A A
A NA Service I.imits G
Protection Against Anductile Failure A A
A A
NA 11 Capacity Conver sions for Class 3 NA NA A
Safety Valves J
Owner 's Design Specit ications for A
NA NA O
Core Support Structure K
Recomumended Aximum Deviations and A
A A
O Tolerances f or Component Supports e
Past Class C vessels were built in accordance with the requirements of ASME VIII (1965), except for inspection and the longitudinal (Category A) and circumferential (Category B) welding requirements noted in Section 4.3.
Past Class C vessels could be classified in accordance with current requirements (1) as either Class 2 (Quality Group B) or Class 3 (Quality Group C).
2.4 PCMPS See Section 4.4 of this appendix.
2.5 VALVES See Section 4.5 of this appendix.
2.6 HEAT E: CHANGERS Heat exchangers were usually designed to the ASME Boiler and Pressure vessel Code,Section III, 1965 Edition (3], and Section VIII [5], which are discussed in Sections 2.3 and 4.3 of this appendix, and to the Standards of the Tubular Exchanger Manufacturers Association (TEMA), 1959 Edition (B).
Discussions regarding TEMA may be found in Section 4.6 of this appendix.
2.7 STCRAGE TASKS Otorage tan <s that must withstand pressures above atmospheric were usually designed to the ASME Boiler and Pressure Vescel Code,Section III, 1965 Edition (3], which is discussed in Sections 2.3 and 4.3 of this appendix.
Aluminum tanks might have been designed to " USA Standard Specification for Welded Aluminum-Alloy Field-Erected Storage Tanks," USAS B96.1-1967 (9].
Storage tan *eee m v coem.au,
4.
COMPARISON OF SIGNIFICANT CURPINT CODE REQUIREMENTS l
AND PAST P2QUIREMENTS 4.1 GENERAL REQUIREMENTS Section 4.1 compares the significant general requirements of the current code (1) with past requirements.
In addition, where feasible, an approach is formulated which facilitates the review of nuclear components and gystems designed and built in accordance with past requirements to be evaluated from the viewpoint of current requirements. The general requirements discussed herein are fracture toughness, quality assurance, quality group classification, and code stress limits.
4.1.1 Fracture Toughness Recuirements Class 1 Comoonents The current code requires that pressure-retaining materials for Class 1 components shall be impact tested to determine T
e arpy V-Notch test, except for materials whose nominal NDT thickness is 7/8 in or less; bolts 1 in or less; bars with nominal sectional area 1 sq in or less; pipes, fittings, pamps, and valves with nominal pipe size 5 in or less; austenitic stainless steels; and non-ferreus materials.
Drop weight tests are not required for martensitic high alloy chromium (Series 4xx) and precipitation-hardening steels listed in Appendix I (le]; however, otner requirements of Na-2332 (1b] do apply.
Class 2 Comoonents Pressure-retaining materials for Class 2 components are required to be impact tested with exceptions as outlined for Class 1 components. Al so exempted are commonly used plate, forging, and casting materials listed in Table NC-2311(a)-1 of Reference ic when used in Class 2 components whose lowest service temperature (LST) * ?xceeds the tabulated nil ductility transition temperature (TNDT) by at least the thickness-dependent value A,
- See Tacle A4-1 for definitions of cccmonly used terms and symbols.
A-30 d Frankin Research Center
% w N %n twe
determined from the curve in Figure NC-2311(a)-1 from Reference lc.
For convenience, the table and the figure are reproduced as Table A4-2 and Figure A4-1, respectively. Materials for components whose LST exceeds 150*F are also exempt from impact testing.
Drop weight tests are not required for martensitic high alloy (Series 4xx) and precipitation-hardening steels listed in Appendix I of Reference le.
Charpy V-Notch testing or alternative testing as described in NC-2331 (lc]
applies for these steels in all thicknesses. For nominal wall thicknesses greater than 2.5 in, the required C values shall be 40 mils lateral expansion.
l l
I 1
l l
l i
l I
I A-31
...) FrankJin Resesrch Center I,
A Men at N v.m mm,,
Table A4-1 Definition of Commonly Used Fracture Touchness Terms and Symbols Symbol Definition TNDT A temperature at or above the nil ductility temperature as determined by a " break, no-break" drop weight test in accordance with ASTM E208.
(The nil ductility temperature is that temperature above which cleavage fracture can be initiated only af ter appreciable plastic flow at the base of the notch and below which cleavage will be initiated with little evidence of notch ductility.)
TNDT is 10*F below the temperature at which at least two specimens show no-break performance.
RTNDT The higher of TNDT or (Tey - 60'F).
T A temperature above TNDT at which three specimens made and ey tested in accordance with SA-370 Charpy V-Notch testing exhibit at least 35 mils lateral expansion and not less than 50 f t-lb absorbed energy.
LST Lowest Service Temperature:
the minimum temperature of the fluid retained by the component or the calculated minimum metal temperature expected during normal operation whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pressure.
1 g
A-32 J.dd Franknn Research Center 4 % of 'he Frarumn lr.ma
Table A4-2 TABLE NC-2311(a)-1 EXEMPTIONS FROM IMPACT TESTING UNDER NC 2311(al-8 Material Material' Condition T..,8 deg. F SA 537 C! ass 1 N
-30 SA-516-Grade 70 Q&T
-10 SA-516-Grade 70 N
O SA-5ca C: ass 1 Q&T
+10 SA-533 cra:e B Q&T
+10 SA-2e.
N
+20 SA 216. Grades Q&T
+30 WCB,WCC SA-36 (P! ate)
+40 SA-5ca-Class 2 Q&T 40 NOTES:
I (1) These materiais are esempt from toughness test:ng een A or LST T.oris ateve the curve in Fig NC 2311(a)-1, for the I
thickness as deficed in NC 2331 or NC-2332.
(2) Material Cordition tetters refer to:
N - Nermalize Q & T - Quencn and Temper HR - Het Relled (3) These values fer T.or *ere estactisted from data on heavy section steel (thickness greater tsan 2% in.). Va.ues fer sections less than 2% in. thick are held censtant untd add.ticnal data is C0tained.
(4) Materrats mace to a fine grain metting prac'Jce.
O A-33
@c
.. a Frank!in Research Center a w m e N v m.aue l
~
s
\\
N w
ceo
>=
<c 2w A
I o
2 w
P-w t
M
+
w e
c w
L W
s.
M w
5 m
3 x
o t
w s
w e
2 a
-m m
_v c
m t,
t m
..e x
=
w g
A o
k a
o 3
m a.
o.
E 4
3 2
a-
=.:
xw P-wo
?9 No z
N
- a
-w Es e
8 o
8 o
e o o
u a
w n
c=
d !* *: 10% _ us,jj. y Figure A4-1 A-34 O,j0J Frankin Research Center i
s w a m w-, w,. a
1 l
Class 3 Components Pressure-retaining materials for Class 3 components are required to be tested, except as outlined for Class 1 components and the materials listed in Table ND-2311-1 [ld) in the thicknesses shown when the LST for the component is at or above the tabulated temperature.
For convenience, Table ND-2311-1 has been reproduced as Table A4-3.
In addition, materials for components for which the LST exceeds 100*F are exempt from impact testing.
l i
The evaluation of materials based on past codes for which fracture toughness requirements may not have been specified or limited is facilitated by the survey forms shown as Tables A4-4, A4-5, and A4-6 for Class 1, Class 2, and Class 3 comporients or systems, respectively.
Example Tables A4-2 through A4-6 and Figure A4-1 will be used to evaluate the resistance to brittle fracture of components whose design is based on past codes for which impact testing may not have been required. The following is an example of how the tables and the figure will be used.
Consider the 42-in primary pipe line between the reactor vessel and steam generator in the Palisades plant. These pipes were fabricated from 3.75-in-thick ASTM 516, Grade 70 plate with a rolled bond 1/4-in nominal cladding of 304L stainless steel. The design temperature is 650*F.
The safety injection system is designed to cool the primary system to 130*F in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with a maximum pressure of 270 psig as noted in Reference 11.
The LST is taken as 130*F.
From Table A4-3, TNDT = 0'F for SA-516 Grade 70.
From Figure A4-1, A = 48' for material 3.75 in thick:
(LST - TNDT) = 13 0 * - O' = 13 0
- F > 4 8 'F = A so tnat this material, if it were a Class 2 or 3 component, would be exempt from impact testing. The fact that the primary coolant piping is Class 1 would not exempt it from impact testing based on present code requirements. However, the fact that the LST exceeds the TNDT by more than 150% of A allows us to conclude that the primary coolant piping material used in the construction of the Palisades plant is adequate, provided that exposure to radiation does not induce an increase of the TNDT sufficient to require the fracture =echanics approach outlined in Appendix G [4e}.
In this regard, note that paragraph NB-2332(b) [lb]
indicates that if the LST exceeds the reference nil ductility transition temperature (RTNDT) by 100*F, then the fracture mechanics approach of Appendix G is not required.
In this example:
(LST - Tgg;) = 130*F > 100*F so tnat the material for tne Palisades primary coolant piping is considered adequate.
A-35 g
...d Franidin Research Center a won or n, r,.a.an in ui.
Table A4-3 TABLE ND.23111 EXEMPTIONS FROM IMPACT TESTING UNDER ND-2311(a)(8)
Lowest Service Temperature for the Thickr:ess Shown Over % in.
Over % In.
Over 1 in.
Over 1% in.
to % in., ir:d.
to 1 in Ind.
to 1% in., incl.
to 2% In., inct Material (Over 16 mn.
(Over 19 mm (Over 25 mm (Over 38 mm Material Condition' to 19 mm, ind.)
to 25 mm, ind.)
to 38 mm, ind.)
to M mm. Ind.)
SA 516 Grade 70 N
-30 F (-34 C)
-20 F (-29 C) 0 F (-18 C) 0 F (-18 C)
SA 537 Class 1 N
-40 F (-40 C)
-30 F (-34 C)
-30 F (-34 C)
-30 F (-34 C)
SA 516 Grace 70 Q&T (2)
(2)
(2)
-10 F (-23 C)
SA 508 C! ass 1
.Q&T (2)
(2)
(2) 10 F (-12 C)
SA 503 Class 2 Q&T (2)
(2)
(2) 40 F (4 C)
SA 533 Grade 33 Q&T (2)
(2)
(2) 10 F (-12 C)
Ctass 1 SA 216 Grades Q&T (2)
(2)
(2) 30 F (-1 C)
(2)
(2)
(2) 20 F (-7 C)
NOT!?:
(1) Material Conciticn letters refer to:
N - N crmalize Q & T - Ct.ench and Temper (2) The lowest service temperature sMown in the cohJmn "Over 1% in. to 2% in." may be used for these thickresses.
(3) Materf.; mace to a fine grain metticg practice.
1 t
5
-36 dJ Franklin Research Center stwena%vusnw.
._= ____.
._. _ -. _ =, -
Table A4-4 Evaluation for Fracture Toughness of Pressure-Retaining Material for Class 1 Component / System Nuclear Power Plant FSAR Page I.
Component / System Data 1.
De scription of Component / System:
2.
Material Description and Thickness: P No.
3.
Design Temperature:
'F 4.
Design Pre ssure:
psi Lowe st Service Temperature ( I (LST) :
'F 5.
6.
Pre ssure at LST:
psi 7.
Fracture Toughness Requirement? Yes No II.
Evaluation Material is exempt (
from impact testing because 8.
(a)
Nominal thickness 5/8 in or less (b)
Bolts 1 in or thinner (c)
Bars with nominal 1 sq in cross section or less (d)
Pipe s, fittings, pumps, and valves, nominal pipe size of 6-in diameter or smaller NOTES:
1.
Lowest Service Temperature ( LST) is the minimum temperature of the fluid retained by the component or, alternatively, the calculated minimum metal temperature whenever the pressure within the component exceeds 20% of the precperational system hydrostatic test pressure
[1].
2.
Welding material used to join materials with P Nos.1, 3, 4, 5, 6, 7, 9, and 11, whicn are exempt from impact testing because of 8(a) through 8(f), is likewise exempt from impact te sting.
- However, exemption 9 does not exempt either the weld metal (NB-2430) or the weloing procedure qualification (NB-4335) from impact te sting.
See paragraph NB-2431 of Reference Ib.
3.
The current code does not exempt Class 1 components from impact testing on the casis of tabulated TNDT and A values as it does Class 2 components.
Item 9 is not an exemption listed in paragraph NB-2311 but a conservative adaptation of NC-2311(a) (8) for Class 2 components to facilitate the SEP review.
A-37
. 2) Frankin Research Center a w o# ne o.n n..
Table A4-4 (Cont. )
(e)
Austenitic stainless steel (f)
Non-ferrous material 9.
Fracture toughness of material appears does not appear to be adequate on the basis of the following evaluation:
(a) for material other than bolting and up to 2-1/2 in thick:
TNDT =
'F (See Table NC-2311(a)-1)
(
Other reference usedI4) :
)
- and,
- F which exceeds 90$F (LST - TNDT)
=
which does not exceed 90'F i
(b) for material other than bolting in excess of 2-1/2 in thick:
RTNDT "
- F (Reference used(4) :
)
- and, F __
which exceeds 120*P (LST - RTNDT)
=
which does not exceed 120*F 10.
For Qo[ ting material in excess of 1-in diameter, reference data has been available has not been available and found to sati sf y not sati sfy 7
the requirements of N3-2333 (4(b)]
l 11.
Fracture toughness cannot be evaluated because of in suf ficient information.
12.
Material is not exempt from impact testing.
NOT2:
4.
When using references other than the current code to obtain TNDT and RTgg7, be sure that the data have been obtained from specimens whose endition matches the material being evaluated (e.g., normalized or quenched and tempered) and that have designatien such as "SA-516 Gr. 70".
n.
1 I
i l
l A-38 J./J Frank'in Research Center ac~ m atvev eaer m a
l i
Table A4-4 (Cont.)
III.
Conclusions Fracture toughness appears to be adequate.
Adequacy of f racture toughness not established; request supplemental test data and supporting documents.
Welding material is is not exempt from impact testing on the basis of foregoing data and Note 2.
1 L
I i
e e
I i
a L
A-39 4
%.) FrankJin Research Center A w.ca w 3. arm. i,
o Table A4-5 Evaluation for Fracture Toughness of Pressure-Retaining Material for Class 2 Cor.ponent/ System Nuclear Power Plant FSAR Page
~
4 I.
Component / System Data 1.
Description of Component / System:
2.
Material De scription and Thickness: P No.
3.
Design Te=perature:
- F 4.
Design Pre ssures psi II 5.
Lewest Service Temperature (LST) :
'F 6.
Pressure at LST:
psi 7.
Fracture Toughness Requirement? Yes No II.
Evaluation II 8.
Material is exempt from impact testing because:
(a)
Nominal thickness 5/8 in or less (b)
Bolts 1 in or thinner (c)
Bars with nominal 1 sq in cross section or less (d)
Pipe s, fittings, pumps, and valves, nominal pipe size of 6-in diameter or smaller (e)
Austenitic stainless steel (f)
Non-f errous material NOTES:
1.
Lowest Service Temperature (LST) is the minimum temperature of the fluid retained by the component or, alternatively, the calculated minimum metal temperature whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pressure
[1].
2.
Welding material used to join materials with P Nos. 1, 3, 4, 5, 6, 7,.
9, and 11, which are exempt from impact testing because of 8(a) through 8 (f), or 8(h), is likewise exempt from testing. Ecwaver, S(g) exemption does ot exempt either the weld metal (NC-1430) or the weld precedure qual Zication (NC-4325) from impact testing. See paragraph NC-2431 of Ref erence lc.
A-40 Jd) Fran'jin Research Center r
A C. amen at The Frershn v.soa.re
I Table A4-5 (Cont.)
(g)
LST of material listed in Table NC-2311(a)-1 (see Table A4-2) exceeds TNDT by at lea st " A" (A depends on tnickne ss). (2)
LST
- F (from FSAR)
TNDT
'F (Table NC-2311(a)-1, Summer 1977 Addenda)
A
'F (Figure NC-2311(a)-1, Summer 1977 Addenda)
(Reproduced on p.
)
LST - TNDT =
F is is not greater than A.
(h)
LST exceeds 150*F.
9.
Fracture toughness cannot be evaluated because of insuf ficient information.
10.
Material is not exempt from impact testing.
III. Conclusions Fracture toughness appear s to be adequate.
Adequacy of fracture toughness not established; request supplemental test data and supporting documents.
Welcing material is is not exempt from impact testing on the basis of foregoing data and Note 2.
A-41 Ds dJ Franklin Research Center
+ = o w r e,=v.
r
Table A.4-6 Evaluation for Fracture Toughness of Pressure-Retaining Material for Class 3 Component / System Nuclear Power Plant FSAR Page
~
I.
Camponent/ System Data 1.
Description of Component / System:
2.
Material De scription and Thickness: P No.
3.
Design Temperature:
- F 4.
Design Pressure:
psi 5.
Lowe st Service Temperature (LST) :
- F 6.
Pressure at LST:
psi 7.
Fracture Toughness Requirement? Yes No II.
Evaluation Material is exempt (
from impact testing because:
8.
(a)
Nominal thickness 5/8 in or less (b)
Bolts 1 in or thinner (c)
Bars with nominal 1 sq in cross section or less (d)
Pipes, fittings, pu=ps, and valves, nominal pipe size of 6-in diameter or smaller (e)
Austenitic stainless steel (f)
Non-f errous material I
NOTES:
1.
Lowest Service Temperature (LST) is the minimum temperature of the fluid retained by the camponent or, alternatively, the calculated minimum metal temperature whenever the pressure within the component exceeds 20% of the preoperational system hydrostatic test pre ssure [1].
i 2.
Welding material u sed to join materials with P Nos.1, 3, 4, 5, 6, 7, 9, and 11, which are exempt from impact testing because of 8(a) through 8 (f), or 8(h), is likewise exempt from testing.
However, exemption 8(g) does not exe.2pt either the weld metal (NC-2430) or the weld procedure qualification (NC-4335) from impact testing. See paragraph NC-2431 of Ref erence ld.
A-42 ddO5nklin Research Center A c m,oa w n r
Table A4-6 (Cont. )
in Table NC-2311(a)-1 for the (g)
LST equals or exceeds TNDT material and thickness being evaluated. (2)
(h)
LST exceeds 100'F.
9.
Fracture toughness cannot be evaluated because of insufficient information.
10.
Material is not exempt from impact testing.
III.
Conclu sions Fracture toughness appears to be adequate.
Adequacy of fracture toughness not established; request supplemental test data and supporting documents.
Welding material is is not exempt from impact testing on the basis of foregoing data and Note 2.
)
A-43 g
..;J Frankjin Research Center
- >w e m r,em,e,au,
4.1.2 Quality Assurance Recuirements The current code [1] requires that activities in connection with the i
design and construction ( ' of ASME III nuclear power plant components and systems be performed in accordance with a quality assurance program that provides adequate confidence in compliance with the rules of Section III.
The program is to be planned, documented, controlled, managed, and evaluated in accordance with Article NCA-4000 (3) for Class 1 and 2 items, and in I3) and NCA-8122 (3) for Class 3 items. The accordance with NCA-4135 quality assurance program is to be established and documented prior to the issuance of a Certificate of Authorization by the American Society of Mechanical Engineers af ter the program has been evaluated and accepted by the society.
Por class 1 and 2 items, the program is to be documented in detail in a quality assurance manual which should include policies, procedures, and instructions which demonstrate provisions fort a.
an organization with sufficient authority, freedom, and independence from cost and schedule considerations tot 1.
identify quality problems 2.
initiate, recommend, or provide solutions 3.
verify Lsplementation of solutions 4.
limit and control further work on nonconforming ite=s until proper disposition, and with direct access to appropriate levels of management to assure proper execution of the program b.
indoctrination and training of qualified personnel c.
notification of the authorized inspection agency of significant enanges in the program 1.
Quality assurance requirements have been determined to be outside the scope of SEP Topic III-l according to the letter frco S. Bajwa to S. Carf agno 4
dated December 10, 1981. This discussion is provided as general information.
2.
Construction under Division 1 includes materials, design, fabrication, examination, testing, installation, inspection, and certification.
3.
See Summer 1977 and Su=mer 1973 Addenda to ASMZ III (1977) General Requirements.
A-44 1
I e
d.
control of the design to assure compliance with the design specification of Section III e.
design review and checking by individuals or groups other than those who performed the original design f.
documentation for procurement of materials and subcontracted services requiring compliance witn Section III g.
document control with provisions for review of changes h.
identification and traceability of materials i.
the control of construction processes 3
examination, testing, and inspections verifying the quality of work by persons independent from supervisors immediately responsible for the work being inspected, and using measuring and test equipment calibrated against measurement standards traceable to national standards (where such standards exist) at intervals sufficient to maintain accuracy within necessary limits k.
proper handling, storage, snipping, and preservation of materials and components 1.
identification of items with suitable marking to indicate the status of examinations and tests, including conformance or non-conformance to the examination and test requirements I
i m.
prompt identification and corrective action of significant conditions I
adverse to quality, with documented measures to preclude repetition n.
maintenance of quality assurance records as specified in NCA-4134.17 of Reference 1, including maintaining for the life of the plant as a minimum, the following: a permanent record file, certified design and construction specifications, drawings and reports, data reports, l
certified stress reports, certified as built drawings, material test reports, non-destructive examination reports, and test treatment reports o.
a comprehensive system of planned and periodic audits with documentation of results, follow-up action, and re-audit of deficient areas.
Class 3 items are to be designed and constructed in accordance with the quality control requirements of NCA-4135 of Reference 1, wnich include:
a.
an organizaticn chart which reflects the actual organization b.
a quality control system suitable to the complexity of the work and size of the organization A-45 Os MJ Franklin Research Center 4 w eerwa m
c.
persons who perform quality control functions with suf ficient responsiblity, authority, and independence to implement the quality control system, identify problems, and initiate, recommend, and provice solutions.
The quality control system for Class 3 construction is evaluated for compliance with the requirements of Section III (1) by the authorized inspection agency and either a representative of the American Society of Mechanical Engineers or the jurisdictional authority at the construction site as required by NCA-8122.
If the jurisdictional authority also performs duties as an authorized inspection agency, a representative of the National Board of Boiler and Pressure vessel Inspector s or a representative of the facility will participate in the evaluation.
If jurisdictional laws do not require inspection or permit inspection i
personnel to participate in the evaluation of the quality control system, then the evaluation will be performed by a representative of the National Board or the Society.
Past codes did not provide for a quality assurance program for Class 1 and 2 construction, nor for a quality assurance system for Class 3 construc-tion, as required by the current code.
Although an integrated program or system was not required by past codes, many quality assurance features were required.
Although the program or system was not specifically required, neverthe-less, construction organizations typically did operate under "in-houso" quality assurance programs which provided for the inspection, testing, and j
surveillance of components and construction activities.
l Cesign organizations did not typically operate under an integrated program.
Two nuclear plants were reviewed by the author as part of the design adequacy task of the Reactor Safety Study.*
Approximately 20% of the items reviewed for one plant either did not fully comply with. the FSAR criteria or were not adequately documented for assessment.
Similarly, 40% of the items examined for the other plants could not demonstrate full compliance with FSAR criteria.
- Appencix X to the " Reactor Saf ety Study - An Assessment of Accident Risks in U.S. Ccmmercial Nuclear Power Plants," WASH-1400, USAEC, Draf t August 1974.
l i
I j
A-46 i
C Franklin Research Center l
<c m n rw m saua
It is recommended that the quality assurance program used in both the design and construction phases for each SEP plant Class 1 and 2 item should be compared with the current requirements previously outlined.
If the comparison shows a weak or non-existent program with design and/or construction phases, then the operating history of the plant should be examined to determine the frequency and origin of incidents in which the pressure boundary has been breached.
If subsequent repairs or replacement of the breached boundary have not provided a permanent fix, then it is reasonable to conclude that a design deficiency exists. The following would then be ra; commended:
1.
a design review of the deficient area with design change recommendations 2.
a technical audit to determine design adequacy of selected Class 1 and Class 2 items for the complete plant.
4.1.3 Quality Group Classifications [6]
Nuclear power plant components are currently classified as Class 1, 2, 3, MC, or CS.
Class MC and CS are for metal containment vessels and core support structure s and are outside the scope of this study. Current classification standards are as follows:
Quality Group A (Class 1)
A component of the reactor coolant pressure boundary is currently designated as a Class 1 component.
Quality Group B (Class 2)
Ccmponents are currently designated as Class 2 provided that:
1.
They are not part of the reactor coolant pressure boundary, but part of:
a.
emergency core cooling g/ stems, posr-accident heat removal systems, post-accident fission product removal o.
reactor shutdown or residual heat removal systems A-47 g
...; Franklin Research Center
%, s w %,
v.
e c.
BWR main steam components described in Reference 2:
main steam line from second isolation valve to turbine stop valve main steam line branch kines to first valve main turbine bypass line to bypass valve fir st valve in branch lines connected to either main steam lines or turbine bypass lines d.
FW3 steam generator steam and feedwater systems up to and including outermost containment isolation valves and connected piping up to and including the first valve that is normally closed or capable of automatic closure during normal reactor operation systems connected to the reactor coolant pressure boundary not e.
capable of being isolated f rom the boundary by two valves normally closed or capable of automatic closure during normal reactor operation.
2.
They are part of the reactor coolant pressure boundary, but are not designated as Class 1 because either the component is not needed for safe shutdown of the reactor in the event of an accident or the component can be isolated by two valves as described in footnote (2) of Section 50.55a of Reference 2.
Quality Group C (Class 3)
Class 3 components are not part of the reactor coolant pressure boundary, nor designated Clacs 2, but are part of:
1.
cooling water and auxiliary feedwater systems important to safety, l
such as emergency core cooling or post-accident heat removal 2.
cooling water and seal water g/ stems that are designed for functioning of components important to safety, such as cooling water systems for reacter coolant pumps, diesels, and control room 3.
systems connected to the reactor coolant pressure boundary that are I
capable of being isolated from the boundary by two valves normally closed or capable of automatic closure during normal operation 4.
systems not previously defi..ed, other than radicactive waste management systems that contain or may contain radioactive material, and whose postulated failure would potentially result in off-site doses that exceed 0.5 rem.
A-48 4
.KJ Fran'4in Research Center A >w o. rv
<mu,
Comparison with Past Codes The past 'B31.1 (1955) piping code does not designate quality classes for piping or valves.I '
Comparison of the component classification designa-tions in the FSAR with the standards previously described for each SEP plant is required beforc a comparison with current code requirements can be initiated.
The past pressure vessel code, ASME III (1965), designates Class A vessels which are essentially equivalent to currently designated Class 1 vessel. Class B vessels designated in accordance with the past code would currently oe classified as MC vessels, wnich are outside the scope of this review. Previously designated Class C or ASME Section VIII vessels may be currently classified as Class 2 or Class 3 vessels. Vessels previously classified as Class C but currently classified as Class 2 should be evaluated carefully against current Class 2 requirements sucn as the quality assurance program.
4.1.4 Code Stress Limits Strength Theories Past codes (4,5], except ASME III (1965) for Class A vessels, have been based on the assumption that inelastic behavior begins when the maximum principal stress reaches the yield point of the material, S.
It has been y
commonly accepted that both the maximum shear stress theory (Tresca crit 9rion) and the maximum distortion energy theory (Mises criterion) are much better than the maximum principal stress assumption in predicting yielding and fatigue failure in ductile metals. Although most experiments show that the Mises criterion is more accurate than the maximum shear stress theory, the present code [1] uses the maximum shear stress theory of strength for Class 1 components because (i) it is more conservative, (2) it is easier to apply, and III (3) it facilitates fatigue analysis. Class 2 and Class 3 components continue to be designed in accordance with the maximum principal stress assumption.
1.
Code Case N-1 classifies piping into two categories:
nuclear piping, designed to contain a fluid whose loss from the system could result in a radiation nazard to either the plant personnel or the general public; and conventional steam and service non-nuclear piping.
2.
Except for Class 2 vessels designed in accordance with the alternative rules of NC-3200.
4 A-49 1.0 Franidin Research Center A emen a n.n.wn u.
If the principal stresses at a point are c1 > c2 > 03, then yielding occurs when:
I (1/2) Sy max = (1/2) (et - o3)
=
according to the maximum shear stress theory. For convenience,'the present code uses the term " stress intensity," which is defined as 2Tmax = the largest algebraic difference between any of two of the three principal stresses.
Examole: Consider a thin-walled cylindrical pressure vessel or pipe, away from any discontinuities and subjected to an internal pressure, p, which induces a hoop stress e and an axial stress 1/2c. The three principal stresses in descending magnitude would be:
c1=
c C2= (1/2)c c3 - -p According to the current code, the " stress intensity" ist (c + p)
(r1 - c 3) which together with the stress limit controls the design. According to past codes, the design would be controlled by the maximum stress together with the stress limits used in the past codes.
Stress Categories The current code recognizes the advances in computer-aided structural analysis capability which enable a more ccmprehensive and detailed determina-tion of stress and strain fields, in both the elastic and plastic states due to thermal as well as mechanical loads, gross structural discontinuities, and local structural discontinuities such as small holes and fillet radii.
Accordingly, the current code recognizes various stress categories defined in NB-3213 of Reference Ib and briefly summarized as follows:( }
1.
Primarv Stresses Any normal or shear stress induced by an imposed load which is necessary to satisfy equilibrium between the external and internal 1.
See Figura NB-3222-1 [lb].
A-50 LL Frankhn Research Center w orm w wasa
forces and moments. A primary stress is not self-limiting. The existence of primary stresses in excess of the yield strength across the thickness of the material will result in failure due to gross distortion or rupture, inhibited only by the strain hardening characteristics of the material. Primary stresses are further categorized as:
a.
General Membrane Stre ss.
The average primary stress across a solid section excluding the effects of gross and local di scontinuitie s.
The six stress components associated with a primary general membrane stress are symbolized by P
- m b.
Local Memorane Stress. The average stress across any solid section induced by a combination of mechanical loading and gross discontinuity which may produce excessive distortion when transferring the load from one portion of the structure to
- another, e.g.,
in the crotch region of a piping tee due to internal pre ssure. The stress components associated with a primary local membrane stress are symbolized by P.
L c.
Bending Stress.
That component of a primary stress which is proportional to the distance from the centroid of a solid section, excluding effects of gross and local structural discontinuities, e.g., the bending stress across the thickness of the central region of a flat head of a vessel due to internal pressure.
The stress components associated with a primary bending stress are symbolized by P.
3 2.
Secondary Stresses Secondary stress is a normal or shear stress induced by an imposed strain field necessary to satisfy compatibility and continuity requirements within the structure. Secondary stresses are "self-equilibrating" and limited by local yielding and minor distortions so that f ailure due to secondary stresses induced by the application of one load will not occur.
Secondary stresses are further categorized as follows:
a.
Secondary Exoansion Stre sse s.
Induced by the constraint of free end displacements due to gross structural discontinuities, such as the stresses in a piping element of hot piping system whose ends are constrained; does not apply to vessels. The stress components of the expansion stress are symbolized by P.
e b.
Secondary Memorane and Bending Stress.
Occurring at gross structural discontinuities and caused by mechanical loads, pressure, or differential thermal expansion, symbolized by Q.
A-51 iJ Frank!in Research Center s o~uw s ne r ones eau,
3.
Peak Stresses Peak stresses are induced by local discontinuities such as notches or thermal loads in which the expansion is completely suppressed, such as the local thermal expansion coefficient of the austenitic steel cladding of a carbon steel component.
Code Stress Limits for Material Other Than Bolting Class 1 Components Current code stress limits depend on the code class and service levels being considered. Design stress intensity values, S,, for Class 1 compo-nents are given in Tables I-1.1 and I-1.2 of Appendix I of Reference le for ferritic and austenitic steels, respectively. For materials other than bolting, the design stress intensity value S is essentially the lower of 1/3 (UTS) or 2/3 (YS) at design temperature for ferritic steels.I I Fo r austenitic steels, S is the lower of 1/3 (UTS) or 0.9 YS at design tempera-oom temperature.I I ture or 2/3 (YS) at Assuming that S, is essentially the lower of 1/3 (UTS) or 2/3 (YS),
then the stress limits for the various service level loads and stress category combinations for materials other than bolting may be summarized as follows:
1.
Design Condition (See Figure NB-3221-1 (lb])
Stress Category Limit of Stress Intensity Primary Stresses Tabulated YS UTS P,
S 1 2/3 (YS) i 1/3 (UTS)
P 1.5 S S YS 1 1/2 (UTS)
P
+P
(
b m
I i
1.
See III-2110(a) of Reference le.
2.
See III-2110(b) of Reference le.
A-52 M
U6') Franklin Research Center A em.on or n. Fwa mu.
.,l-2.
Level A and B Service (Operating and Upset Conditions)
(See Fig. NB-3222-1 [lb])
Stress Category Limit of Stress Intensity (a) Expansion Stress Intensity Tabulated YS UTS Pe (not for vessels) 3 Sm 5.2 YS
< UTS (b)
Primary and Secondary (1)
PL+Pb+Pe+Q 3S 52 YS 5,UTS m
(c) Peak Stresses (2)
(3)
PL+Pb+Pe+Q+F S
(See fatigue curves, a
Fig. I-9.0, Reference le) 3.
Level A and B Service Limits for Cyclic operation (NB-3222.4)
Unless the analysis for cyclic service is not required by NB-3222.4 (d) (1) through NB-3227-4 (d) (6) [1], the ability of the component to withstand cylic service without fatigue failure shall be demonstrated by satisfying the requirements of NB-3222.4 (c) as follows:
a.
Determine the stress difference and the alternating stress intensity, Sa, for each condition of normal service, b.
Use stress concentration factors to account for local structural discontinuities, as determined by theoretical, experimental, photoelastic, or numerical stress analysis techniques. Experimental methods shall comply with Appendix II-1600, except for high strength alloy steel bolting, for which NB-3232.3 (c) shall apply. The fatigue strength reduction factor shall not exceed 5, except for crack-like defects and for specified piping gecmetries given in NB-3680.
c.
Design fatigue curves in Figure I-9.0 for the various materials shall be used to determined the number of cycles Ni for a given alterna-ting stress value (Salt)i. The alternating stress determined f rom the analysis should be multiplied by the ratio of the modulus of elasticity given on the design fatigue curve divided by the modulus of elasticity used in the analysis before entering the design f atigue l
curve.
1.
3S may be exceeded provided the conditions of NB-3228.3 are satisfied.
m 2.
For cyclic operation.
3.
2 Sa for full range of fluctuation.
l l
A-53 Ah 2 1...' Franklin Research Center son.sona mv m enue
d.
Cumulative usage for multiple stress cycles is be determined from U = Sum of (M /N )
i i where Mi is the expected number of cycles associated with (Salt)i and Ni is the corresponding number of cycles from the design fatigue curve. The cumulative usage f actor U shall not exceed 1.
4.
Level C (Emergency Conditions)
(See Fig. NB-3224-1 (lb))
Stress Category Limit Tvoe of Analysis Primary Stresses IlI Elastic Pm (pressure and 1.2 S or YS m
mechanical)
Pm (pressure - only 1.1 S or 0.9 YS(l)
Elastic m
for ferritic material) 1.8 S or 1.5 YSil)
Elastic PL m
0.8 (collapse load)
Limit PL+Pb l 8 Sm or 1.5 YS(l)
Elastic 0.8 (collapse load)
Limit 4.8 Sm Triaxial Stresses (2)
Secondary / Peak Evaluation not required 4
Bolting Material Stress Limits - Class 1 Components (NB-3230)
J Design Conditions Pressure-retaining bolts are designed in accordance with the procedures of Appendix E (le), which account for gasket materials and design as well as l
bolting material stress allowables given in Table I-1.3 of Reference le, which are based on the lower of:
1/3 (YS) at rocm temperature 1/3 (YS) at design temperature (up to 300*F).
4 5
1.
Whichever is greater.
2.
Based on sum of primary principal stresses.
i A-54 s.L Frankhn Research Center h w nw w em,=
r
,_.,-,,,,-m.
Level A, B, and C Service Limits (NB-3232)
Actual stresses in bolts produced by a combination of preload, pressure, and differential thermal expansion may exceed the allowables given in Table I-1,3 as indicated below:
- a. Average stress (neglecting stress concentrations) shall not exceed 2 times the Table I-1.3 (le] values, (S )
i 2 (YS) b avy 3
- b. Maximum stress at bolt periphery (or maximum stress intensity if tightening method induces torsion) due to direct tension and bending shall not exceed 3 times the value given in Table I-1.3 (le),
(S )
1 (YS) b max Fatigue Analysis of Bolts Fatigue analysis of bolts is required unless all the conditions of NE-3222.41(d) (1) are satisfied. Suitability for cyclic service of bolts shall be determined as described in NB-3222.4(e) and as follows (NB-3232.3):
a.
Use the design fatigue curve of Figure I-9.4 (1) using the appropriate fatigue strength reduction factor described in NB-3232.3 (c) for bolting having less than 100 ksi tensile strength, b.
For high strength alloy bolts, use Figure I-9.4, provided that (1) the nominal stress due to tension and bending does not exceed 2.7 Sm for the upper curve or 3.0 Sm for the lower curve, (2) the minimum thread root radius is not less than 0.003 inches, and (3) the ratio of the shank fillet radius to the shank diameter is not less than 0.060.
c.
For bolting having less than 100 ksi tensile strength, use a fatigue strength reduction factor of 4.0 unless a smaller factor can be justified by analysis or test.
For high strength alloy bolts, use a fatigue strength reduction factor not less than 4.0.
A-55 O
..'.; Frankhn A ssearch Center s >sen ne name mue
o Code Stress Limits - Class 2 and Class 3 Comoonents Design Allowable Stress values I1I Design allowable stress values are given in Table I-7.0 for Class 2 and Class 3 and in Table I-8.0 for Class 3 component materials. The se de sign allowable stress values are limits on maximum normal stresses rather than the stress intensity values for Class 1 components.
1.
Ferritic Steel Non-Bolting Materials Design allowable stress S for Class 2 and 3 component s as detailed in III-3200 [le] for ferritic steel non-bolting materials is the lowest of:
1/4 (UTS at room temperature) 1/4 (UTS at temperature) 2/3 (YS at room temperature) 2/3 (YS at temperature).
T 2.
Austenitic Steel Non-Bolting Materials The stress allcwable for austenitic steels is the lowest of:
1/4 (UTS at rocm temperature) 1/4 (UTS at temperature) 2/3 (YS at room temperature) 0.9 (YS at room temperature).
3.
Boltina Materials i
Design stress allowables for bolting materials are based on the same criteria as for non-bolting materials, except that for heat-treated bolting naterials, the allowable shall be the lower of:
1/5 (UTS at rocm temperature) 1/4 (YS at rocm temperature),
i t
1.
Except for Class 2 vessels designed in accordance with the alternative design rules of NC-3200, where stress intensity limits are based on Table I-1.0, i.e., the same as for Class 1 components.
gg A-56 h Frankin Research Center i
s cm m vv a +au,
--.-..-,.__-_-______-__1_--
Level D (Faultad Condition) (Appendix F of Reference le)
The rules for evaluating level D service conditions are contained in Appendix F of Reference le.
Only limits on primary stresses are prescribed; thermal stresses are not considered. When compressive stresses are present, component stability must be assured. The potential for unstable crack growth should also be considered.
Component design limits on primary stress intensities for level D conditions depend on whether the system has been analyzed elastically or inelastically.
Elastic System Analysis For an elastic system analysis, the component design limits for level D conditions permit plastic deformations based on loads or stresses determined by:
a.
Elastic Analysis:
in wnich the computed primary stress appears to exceed the YS by as much as 60% but remains within 70% of the UTS, except for piping in which the pressure does not exceed two times the design pressure, in which case the primary stress computed by Equation 9 of NB-3652 should not exceeed 3Sm (2 x YS).i b.
Collaose Load Analysis:
in which the level D loads do not exceed 90%
of the collapse load determined by either a lower bound limit (1) analysis (which assumes an elastic-perfectly plastic material), a plastic analysis which accounts for the strain-hardening characteristics of the material, or by experiment, c.
Stress Ratio Analysis: which is a pseudo-elastic analysis method utilizing the techniques and curves given in Appendix A-9000 (le), in waich the apparent stre ss( 2) is limited to the lesser of 3 S or m
0.7 S except when the methods of A-9000 (lel permit higher limits n
when the type of stress field is taken into account.
Inelastic System Analysis When a system is analyzed inelastically, the level D primary stress or load limits for components permit plastic deformation depending on the component analysis method as follows:
1.
A load which is in equilibrium with a system of stresses which satisfies equilibrium everywhere, but nowhere exceeds the YS at or below the collapse load.
2.
Computed value of stre ss assuming elastic behavior.
A-5
4
.... Franklin Research Center a c~.ea or N % -,
a.
Elastic Analysis:
in which the computed primary stress intensity is limited to the greater of 0.7 UTS or YS + ((UTS - YS)/3).
b.
Collapse Load Analvsis:
in which the load is limited to 90% of the collapse load. The collapse load may be determined by one of the three methods previously described.
c.
Stre ss Ratio Analysis: as described previously.
d.
Plastic Instability Analysis:
in which a plasticity analysis is used to determine the load, PI for which the deformation increases without bound. The load P is limited to 0.7 P or YS + (S - YS) /3 r
g where Sr is the true effective stress ascociated with plastic instability.
e.
Strain Limit Load Analvsis:
in which the load P is limited as de scribed in (d) but not to exceed Ps associated with a specified strain limit.
f.
Inela stic Analv si s:
in which primary stress is limited as in (a).
Comparison with Past Codes The fundamental differences between current and past codes with regard to stress limits are summarized as follows:
1.
The current code for Class 1 items is based on the maximum shear stress theory of failure.
ASME III (1965) is based on the same theory for Class A (equivalent to Class 1) ve s sels.
331.1 (1955) piping code is based on maximum normal stress theory of f ailure.
2.
The current code for Class 2 and 3 items is based on the same theory of failure as past codes.
3.
The current code for Class 1 items considers primary as well as secondary stresses and peak stress categories, as does ASME III (1965). B31.1 (1955) power piping code does not consider peak st re sse s.
ASME I (1965) considers (for piping) primary membrane stresses due to pressure only, except for mitered bends where the required thickness for a straight pipe is multiplied by a factor, (k - 0. 5) / ( k - 1. 0 ), where k is the ratio between the radius of the bend (f rom center of curvature to center of pipe) to the inside radius of the pipe.
4.
The current code for Class 2 and 3 vessels consider s primary stresses for si:e selection, as does ASME III (1965).(1)
The current code for Class 2 and 3 piping consider s primary and secondary stresses, as does the past piping code.
1.
Unless the vessol is designed in accordance with the alternative N3-3200 rules which are based on primary, secondary, and peak stresses.
A-58 4- -w acJ Franklin Research Center s w w m meu,
5.
The current code gives stress limits for the design condition as well as for service levels A and B which are equivalent to past code requirements.
6.
The current code gives stress limits which permit large deformations in the region of discontinuity that may require repair for service level C and overall gross deformations that may require replacement for service level D.
The equivalent of service levels C and D was not specifically considered by past codes.
The FSAR, however, does consider a design basis accident which would be the equivalent of service level D and the stress limits given in the FSAR may be conservative, when compared to current stress limits. Stress limits for the equivalent of service levels C and D should be examined and evaluated based on the information given in the FSAR for the plant being evaluatec.
4.1.5 Welding Requirements Welding materials must currently satisfy the qualification requirements of Section IX of the ASME B&PV Code as well as the mechanical property and chemical analysis test requirements of NB/NC/ND-2430 [1].
A determiaation of delta ferrite shall be performed for A-No. 8 weld material (see CW-442 of ASME IX) except for SFA-5.4, Type No. 16-8-2 and filler metal to be used for weld metal cladding.
A-No.8 weld material would typically be used to join chrome-nickel austenitic stainless steels such as SA-312 Grade TP 316.
The minimum acceptable delta ferrite shall be 5FN and re sults shall be included in the certified material test report.
Full radiographic examination of vessel welds is currently required, depending on thickness of materials joined, weld joint category (see NB/NC/ND-3351 (1)) and code class as discussed in Section 4.3 of this Appendix.
Full radiographic examination for piping, pumps, and valves based on current and past codes, depends on weld joint category, pipe size, and code class as discussed in Section 4.2 of this Appendix.
It is concluded that past welding requirements for vessels were more severe than current requirements, but past code requirements for piping, pumps, and valves were not as severe as current requirements for Class 1 and 2 components.
A A-59
... Franklin Research Center
- c e onNr ma,n.w.
t It is recommended that the FSAR be carefully examined for radiography requirements for pipes, pumps, and valves which would currently be classified as Class 1 or 2.
It is also recommended that welded components and systems in SEP plants made from austenitic stainless steel be spot-checked to determine evidence of hot short cracking in the weld region unless evidence of the use of A-No.8 welding rod with at least 5FN delta ferrite can be provided.
?
4.1.6 Desian Considerations for Bolted Flange Connections Appendix XII of the current code [1) provides supplementary information j
to prevent leakage in bolted flange connections with unusual features such as a very large diameter or under unusual conditions such as high pressure, high temperature, or severe temperature gradients. Appendix XII permits analysis of the joint which considers changes in bolt elongation, flange deformation, and gasket load that can take place upon pressurizaticn and that may indicate a required bolt preload greater than 1.5 times the design val'ue.
This practice is permitted provided that excessive flange distortion and gross crushing of the gasket is prevented. Bolt relaxation under high temperatures should also be investigated. Methods for assuring adequate bolt tightening for large diameter bolts are discussed in Appendix XII.
Past codes did not consider special situations as described above. The I
current considerations of Appendix XII may be useful in evaluating problem connections.
4 l
l i
J I
J t
A-tic a.~ e FranMn Research Center
. swamrm we l
i
4.2 PIPING The current Class 1 piping design requirements are given in NB-3600 of Reference Ib.
The fundamental differences between current and past require-ments are that:
1.
The current code explicitly considers and evaluates the margin against fatigue damage by a for=ulation for peak stress which accounts for local as well as gross dis-continuities. The secondary stress indices C in the current code are equivalent in principal to the stress intensification factor i of the past code [4]. The current code magnifies gross discontinuity stress by sultiplying C by a local stress index K.
The past code (1) considers the effect of cyclic loading by reducing the allevable expansion stress by a factor f which varies between 1.0 for less than 7,000 cycles and 0.5 for more than 250,000 cycles.
Figure A4-2 shows a plot of the allowable expansion stress based on the past code and labelled past " fatigue" curve super-i= posed against the design fatigue curves for carbon, low alloy, and high tensile steels (Fig. I-9.1 of Reference 1e) of the present codes, labelled current fatigue curve.
The past " fatigue" curve is based on a 70 ksi ulti= ate tensile strength (UTS) material whose allowable stress range, S,(2) is f (1.5)(UTS)/(4) (0.9) where 0.9 accounts for 3
the eff,1ciency of a welded joint, and f depends :n the nu=ber of cycles as shown in Table A4-7.
The figure also indicates a value K (cycles), which is the ratio of the present over the past fatigue allowable alternating stress for a given nu=ber of cycles. K varies between K(10) = 25 and the K (1,000,000) 1.0.
Notice that K is the allowable local stress index for
=
a design which is based on the past code and being evaluated in light of the present code, all other things being equal.
Assu=ing that for most piping syste=s the eaxi=um local stress index is not likely to be higher than 5, but higher than 1.4, we conclude from Figure A4-2 that piping systems designed in accordance with the past ar.d the present code:
a.
are conservative for services with less than 500 cycles b.
possibly are unconservative for services with cycles greater than 500 but less than 100,000 c.
are probably unconservative for services with more than 100,000 cycles and significant load changes.
1.
B31.1 (1955) only; ASME I (1965) does not explicitly consider cyclic loads.
2.
SA = f (1.25 Se + 0.25 S ).
Using Se approximately equal to Sh and h
Sh f. 0.9(1/4 UTS) gives SA 5,f (1.5(UTS)/4)0.9.
A-61 D
ED Franklin Research Center ao e N r man.
. _ _ _ _ _ ~ _ _
10'
- -d hb G
en
/$'
PAST CODE COflSEllVATIVE' PAST CODE f.1 AY 4-y
,4 m
m (LESS THAtl 500 CYCLES)
BE UtlCOilSEftVATIVE PAST COD 5PHOBf Y
8 n
10
,UtlCOilSERVATIVE -
3
.3 K(10)- 25 (DETWEEN 500 AtlD 100,000 CYCLES)
FO'H SIGNIFICAtlT 4
~6 LOAD Cl(AllGES g ClIHREriT FATIGtiE CilRVE (LtOHE THAN un N
100,000 CYCLES)
~
~
N
[
g K(500) - 5.0 a
8
.!r, j
K(10 )- 1.4 N
K(1000) - 3.75'
% 'N a
~
K(10*) - 1.8 K(10*) - 1.0
- PAST "FATIGtlE" CtlHVE -
30 24
'N g
20 3
~' ' '"-
~~I 17' 32' 3 f 14i
=/
b iO' I
I l_Li l.11 I
I l_1.1.Lij l
l_LLLill l
I I II Lll i
I I 1111L 8
8 to 10' 10 10*
10 10*
tlOTE: E - 30 = 10* kal U TS 4 80.O ksi
--- UTS 115.0130.0 kal Interpolate for UTS 80 0-115.0 kal Figure A4-2.
Design Fatigue Curves for Carbon, Low Alloy, and Higli Tensile Steels (For !!et al Temperatures tiot Exceeding 700*F) (Reference 4e) t 9
e
2.
The current code considers the influence of ther=al gradients through the thickness of piping elements, together with the effects of the range of pressure and moments due to changes in service temperature and pressure, when determining the peak stress intensity S.
3.
The current secondary stress indices C are either equal or less than twice the corresponding stress intensification factor i of the past code. This implies that expansion stress computations based on the past code are conservative from the viewpoint of margin against excessive distortion.
NB-3653.2 gives a simplified expression for S which conservatively p
estimates the sum of primary and secondary and peak stresses as follows:
PD D
S
=KC
+ gC2NM
~
p 11 2t i
2(1 - v) K r3"alaTl l
b bl x l a,T, - 2
+KCE T
33g 4
(1 - v) Ea l aT,, l (1)
+
where:
K,K 'K
= 1 cal stress indees 2 3 AT = linear portion of thet=al gradient through the thickness AT,, = non-linear portion of thermal gradient through
~
the thickness M, = resultant range of =o=ent due to service changes in temperature laTy or mechanical loads such as
~
earthquake C,C,C
= secondary stress indices 7 2 3 P, = range of service pressure v = 0.3 Ea = modulus of elasticity times the mean coefficients of thermal expansion D = outside diameter of the pipe o
h_
A-63 Mu Franklin Research Center 4 ow w r wwa -
t = nominal wall thickness I = sectional roment of inertia T,(T ) = range of average temperature on side a(b) of 3
gross struc~ ural or material discontinuity.
Values of stress indices for the various piping elements are given in Table NB-3682.2-1 of Reference Ib and reproduced as Table A4-8.
For the purpose of the discussion which follows,(1) the fo'urth term in the expression for S is neglected since it is atypical._.
p The pas t piping code [4] sets limits on the first two ter=s in the expression for S which will be derived herewith. Equation 13 of Section 6 of p
Reference 4, neglecting contributions due to torsion, is giveh by:
S
=1 75,Sg 3
where:
g = resultant coment due to change in temperature from the mini =um operating temperature (usually taken as erection te=perature 70*F as noted in Section 619(b) Section 6 of Reference L to the maxi =um nor=al operating te=perature plus cove-ments of pipe ends attached to equipment.)
Note that M =A21' Ib (approx) where:
g aT 1
21
[(T,) - 70* F]
a Z = see:1on zodulus of pipe = (I)/(D,/2) i = stress intensification factor given in Figure 2 in Section 6 of Reference 4 for various piping elements.
l Substituting the expression for Z in S, we obtain:
g D
= 1[g S3 Comparison of the stress intensification factors, 1, given in Section 6, Figure 2 {4] with the stress indices C, given in Table NB-3682.2-1 reveals that C i* *EP#**I**t'i? 1*9
- i*
2 1.
This discussion can be used to compare current requirements with piping
- l designed to 331.1 (1955).
It is not applicable to piping designed to ASSE I (1965).
A-64 USJEnkjin Research Center smawm me
4 Note further that the 11=1: on S is :
C S = #(1.25 S + 0.25 S )
A c
h where:
S = allowable stress in cold condition S = all wable stress in hot condition h
f = stress-range reduction factor to account for cyclic conditions as given in Table A4-7.
Table A4-7 Stress Redue:1cn Fae:or
~
~
No. of Full Te=perature Stress Reduction Cveles over Excected Life Factor. #
7,000 and less 1.0 14,000 and less 0.9 22,000 and less 0.8 45,000 and less 0.7 100,000 and less 0.6 i
250,000 and less 0.5 Note that for ferritic steels, both S and S appt x =ately equal 0.9* (1/4 UTS) h such that:
= 1.5 f(0.9) UTS = 0.34 f(UTS)
S 3
(ferri:1c steel) 4 For austenitic steels, S is approximately equal to y S so that:
h c
5 4
S =.
- L'rs
+1-
- LrS (0.9) 4 4
4 5 4
A a
S = 0.33 f (UTS) g (austenitic steel)
I i
- The factor of 0.9 is used to account for a butt-welded join: efficiency.
l A
}J2EnWin Research Center 4 c m.an
- N %===,=
I 1
i o
Table A4-8 Tame NB.3652.21 SECDON !!! DIY1SION I-SUBSECnON NB TA8LE NB.2682.21 STRESS INDICES FOR USE WITH ECUATICNS LN NB.3630 (Not Appiacab.e for D,/t>100)
Irrtarnal Moment Thwmal Presswt Landinga Laoitig P'ang PrWxts a:4 Joints 8.
C.
K.
8:
Cs K.
C, C.
K.
Stranget >pe. remete far
Q.5 L1 L2' Le LO La LO QJ L7 G) as wowed IS3/16 4 (or 3/r>0.1]
QJ 1.1 L2' Lo Le 2J LO 04 L7 c.rta nnet a.o societ =f f! tunes, we on ftanges 'x soczet =cene narr,as 0.73 2.0 3.0 1.3 L1 2.0 La LO 3.0 Langitucew tstt wees in strasent pepes (a) fhesa OJ La L1' LO L0 La Lo L1 (b) as wcoes t>3/16 et.
0.5 L1 L '.
La L2 IJ 1.0 L2 (c) as metoed ts3/16 in.
aJ L4 2J8 Lo L2 1J LO L2 Tamered trarmtion arts ser 'IB-a413 and 0. 'f S-42331M 9
(a) fa.sn er no grrtn we+4 ctes4e twin vn OJ L2 to L1 Lo L1 (b) as accee QJ L2 La La LO L7 Branca con vetens se NS.3643ua 1,g 2,3 L7 L3 Lg L7 Curv=o moe w wtt we.een, ecows pr 2h ANS1 Sik9. ANS1316.24 LO gg, e
La LO 0.5 LO w W$$ SP44 u I'#*f*
8.stt.weiong.tses per ANSI 316.9 ar MSS SP44 u La LS 4.0 e
LO 1.0 13 L3 3utt-= ecog reoucers per ANSI 31 A9 w M55 574eu LO
'8 LO Lo 0.5 Lo NOTES:
it)tel *Ne veewee of f,
inoan *or mese s-a m are tel lf 0
.e - 0,. 's act wester men 0.CS D,. ans ae.
aoss. cam.a lor :oraconents nets out of rowner eas not cretsoie we.ue of g, mov ee cota.ead by mwtasaving the yeetse 'naa 0. Cat =cere out et roundaess.s cet: nee as
.sou.ates waevet of X, by the factor 7,o:
Osutea
- ENa, *00 0,... = 'runitruarn outs.oe aewter of woe sectson. in.
- ' * * '
- A c,2r O
- neaivawm outgico ciserstar of crom section,in.
- novennas me.1 inicmaeus..n.
apere M = 2 for terretse stae's and aonferrows ears escent n.c ederwron aslovs ane M If Tee Goes lect'en 's out si rowth$ eWCfl f%et trte (Tots anches.oronerome 48 tov $
sect, a.s.ocrti..
twv
.cnic.i..n enevei..
of u - L7 to, e,,eren,t.
e s.,,cevero.wn,on
( mov De Cota. Pad DV NIuft'o'F'ng tne Laos &ated W4wes Fovt, ed a.cket-ston cr*fo*Flowm salovg af e, ir, - e scor e...
sy. v.e.a sirenam at se.,a weermee.==
a (Tec.es 1.LO3
- 3.,, - o _,, '
is 1
7 ces ea e nan. s=
- r.. i -
, d,s.,
o,,,,,,,,.
f.
,a, an, w.
f )
7.,
1 *C 465 g23 weios en acecroame aim ao roovere wnes af tes suc ect.en.
.ae. O,
-er a.i ou rs.oe o.a,neter oa.
- w. pw,.e.o, ne oee.a.4 es inoie me.as n.ca se e e,een row a on osu..a.enw.no ener.or i.,r4 ace to re. move
.e..nier ei arewwre..,.
a
'wte 9easiewr9 waswe 30 prosaufw.R f Ple 404G esped irregw.antiet and survyt martges in Jetour dwe to ewC'e wnGar Cartsatteretsore Thurstw9ss af we'd ren1Fertsr'tBat (totaf I a =uwaws as es<st.c,r, at metenas at aoom *se.
,,*see ano outs ael s%i so 6scesa o it. *#o concavery o-sewee. 09 on ine, or we.. ce,nscies. 7 e..aees con ove i.wi cuwe eves ne o.e.neo n i.i.
no
.co na e. s.oc. :an,e me.s so eram ae,ea, io N
4 A-66 M F snk.'in Resesrch Cer ter a : - n a w.sn e=n.e
i Table A4-8 (Cont.)
NB4000 - DESIGN Table.NS FA2.21 turface of c.co or, on tacerns transstion sace of we+d. to For Me tfie nomsnas tranetson wrfacel greeter inen 7 des see samen below.
Mg e v'M*,, *M *,, *W',,
- reeustant moment on tirenen I **b '****
For M, 7 aos man.
g M, e VM',,em',,.w',, a roo.itant moment un run I
T-I
.iner M,,, M,, ans M, rare omenenes as teso==:
j f7 duS men, g 7 ces man.
If Mg, anti M, neve tne same aigsare.c s.gn, snen M,,4. Il i
M, and M,,
nave def ferent a.geora.e s. ens, toe. f w;,.e tne 7
N f
amener of M, or M, anere s = e, r, s.
i s
f f
For brancn connections of tees tne M term of Eauss.ons i
t&) As.svees >s ce%ned a==.cs not neering tne soscias 191.110),1111 or 1121 sness be reonoces av ene fono ng reau.rements for nuwe.esset At ene.avernecpon of a amers of terms:
lonytudinal Dutt evend in strengnt poos vusta e prtft Outt oseed or prtft fillet eased, M.
g
- 8 r Ze Eeustion (9) 8,a Ze 8,
- 0.5 ene 4,
- 1.0 e,. $, *e,,w.
N e,. 4, e,. <,.an. <,.noiseinmin. m.cro.uet E enons non a itu of me mve neces for tne sorgruainas weed and
'a
'e prive asne. For enamoee, at tne meersection of an g,,g,,1 g,,g,, U a>= weed prtn butt weed sain an as aea longituames go,,,,,, n j g tsutt sed. C,.s 1.1 x 1.1
- t.21, C, 'or a giren fines we.d le 2,
invernmenne a sorgtvennes veed sness to tamen as 2.0.
( 1h surve.necas even are sons.omose on'y to brancn anere G3nnecttone in Btro*Ftt Dios vuotit threnCR asas nornes to tfte por surtees and anscn nest ene aanwasiones reau.renunts Ze
- eir'els y, med hastenone of N846BS arme F.g. NS 0646.1 1, f 4) #
- curtett ospe or
- Deus rassus, an.
I, * #Am' Ir a
- mean radan.s of crees section,.n.
= (0,- rU2, venere r e nom. nee sis vn. canes (SI N nsues of moment. Mr. snans be ocea.aee from an Far tronen connectione per NS.36A3 toe Foomote 3 anose Inn, T, #m. anos T, are oefeaed m F q. N S-3686.1 t eiervers of me posag tvivem in accordancs amen NSO672.
3 My se echamit as trte range of treenempt losseng soossed during For tButteseneeng toes oor AN53 016 9 or MS SP 44:
f the nosofies couronag evos, c',,, a nimma resus of ose gnered tirencn moe
, M, r,
- norrenas vene tnecanoes of senaysatas brancn ljA 4,, o m an ressus of eens,naese run moe see.g., rhnso,e p oe e
v3 r, nomine a m.cnnens of aes., sated run moe Mt
- momsat at Po.nt A 8
i$'lattsces are somecanee to taperett treamtion scenen watit a girtn y
i e
i 2
taart need at tee men end of me ersnation.
f,9 9 M, C, a 1.3 + 0 003 (0,r)
- I 5 isnt t
A but not yeever tnen 2.0
- #*'8' " *'U8**
/
C,
- 14
- 0 004 'O,tt
- 3/J I4/rl
/"
y3 Due not greater tnan 2.1 l
"f*""*"'#'*'^
C,
- 1 J + 0.008 ta,rl e
a
(
Wr vu, *w, e, 3 "2
e (71 d,a = 0 75 C a ut not seen vnan 10 c
t i
- ,,
- 0.75 C,, out not less enan 10 l
Bewann Moe c,, e ggs,,,,7,) a n g,,,p,,) o n g r s7,) tr',,r,1, mut not e
i te. m.a 1 s Momenes cascuiated 'or comat at crornectson of run and A,,,
- f. r,,, Pa. anes e, are dehaed in F g.
Orense comer haes N S-3686.1 1
& M,3 t',a
- I O C,,
- O a ( A,,,7,)*#' tr'.,, A
). Out not tous tnen 10 af,,
- 2.0 N proeucs of C,K,,inass he e trea.nmen of 3.0 2 ",3 A FI T
A "r2
- 3) c, *
. put aee tous enan f.S; 8,
- 0.75 C, y
1 re i
e,., anere r = nom.no once ami micenne
/
6 I
A = sene ree.us et cuevos o.oe or e*tio=
g.2 nw/
el
=
,. ~
.oe.e..u.
- ic - tu2
,,st
, 4, g e
I'$
i l
A-67 i.UU Frank lin Research Center
= o on w nie %n in ua
a
__ an w Table A4-8 (Cont.)
Tame.NB-3632.21 SECTION !!!. DIVISION I - SU3SECTION NB (g) d,a
- 8,p
- 0.75 C,e tcf mee,cers in evnaca r, and r, e 0.10, Osa = C,,
- 0.57 f A,/1",;, but not ress then 2.3 Ama evnen rea us of doingnated run poo T, e norrenas med tn.cknem of or%gnated 'run moe C,
- 1 + 0.0C58 ova,te K,a*K,p*12 (10) The X indices van for fittings per ANSI 816.9, ANSI Co
- 1
- M8 8 'lO*I*
,,"J, = 0.51 318.23,or s. ass $7a8 acoev only to scenes fittw ustn no cunnnectsons, artacnte=nu, or otner entraneous stree emesars anere Oct,, e the Larger of D,,t, sus 0, A,.
on me boa.es mereof. For rittings seen engstwenal taart
. sends,the K.ndeces snown snan be erwatics.ed by time 1.1, tar
,ia; m sucers an =euca r, and/or r, <110, e
fesst imeMrs as oefined en Note 2; by I.3 for teees not Preet'at "* r*3u'renents for neee==sian.
C,
- 1 + 0.03485e'#' LO,,,t )***
(11) The strees eno css green creescs stresses
- n. cst oomar in the boev os e fitting. It is not teouwed to tsaa ine creduct of c,.1 + 0.0188svO,re aress indices for two a.o rig proouc s sucn as a tae and a rechacer, or a f ee er48 8 girth butt swete$ sonen weesced ICgetner estere 0,.tg is the large of O,,T, arid 0, A,.
escent for the esse of curwd pas or taalt essening above==*ded togetner or toined try e zece of stra gnt poo ee sength of an.cn es sees than 1 moe o.arrerer. For tnes (tel The K inosess seven in tal, !bl, erie (c) accey for ra'*****
soeerf+e case the stress endes f or ene curwd o.co or butt attecned =o tne connectmg poe vnta Fust ar aewooser wendsrug etbove must be multsossed Dv tnet for tne gwtn thsti girtn vuescs as cef sted in footnots (2). Note mat tne vwend. Esclumed from the erketsC#ecatton are the 5, and C',
comtecnnt gleTn Webd Rhet SM De @ecaed sacWetery U
<=s. css. Their weave es to be: 3.
- 1.0, C*,
- 0 30.
comoeienna (12141 eefe=1 as ene me rruni pernewtWe mrstruten as snovan
(*) F*' rooucars connected to poo witn Muse girth tntet en 84 N64:231. A voue of 6 'ess men 3/32.rt may se
- s'88:
used artraeed too strene missenten i sonoried for g,,,
fearication. Por trusa or, aet.ned m footnote 12),4 mey K.
- 1.1 - a t
, but nas iets tnwi 1.0 De taaen at fero, yTerlas
. (13) taJ Narnanc'ature L,,,
K,
- 1.1 - 0.1 - easseur, tmst not seus t%en 1.0 L,
v Gortas e-ed VO,,c, e -no ensaer of L,/yG and wnere R%%,,',' / -
6 431 For reaucers eteirected to rrce entn ar enres yrth
/
,d
?, r2 butt newts veiers r,, t, > 3/16 in. sid 4,,T,, 4,,t, 4
/
,f 0.1:
ri r,
7 D//?,
m g
g a
I Le l
l K,
- f.2 - 12
, but not teor than 1.0 MO*'m tg Oj2 4
e t
X,
- 1.8 - 43
, tn t not peu tran 1.0 v Gmim t, e ncyninal easi tneeness, targe end
- ne'eE/vdelm is the o' nester of L,/gM and t,
- nornens.voi en.ca mess, ensi and Le/v 0:r -
e 0, e aem nas outs.ee eemetee,'arge sed I
J.
- aornecas outs,ce sewter, vues and Ic/ 8er reducers.ennected to >ce wetn as-meaaect gets l
.. cone 'angie, eeg.
outt wees, enere r, or r, 4 2/16 en. or 4,ir, or 4,,t, >
b 0.11
'
- 4/ TM.co.ces 3,ven.a e s and (d) acory.I me feaosang Le c
conetons are enet.
K.
- 1.2 - a2
, but not neus men 1.0 f ff Cans anee. 4 ooes act e scoed 50 oss and tne 5 0 *'*
l recueer e concentric.
l (21 The we.i taae kr.ers is not less then t,,,, inrous*=mst toe K.
- 15 - 1.5
' tnet not hoss man 10 body of tne rooucar, escsot in and.mmedtstoey
,p adecent to the Cvieracracas ocrtion on the smsd end, asure r is t%ceries smasi riot to ess than t,m. Wad
,g a
inar t nesses r, m avus tem are to be otneined Dv L'/ mNOf*
E ouasion it), N S-341.1.
t 1M Q,;
A-69 dJ Frankin Research Center 4 > on s N rrannan wem.
e Noting that M =AM and conservatively assuming that a nuclear power plant g
b designed in accordance with past codes is such that SE" A and recalling that C : 1.9 1, the second term in the expression for S becomes:
3 p
D l D
\\
[M KC A
K 1*9 i
M
=
3 3 2 b
1.9 Sg (A21 2K)=
0.65f A E2 (UTS)
(2a)
=
21 for ferritic steels 0.63f A,1 K3 (UTS)
(2b) for austenitic steels Past piping codes determine pipe thickness in accordance with the for=ula(1}
PD t, = 2(S, +0.4P) 9""E #"
""U "1
- "O*
)'
where:
P = design pressure D = outside pipe diameter i
C = allowance for corrosien S = all wable stress at temperature h
i e = mini =um pipe wall thickness IJhen C is small ce= pared to the thickness and 0.4P is sr.all compared to S, the minimum thickness is approxi=ated by PD0
, =
's 2S Since the actual pipe daickness, t,
is noc less than t, we have l
1 PD
- (UTS)(0.9) ferritic steel 18h" 2t f- (UTS)(0.9) austenitic steel 1.
Based on y = 0.4 for ferritic =aterials below 900*F.
A-69 4
JULb%nklin Research Center
%w % %-
i Assuming that the range of service pressure P is a fraction A of the design t
pressure, we have PD A,PD 1/4 A,(UTS)(0.9) ferritic steels 0 0 0
2t 2e
- As h
1/5 A (UTS)(0.9) austenitic steels 1
s o that the firs't term in the expression for S may be put in the form
[P D \\
10 1 (UIS) K C1 1 (0.9) ferritic steel (3a)
A C
=
1 2t j y
yA (UTS) "gG (0.9) austenitic steel (3b)
Substituting Equations 2a and 3a on Equation 1 and neglecting the fourth term in Equation 1, we obtain:
1 S = - (0.9) A K C1 (UTS) + 0.65f A,,,K.,
(UTS) p r
(1 1) Ec %
3 9
2 3 2(1-v) 1 (la) for f erritic steels.
Similarly substituting Equations 2b and 3b in Equation 1 and neglecting the fourth term in Equation 1, we obtain:
A1 (UTS) K C + 0.63f A.,,
K.,
(_UTS)
S
=
11 p
3 I
r2
+ (1-v) EalaTl+K laTl (lb) 2 3 20 - v) 1 for austenitic steels.
l
{
These expressions can be further simplified by noting from Tables I-5.0 and I-6.0 (lal (*4intar 1973 Addenda) that:
3
-6 Es 2 7.9 x 10 x 7.3 x 10 ksi
~
(1 - v)
- 0.7
.r err c steels op 3
-6 Ea 28.3 x 10 x 9.4 x 10 ksi 0.380 for austenitic steels
=
=
(1 - v) 0..,.
a-A-70
'JDJ Franidin Research Center som %vm m
s
)
Substituting appropriately in Equations la and lb and multiplying the second term by 1.3 to account for movements of pipe ends attached to equipment, we have:
1
+ 0.85f i K, (UTS)
S = 0.23 A1 (UTS) K C11 21 p
+0.291laTl+0.145K3laTl (la) 2 y
for ferritic steels
+ 0.82f A K
S = 0.18 A1 (UTS) K C11 H 2 (UTS) p j
+0.380laTy+0.190K3laTl (lb) 1 for austenitic steels where:
p
=(rangeofservicepressure)/(designpressure)=f A
7 UTS = ultimate tensile strength of material at 70*F i
f = stress-reduction factor (see Table A4-7) th A.,, = [ Change in te=perature f or i service cycle] divided by (maximum operating temperature - 70*F]
l
=laTl/l(T)
-70*Fl l
1 o mx g,C,
,laT l,K,laT l = previously defined.
7 3
3 y
Tae alternating stress intensity, Sg, is one half of the peak stress intensity, S ; that is:
=1 S
S alt 2 p For a given value of alternating stress corresponding to actual n service cycles, the nu=ber of such cycles N allowed may be found from the applicable design t
fatigue curve, Figure I-9.0 [le]. The usage factor for the given a service t
l; cycles is defined as:
l "i
i "Ii i
4 A-71 rd) Franklin Research Center haw awm
e
.~
=-
The cumulative usage factor, U = CU shall not exceed 1.0 as required by f
NB-3222.'4(e)(5) of Reference lb.
Equations la and lb may be used to evaluate _ Cla s 1 piping designed in accordance with past code requirements from the viewpoint of present code re-quire =ents.
Some exa=ples will be used to illustrate use of the formulae.
Examole 1 Consider the 42-in ID primary coolant piping between the reactor vessel and steam generator for the Palisades plant (11). These pipes were fabricated from 3-3/4-in thick ASTM 516, Gr. 70 plate with a rolled band 1/4-in nominal cladding of 304L stainless steel. A review of transient conditions given in Section 4.2.2 of Reference 11 indicates the following step power change service cycles:
1.
15,000 cycles of 10% full load step power changes increasing from 10% to 90% of full power and de-creasing from 100 to 20% of full power
- j-2 2.
500 reactor trips from 100% power.
Examination of Figure 4-8 of Reference 11 shows the reactor coolant temper-ature as a straight line function of NSSS power. Considering the hot te=perature function, note that this full power T = 594*F and at 0% power T = 532*F.
This indicates that the temperature change associated with the reactor trips is 62*F.
For each AT, we shall assu=e that aT = 0.75 AT and AT = 0.25 aT.
3 A = ore accurate determination of ST and aT3 =ay be obtained from Reference y
12, so that:
y = 15,000 AT of Service Cycle 1 = 62'F Se rvice Cvele - 1 n
aT = 0. 75 x 62 = 46.5'F f = 0.3 AT = 0.25 x 6.2 = 15.5'F 2
A-72 Od' nidltt Research Center a cm :. e N rm e.
e Service Cvele - 2 n, = 500 AT of Service Cycle 2 = 62*F aT = 0. 75 x 62 = 46.5'F 7
f = 1.0 AT = 0.25 x 62 = 15.5'F 2
Elbow Consider an elbow in which the bend radius R is 5 times the pipe diameter 2r 2r = 42.5 + 3.75 = 46.25 r = 23.13 R = 5 x 46.25 = 231.25, 2R = 462.5 From Table A4-8 for curved pipe or a butt welding elbow K = 1.0 y
C = (2R-r)/[2(R r))
7
= (462.5 - 23.13) /[2 x (231.25 - 23.13)]
i
(
= 1.06 K, = 1.0, K = 1.0 3
Lenzitudinal Butt '4 eld-S traight Pipe A longitudinal butt weld flush in a straight pipe would be a more critical element to *.nvestigate since for this element:
K = 1.1, C = 1.0, K, = 1.1, K = 1.1 y
7 3
3 ranch Cennections A b*canched connection which may possibly have been used to connect the 12-in Schedule 140 316 stainless steel surge line from pressuri:er to the hot leg would have stress indices as follows:
K, = 2.0, K, = 1.7, K = 2.2, C = 1.5 a
1 1
and obviously would be most critical. These K and C values are taken from the 7
Su=mer of 1979 Addenda (1].
5 ranidin Research Center
- wwwr- -
I t
.~., - -
e 1
Determination of Usage Factors UTS = 70 ksi (ASTM 516 - Gr. 70) ch For the i service cycle:
(S )i 0.23A x 70 x I C1 1 + 0.35 70 *# K 1
=
p 1
2 21 l
+0.291laTl+0.145KlaTl 2
3 1
Asst: ming that the pressuri:er maintcins pressure within p_50 psi during these seriice cycles, then:
100 A1 =,,500 = 0.04 so that (S )i 0.644 K C1+59.5A.,,fK.,+0.291laT
+0.145KlATl
=
p 2
3 1
AT i
Determination of \\
for each service cycle A
=
21
[(r ) ax-70a) r oC (T )
=aximum operating te=perature = 594'F
=
ST of Seriice Cycle 1 = 62*F a! of Seriice Cycle 2 = 62*F 1
- 62/(594 - 70) = 0.12 21 A,,, = 6 2/ ( 5 94 - 70 ) = 0.12 1
finally (Sal,), = $. (S ),
p.
A su=cary of the resulta for each of the two ser/ ice cycles as it af fects the usage of the three elements is given in Tables A4-9 through A4-il.
It is apparent i
fro = the usage factors calculated in these tables that cu=ulative da= age from j
cycles 1 and 2 is negligible.
~
l A-74 4'J Franklin Research Center 3
dud A Drummon W TN Fearwen meaus
4 Table A4-9 Usage Factors Due to Thermal Gradient Through Thickness Example: Hot Leg of Palisades Primary Coolant Piping Piping Element: Elbow K = 1.0, C = 1.06, K = 1.0, K = 1.0 1
2 3
Service Cycle - 1 n = 15,000 f = 0.3 A
= 0.12 1
21 4
aT = 15.5*F AT = 46.5*F 2
0.644 K C + 59.5 f K A
- 0. 291laT l 0.145 K fat l = 18.7 ksi S
=
2 21 2
3 1
=1 S
S
= 9.3 ksi alt 2 p 6
- 1 N > 10 (See Figure A4-2)
U
= 0.02
= g1 7
Service Cvele - 2 n = 500 f = 1.0 A
= 0.12 2
22 AT = 15.5'F AT = 46. 5 *F 2
7 0.644 K C11+59.5fK,A,,+0.291laTl+0.145KlaTl=19.1ksi S
=
2 3
1 p
=1 S = 9. 5 ks i Salt 2 p 6
"'d l
N, > 10 (See Figure A4-2).
U, = g = 0
(
2 l
U
+U
= 0.02 7
2 l
i l
l A -75 A
3bJ5nidin Researen Center s w a m pm emma.
.i d
Table A4-10 Usage Factors Due to Thermal Gradient Through Thickness Example: Hot Leg of Palisades Primary Coolant Piping 1
.=
Piping Element: Longitudinal Butt 'Jeld-Straight Pipe K = 1.1, C = 1.0, K = 1.1, K = 1.1 1
2 3
l Service Cycle - 1 n = 15,000 f = 0.8 A
= 0.12 1
g AT = 15.5*F AT = 46.5'F 2
S = 0.644 K C + 59.5 f K A2 n + 0.291laT l ' O.145 K {aT l = 18.9 ksi 2
3 y
=1 S
S = 9.5 kai alt 2 p i
6 1
N1 > 10 (See Figure A4-2)
U
= 0.02 i
=
Service Cycle - 2 n = 500 f = 1.0 A
= 0.12 2
g AT,, = 15. 5 ' F aT = 46.5'F 7
S = 0.644 Y C + 59. 5 f K.,1,,,
+ 0. 291 l aT,, l + 0.145 K l aT l = 20. 5 ks i p
11 3
1 S
=1 S = 10.2 ksi l
alt 2 p 6
N2 > 10 (See Figure A4-2) p=0 2
U
+U
= 0.02 2
T A-76
_nklin Rese_ arch _ Center
t Table A4-ll Usage Factors Due to Ther=al Gradient Throurh Thickness Exa=ple: Hot Leg of Palisades Pri=ary Coolant Piping Piping Ele =ent: Branch Connection (K and C from her M Menda W) 7 y
K = 2.2, C = 1.5, K = 2.0, K = 1.7 1
2 3
Service Cycle - 1 n = 15,000 f = 0.8 A
= 0.12 21 aT = 15.5 7
~
aT = 46.5 r 2
1 0.644 K C1 1 + 59.5 f K,A,i + 0.291laT l + 0.145 K laT l = 29.5 ksi S
=
2 3
p
=1 S '= 14.8 ksi S alt 2 p 6
"1 N > 10 (See Figure A4-2)
U = 7 = 0.02 y
1 Ser rice Cycle - 2 n = 500 f = 1.0 A
= 0.12 2
22 aT = 15.5 7 aT = 46.5 r 2
1 0.644 K C, + 59.5 f K, A,,, + 0.291laT l + 0.145 K laT l = 25.2 ksi S
=
1.
2 3
1 p
=1 S = 12.6 ksi Salt 2 p 0
n, U, = p. 2= 0.0005 N, = 10 (See rigure A4-2) s
+U
= 0.0205 2
4 A-77 D
Jb Frank!in Research C. enter 4ca aw
>=r -
am
i Examole 2 The same Palisades primary coolant piping will be considered as in Example 1, except that only a branch connection will be considered for service cycles in which there is a more significant change in average metal tempera-ture as follows:
Service Cycle cri 121 (lEr l/(524))
1-ni
(*F) i 1-15,000 (10% to 100% full power) 59*
0.113 2-15,000 (50% to 100% full power) 31*
0.059 3-15,000 (10% to 90% full power) 55' O.105 4-15,000 (100% to 20% full power) 49' O.094 Comparing the above values A with the value of 0.12 obtained in 21 Example 1, the usage factors associated with the above four additional cycles are negligible.
i Cocoarison With ASME I (1965) Requirements Piping f rom a BWR reactor vessel up to and including the first isolation valve external to the drywell could have been designed and built to the j
following requirements:
il a.
ASME I (1965) b.
ASME I (1965) and 331.1 (1955).
If requirement (a) was invoked, expansion stress limits due to cyclic thermal loading are not specifically imposed. However, ASME I (1965) does require consideration of loads other than working pressure or static head, wnich " increases the average stress by more than 10% of the allowable working stress."
For example, the allowable working stress for welded alloy steel SA-250-Tl at 600*F is 11,700 psi.
Expansion stresses would typically be in excess of 1170 psi and should be considered.
Licensees that designed their piping based on ASME I (1965) criteria should furnish details as to how thermal stresses were considered.
t i
I A-78 Q.,a 1
2 'J Franklin Research Center ar e N r u n.u.
If requirement (b) was invoked, then paragraph 102(b) of Section 1 [4]
requires that valves, fittings, and piping for boilers as prescribed in ASME I are within the scope of B31.1, but provisions of ASME I shall govern where tney exceed corresponding requirements of B31.1.
Accordingly, piping built to requirement (b) would have to satisfy the specified expansion stress limits of B31.1 due to cyclic thermal loads as well as the full radiography requirements for all longitudinal and circumferential fusion welded butt joints of ASME I.
Welding Recuirements Full radiography of welded joints in piping, pumps, and valves as stipulated in past (4] and current codes (1,13] depends on weld joint category, pipe size, and code class as shown in the following table Full Radiography Code Requirements for Welded Joints in Piping, Pumes, and Valves Current Codes ASME III (1977)
ANSI B16.34 (1977)
Past Codes (l)
Description of Class Class ASA B31.1 (1955)
Welded Joint 1
2 3
Standard Special
& ASME I (1965)
~
A. Longitudinal Yes Yes No No Yes No B. Circumferential Yes Yes No No Yes No C. Flange connection Yes Yes No No Yes No D. Branch and piping Connections to j
pipes, pumps, and valves of nominal pipe size exceed-t ing 4" as follows (1) Butt-welded Yes Yes No No Yes No (2) Cornet-welded Yes Yes No No Yes No full penetration l
(3) Full penetration Yes Yes No No Yes No 1.
Full radiography of butt-welded joints may be specified under B31.1 (1955) but it is not mandatory.
Full radiography is required for all longitudinal and circumferential fusion welded butt Joints for pipes built to ASME I (1965) requirements.
2.
Except when specified by material specification for piping in excess of 2 in nominal diameter.
3.
When either member thickness exceeds 3/16 in.
A-79 A%
...; Franklin Research Center swamnw nw.
r i
i 4
In conclusion, full radiography was not required by the past code, but it is a current requirement for Class 1 and Class 2 welded joints for piping, pumps, and valves.
It is recommended that welded Class 1 and Class 2 components and systems be checked to learn what radiography requirements were j
j enforced.
j l
4.3 PPESSURE VESSELS The past code requirements for pressure vessels are given in one or more of tne following ASME Boiler and Pressure Vessel Codes depending on the SEP nuclear plant group as defined in Table Al-1.
Group Pressure Vessel Ccde I
ASME III (1965)
ASME VIII* (1965)
II ASME VIII* (1962) i III ASME VIII* (1956, 1959) 4 The current code requirements [1] and the past ASME III (1965) code are essentially the same with regard to significant items with the following l
exceptions:
Fracture Toughness - Class A Vessels The current code, except for exempt materials as noted in Section 4.1, requires greater toughness than the past code.
A comparison of current and past Charpy V-Notch acceptance levels at temperatures at least 60*F below the temperature at which the vessel is to be pressure tested is as follows:
Past Current Minimum Absorbed 15 to 35 ft-lo 50 ft-lb Energy depending on yield strength Mimimum
)
Lateral Not specified 35 mils Expansion "Plus nuclear code cases.
I A-80 d2 d!J FrankJin Research Center 4
= x.aa w wv.= m u.
6 It is recommended that past Class A vessels should be evaluated from the viewpoint of current Class 1 fracture toughness requirements as outlined in Section 4.1.
Fracture Toughness - Class B Vessels (Outside Scoce)
The impact test requirements for Class B vessel materials built in accordance with Subsection B of ASME III (1965) are the same as for Class A vessel materials, except that the maximum test temperature should be at least 30*F lower than the lowest service metal temperature (LST). The current code permits Charpy V-Notch testing at temperatures up to the lowest service metal temperature. The acceptance standard for the C test of the current code, y
however, requires a lateral expansion between 20 and 40 mils and sets no absorbed energy requirement. The current code provides for exemptions from impact testing. Where the exemption does not apply, drop weight testing for materials exceeding 2.5-in thicxness shall demonstrate a nil ductility transition temperature below the LST by 30*F for 2.5-in thick material, and increasing to 87'F for 12-in thick material as show in Figure A4-2.
Class B vessel materiale built according to the past code and evaluated in accordance with the current fracture toughness requirements:
1.
would satisfy current requirements provided the material thickness is less than 2.5 in 2.
may not satisfy current requirements for thicknesses in excess of 3 in (exclusive of cladding) for those materials not otherwise exempt f rom impact testing as noted in Section 4.1.
l Fracture Tcuchness - Class C Vessels Materials for Class C vessels built in accordance with ASME III (1965) were required to satisfy impact testing provisions of ASME VIII (1965) (5).
Paragraph VCS-66(c) of Reference 5 exempts materials whose LST is -20*F or greater. Apparently, impact testing was intended primarily for outdoor vessels. The current code exempts materials for vessels whose LST exceeds i
A-81 4
..J Franklin Research Center a o~ wi a N r2n. %
i 100*F.
Therefore, all Class C vessels built in accordance with the past code should be evaluated in accordance with Section 4.1 Class 3 ( I criteria to determine if current Class 3 requirements would be satisfied.
Design Requirements Class A vessels designed in accordance with ASME III (1965) are based on an analysis which determines the stress distribution in the vessel. Stresses were combined, categorised, and limited in the same manner as is currently required for Class 1 design condition as well as the equivalence of service levels A and B, i.e., for expected operating and upset conditions which the vessel must withstand without substaining damage requiring repair.
The basis for establishing design stress intensity values, S, as noted in Appendix II
[3] as well as the basis for establishing f atigue curves is the same as current code requirements.
In conclusion, Class A vessels designed in accordance with ASME III (1965) would satisfy current Class 1 vessel requirements for the design condition as well as service levels A and 3.
Class A vessels were not, however, required to withstand loading conditions which may produce large deformations in the areas of gross structural discontinuities (service level C) or conditions which may produce gross general deformations (service level D) requiring removal of the vessel f rom service for repair.
The past codes do not specifically consider loading conditions, other than design, cperating, and test.
The FSARs for rpecific SEP plants may, however, consider the equivalent of emergency and f aulted conditions. A discussion of the evaluation of the FSAR stress limits for these loads against current limits is presented in Section 4.1.4 cf this appendix.
Class 3 vessels, as defined by ASME III (1965), are containment vessels, which are outside the scope of this study.
Class C vessels are designed in accordance with ASME VIII (1965) except that:
1.
Class C vessels currently designated as Class 1 or Class 2 should be evaluated against Section 4.1 Class 1 or Class 2 criteria.
A-82
..c) Fran'Jin Research Center a w
- N % w ueve
- 1. the exemptions f rom inspection defined in U-1(g) of Reference 5 are not applicable
- 2. longitudinal and circumferential welds for those Class C vessels which are or may be connected to the reactor coolant or moderator system during operation and Class C chambers in a multi-chamber vessel having at least one Class A chamber shall be full penetration welds and shall be fully radiographed, and shall satisfy the requirements of N-462.1 and h-462.2 of Reference 3 for Category A and B joints, respectively.
Stress limits for Class C vessels which would currently be classified as Class 3 vessels are essentially the same as for Class 3 vessels designed in accordance with the current code.
The past code allowable normal stress was the lower of 1/4 (UTS) or 0.625 (YS) compared with a current allowable of the lower of 1/4 (UTS) or 0.677 (YS). The past code is at least as conservative as the current code.
The current code does set limits on combinations of primary membrane and bending stress at 3/2 S = YS.
Class C vessels which would currently be classified as Class 1 or Class 2 vessels should be evaluated against current Class 1 or Class 2 code require-ments, with special attention being given to current radiography requirements.
Evaluation of past vessels for the equivalent of service levels C and D for stress limits set in the FSAR should be compared to current stress limits for these service levels.
Examole The Palisades FSAR classifies the pressurizer as a Class A vessel and j
defines the stress limit for the design basis accident (equivalent of service level D) as 10% above YS based on an equivalent elastic stress. Current requirements permit computed stress levels to exceed the YS by as much as 20%
for an elastic analysis. We conclude that Class A Palisades vessels satisfy current requirements for level D loads.
l l
A-83
.... FranWin Research Center a re a o. a..m
,t.
I Welding Requirements The following table provides a comparison between current and past code requirements when radiographic examination of butt-welded joints is mandatory.
The values given are thickness limits above which full radiographic examination of butt-welded joints is mandatory.
~
From the table, it can be seen that:
1.
Vessels built to ASME III (1965) Class C requirements and currently classified as Class 2 or Class 3 would more than satisfy the current radiography requirements for joints of Category A or B.
(Refer to NB-3 351, NC-3351, and ND-3351 for definitions (1).)
2.
Joints of Category C (Refer to NB-3351, NC-3351, and ND-3351 for definitions [1]) in a Class C vessel currently classified as Class 2 would have been examined in accordance with ASME VIII (1965) requirements, which do not satisfy current Class 2 requirements.
3.
Vessels built to ASME III (1965) Class A or ASME VIII (1965) would satisfy current requirements for Class 1 and Class 3 vessels, respectively.
It is concluded that current Category C joints in Class 2 vessels built to past Class C requirements do not satisfy current radiography requirements.
4.4 PUMPS Pumps furnished under the requirements of the Hydraulic Institute Standards (14] were designed to satisfy functional requirements.
Integrity of the pressure boundary was not covered by this standard. The design of the pu=p pressure boundary should be evaluated in accordance with the current requirements of NB/NC/ND-3400 (1).
See Sections 4.1.5 and 4.2 of this Appendix for discussion of pump weld-ing requirements.
4.5 VALVES Class 1 valves current design requirements are given in Subarticle NB-3500 of Reference Ib.
All Class 1 valve materials must meet the fracture toughness requirements of NB-2332.
All Class 1 listed pressure rated valves should have a minimum body wall thickness as determined by ANSI B16.34 [13],
A-84 gg J...:mFranklin Research Center w s wmmann
t-4 e
2)
Ya t/
g3 P-No.
Current Code Requirements Past Code Requirements a5 Material Code Class ASME B&PV Sect. III (1965)
ASME VIII (1965)(6) h Cla ssi t icat ion 1
2 3
Class A Class C(l) a fd-II I*
I' 1
O 3/16 in 1 1/4 in O
O 1 1/4 in ln
=
3 0
3/16 3/4 0
0 3/4 2
m 4
0 3/16 5/8 0
0 5/8 5
0 3/16 0
0 0
0 7
0 3/16 5/8 0
0 See Note 3 8
0 3/16 1 1/2 0
0 See Note 3 9
0 3/16 See Note 3 0 0
5/8 m
10 0
3/16 5/8 0
0 0
11 0
3/16 5/8 0
0 See Note 3 1.
ASME B&PV Code Section III, 1965 Edition, Class C may currently be classified as Class 2 or 3 ot the current code.
2.
All thicknesses require full radiography when "0" is indicated.
3.
Requirements not specified for this P-No.
4.. These requirements are f or f ull penetration welded joints of Categories A, B, or C (N-463
[3)).
5.
These requirements are for full penetration welded joints of Categories A or B (N-2113
[3]).
Butt-welded joints of other categories shall satisfy the requirements of ASME B&PV Code Section VIII, 1965 edition.
6.
Vessels containin3 lethat substances shall have welded joints for materials of all thicknesses f ully radiographed.
4
1 I
except that the inside diameter, d, will be the larger of the basic valve body inside diameters in the region near the welding ends. Class 1 valves may be designed in accordance with either the standard design rules of NB-3530 through NS-3550 or the alternative design rules of NB-3512.2.
Alternative design rules require either computer analysis or experimental stress analysis procedures.
Listed pressure rated Class 1 valves should be hydrostatically tested to assure integrity of the pressure boundary (leakage thecugh the stem packing is not a cause for rejection) at not less than 1.5 times the 100*F rating rounded off to the next higher 25-psi increment as required by Reference 13, except that valves with a primary pressure rating of less than Class 150 will be subjected to the required test pressure for Class 150 rated valves.
Class 1 valves may be subjected to normal duty within the cyclic load 4
limits of NB-3550; otherwise the valve may have to be designed in accordance with the alternative design rules for severe duty applications.
Class 1 valves are to be designed for service levels A, B, C, and D with stress limits of NB-3525 through NB-3527 (1b]. Stress limits for level B loads are based on 110% of operating limits. Level C pressures are limited to 120% of operating limits. Pipe reaction stresses for level C loads are limited to 1.8 S, for the valve body material at 500*F, with S taken at 1.2 YS for the pipe at 500*F.
Primary and secondary stresses for level C loads 1.5, Q = 0, and limited to 2.25 S,.
Level D loads are based on C
=
7 may be evaluated in accordance with Appendix F (le).
l A design report for Class 1 valves will be prepared in accordance with the requirements of NB-3560 [lb].
Class 1 valves designed in accordance with the standard rules must satisfy tne body shape rules of NB-3544 which are intended to limit the local stre ss index to a maximum of 2.0.
Primary and secondary stre as intensities may then be calculated by the formulas given in NB-3545.1 and NB-3545.2 [lb],
re cpectively, and sub) ct to the stress limits described in Section 4.1.1 for Class 1 items.
Fatigue evaluation is performed by the rules and formulas of NB-3545.3.
A-86
-eT@s3
.;.J Franknn Research Center a > oaw N sme me
e Comparison With Past Requirements The past code (4) required that steel valves for power piping systems:
1.
be recommended for the intended service by the manufacturer 2.
be made from code materials suitable for the pressure and temperature 3.
have a minimum body metal thickness as required for ASA B16.5 fittings (15]
4.
shall be hydrostatically tested as required by Reference 15, i.e.,
1.5 times the 100*F rating rounded off to the next higher 25-psi increment, using water not above 125'F, with no leakage through the shell.
Note that the minimum body thickness of valves based on the current code would be based on ANSI B16.34 (13).
As an example, consider a 2500-lb valve designed in accordance with the past code (15).
Body thickness would be based on Table 33 (15]. Comparison with current requirements may be obtained from Table 3 (13] as shown in the following table:
Minimum Wall Thickness Based on Past and Current Codes 2500-lb Class Minimum Wall Thickness Nominal Pipe Inside Past Code Current Code Size (in)
Diameter (in)
Table 33 (15] Table 3 [13]
4 2.88 1.09 1.09 5
3.63 1.34 1.34 6
4.38 1.59 1.59 8
5.75 2.06 2.06 10 7.25 2.59 2.59
+
12 8.63 3.03 3.03 Notice that past valves would satisfy current thickness requirements.
It is concluded that Class 1 valves designed in accordance with past requirements would satisfy current requirement,s with the following possible l
exceptions:
i l
A-87 Og
...J Frankhn Research Center
=x.,oaww % nm.w.
i 1.
Fracture toughness requiremen5s may not be satisfied. Evaluate as recommended by Section 4.1 of this appendix.
2.
Valves may not sat.
i the primary, secondary and peak stress combination limits if body shape differs significantly from the rules of N3-3544 [lb).
3.
valves may not satisfy the primary plus secondary stress limit for service level C.
It is recommended that SEP Class 1 valves be evaluated on a case-by-case basis as follows:
1.
Use fracture toughness evaluation forms given in Section 4.1 of this appendix.
2.
Compare actual body shape with body shape rules of NB-3544 [lb].
If not significantly different, the valve would be considered adequate.
~
If significant differences are found, the Licensee should be asked to provide calculations and an evaluation based on alternative rules for the valve in question, unless it can be shown that the valve has been subjected to level C conditions and did not have to be replaced.
Class 2 and 3 valves are currently designed to the requirements of subarticle NC-3500 (lc] and ND-3500 [1d), re spectively. Class 2 valves satisfying the standard design rules comply with the standard class requirements of ANSI B16.34 except that valves with flanged and butt welded ends may be designated as Class 75 in sizes larger than 24-in nominal pipe size provided that NC-3512.l(a) is satisfied. Valves with flanged ends in sizes larger than 24-in nominal pipe size may be used provided that NC-3512. l (b) is satisfied. A shell hydrostatic test satisfying ANSI B16.34 is required. Class 2 and 3 valve stress limits for cervice limits A, B, C, and D are as given in Table A4-12.
Class 2 and 3 valves with butt welding or socket welding ends conforming to the requirements of NC-3661 and ND-3661 should satisfy the special class requirements of ANSI B16.34 except that:
a.
the nondestructive examination (NDE) requirements of ANSI B16.34, special class, shall be applied to all sizes in accordance with NC-2500 for Class 2 valves and ND-2500 for Class 3 valves.
b.
stress limits for service levels 3, C, and D shall be as shown in Table A4-12.
A-88 4
bV Frankhn Research Center 4 % or % %nea.
i Table A4-12 Level B, C, and D Service Limits for Class 2 and 3 Valves TABLE NC-35211 LEVEL B, C, AND D SERVICE LIMITS Service Umit Stress Umitsid P.S e,$ 1.1 S Level 8 (s. or e )+a,$ 1.655 1.1 t
e,$ 1.5 S Level C (e, or e,)+e,$ 1.s 5 1.2 a.$ Z.0 5 Level O (e. or e,)+e,$ 2.4 5 1.5 s
NOTES:
(1) A CJsting Qua4ty f acter of I shall te assurr'ed in satisfying these stress limits.
(2)These requirements for the acceptacility of vafve des.gn are not intended to assure the functional acecuacy of thevalve.
(3) Des.gn rew.re-ects hsted :n this tacle are net acclicab!e to vafve disks, stem, seat rings, er other parts cf the va!ves *nien are contamed withn tre confines of the body and toneet.
(s) T;1ese rules do not acosy to saf ety relief valves.
(5)The maximum cresssre snail 9et exceed the taculated factors tested under P times the Design Pressure er times the rated pressure at the acclicable semice tem;erature.
i I
A-a9
...a Franklin Research Center a ww w N r,., on -.
A openings for auxiliary connections shall satisfy ANSI B16.34 and the c.
reinforcetant requirements of NC-3300 and ND-3300.
Comoarison With Past Recuirements Class 2 and Class 3 valves designed by past code requirements would have the required minimum body thickness but may not comply with pressure-tempera-ture ratings of B16.34, which depend on material group and a rational formulation as compared to the empirical basis of B16.5.
It is recommended that the pressure-temperature rating of Class 2 and 3 SEP valves be compared with the current pressure-temperature rating of B16.34.
For example, the isolation valves of engineered safeguard system of the Palisades plant would be considered Quality Group B (Class 2) components by current standards. These valves are 150 lb rated valves designed to withstand 210 psig at 300*F by Table 2 of the past standard ASA B16.5 for flanged fittings. The current standard ANSI B16.34 gives an allowable pressure at 300'F which depends on the material group as shown in Table A4-13.
It is apparent f rom Table A4-13 that the engineered safeguard isolation valves for the Palisades plant would satisfy the current standard provided tha t the valve material was in one of the tabulated material groups other than 1.12, 2.1, or 2.3.
4.6 HEAT EXCHANGERS Heat excnangers are currently designed and constructed in accordance with i
the rules of ASME B&pV Code Section III,1977 Edition [1]. The design requirements for the pressure boundaries of the heat exchanger are found in i
the following sections of the current code:
Section Shell Side 3300 Tuce Side 3600 rube Sheet 3300 Shell Flange 3200 (Class 1); Appendix XI (Class 2 and 3)
Heat exchangers designed to ASME III (1965) or ASME VIII (1965) are compared as pressure vescels with current requirements in Section 4.3 of this Appendix.
D A-90
.... Franen Research Center s w ar ne rva nw.
i o
Table A4-13 Allowaole Working Pressure (l) for a 150 lb Standard Class Valve at 300*F Material Group Allowable Pressure (psig) 1.1 230 1.2 230 1.3 230 1.4 210 1.5 230 1.6 215 1.7 230 1.8 215 1.9 230 l
1.10 230 1.11 230 1.12 205 1.13 230 1.14 230 i
2.1 205 2.2 215 2.3 175 2.4 210 2.5 225 2.6 220 2.7 220 l
l 1.
Based on ANSI B16.34 (1977) l t
l l
A-91 j
s..J Franidin Research Center
- c~ a w N r,-. %,,
\\
o Heat exchangers designed to the standards of the Tubular Exchanger Manufacturers Association (TEMA) 1959 Edition [8] require that "the individual vessels shall comply with the ASME Code, for Unfired Pressure vessels." TEMA Class R heat exchangers are for the more severe requirements of petroleum and chemical processing applications. TEMA Class C heat exchangers are for the moderate requirements of commercial and general process applications.
The TEMA standards give design rules which " supplement and define the code for heat exchanger applications."- Allowable stress values, identical with Tables UCS-23 and DCS-27 of the 1959 edition of the ASME Code for Unfired Pressure Vessels, are reproduced in TEMA as Table D-8 for carbon and low alloy steels and as Table D-8W for carbon and low alloy pipe and tubes of welded manufacture; the stress values are one-fourth the specified minimum tensile strength multiplied by a quality f actor of 0.92.
Group I heat exchangers designed to TEMA (1959) would be governed by the code requirements of ASME VIII (1965). Comparison of ASME VIII (1965) with current requirements is as follows:
1.
Class 1 heat exchangers shell flanges would have to be designed by computer analysis to determine primary, secondary, and peak stress intensities, rather than design formulas as previously used.
2.
Materials for Class 1, 2, and 3 heat exchangers must comply with current fracture toughness requirements outlined in Section 4.1.1 of this Appendix.
3.
Radiography requirements for vessels designed and constructed to ASME III (1965) or ASME VIII (1965) are compared with current requirements in Section 4.3 of this Appendix.
t 4.7 STCRAGE TANKS Storage tanks may currently be classified as Class 2 or Class 3 and are designed in accordance with the rules of NC/ND-3900 [1] for atmospheric tanks or 0 to 15 psi tanks, respectively. Atmospheric tanks may be within building structures or above grade, exposed to atmospheric conditions.
Storage tanks of 0 to 15 psi design are normally located above ground within building structures.
l A-92
.. ] Franklin Research Center
% = m rwe,mu.
l Atmospheric Storage Tanks Atmospheric storage tanks are currently required to satisfy the general design requirements of NC/ND-3100 and the vessel design requirements of NC/ND-3300 except that a stress report is not required. Stress limits on the maximum normal stress for Service Levels A, B, C, D is as shown in Table A4-12.
Minimum size of fillet welds should satisfy NC/ND-4246.6, i.e., 3/16 in for 3/16-in thick plate, and at least 1/3 of thinner plate thickness for plates greater than 3/16 in but not less than 3/16 in.
Nominal thickness of shell plates should be at least 3/16 in for tanks of nominal diameter less than 50 f t or 1/4 in for tanks of 50 to 120 f t nominal diameter, but not greater than 1 1/2-in thick.
Roof s shall be designed to carry dead load plus a uniform load of at lea st 25 psf for outside tanks or at least 10 psf for inside tanks. Minimum roof plate thickness is 3/16 in plus corrosion allowance. Allowable ' stresses are summarized as follows:
a.
tension - for rolled steel, net section:
20 ksi; full penetration groove welds in thinner plate area:
18 ksi.
b, compression - 20 ksi where lateral deflection is prevented, or as determined f rom column formulas of NC/ND-3852.6 (b) (3).
c.
bending - 22 ksi in tension and compression for rolled shapes satisfying the shape requirement of NC/ND-3852.6 (c) (1) ; 20 ksi in tension and compression for unsymmetric members laterally supported at intervals no greater than 13 times the compression shape width; and for other rolled shapes, built-up members, and plate girders: 20 ksi in tension and compression as determined by the buckling formulas of NC/ND-3852.6 (c) (4).
d.
snearing - 13.6 ksi in fillet, plug, slot, and partial penetration groove welds across throat area, 13 ksi on the gross area of beam webs where the aspect ratio (h/t) is less than 60 or:
19.5 (h/t)-
7200 A-93 g
i.4 Franklin Research Center a>
oae N % m.v.
4 O to 15 osi Storage Tanks i
Storage tanks which may contain gases or liquids with vapor pressure above the liquid not exceeding 15 psig are currently designed in accordance with the requirements of NC/ND-3920. Maximum tensile stress in the outside tank walls is as given in Table I-7.0 of Reference le if both meridional and latitudinal forces are in tension, or this value multiplied by the tensile stress factor N (less than 1.0) determined from the Biaxial Stress Chart, Fig.
NC/ND-39222.1-1 (1) if one of these forces is compressive. Maximum compressive stress in the outside wall shall be determined by the rules of NC/ND 3922.3 (1]. Maximum allowable stress values for structural members shall be as determined f rom NC/ND-3923. The O to 15 psi storage tank shall be designed in accordance with the detailed rules of NC/ND-3930.
Comparison with Past Code Recuirements Storage tanks in Group I SEP plants were designed either in accordance with A/E specifications, USAS 396.1 (1967) (9), API-650 (1964) [10), ASME III (1965) Class C, or ASME VIII (1965). Stress allowables for ASME III (1965)
~
Class C vessels are as given in ASME VIII (1965).
Examination of the ASME VIII (1965) allowable stress valves for carbon and low alloy plate steels indicates that the values do not exceed 20 ksi except for SA-353 Grade A and 3, with allowable stresses of 22.5 and 23.75 ksi, respectively. ASME VIII (1965) does not consider biaxial stress fields with associated reduction in tensile allowables. Stress allowables for roofs in Reference 10 are the same as for current atmospheric storage tanks.
A comparison of API-650 (1964) roof design requirements, including stress allowables, shows agreement with current requirements; shell material and tensile stress allowables may, however, not satisfy current requirements. The past code allows the use of A-7 plate material not currently listed as an acceptable material.
The past code permits an allowable tensile shell stress 21,000 psi times the joint ef ficiency. Assuming spot radiography of a double welded butt vertical shell joint made from A-283 Grade C or A-36 plate material, the allowable stress would be 17,850 psi based on 0.85 joint efficiency, which exceeds the current 12,600 psi allewable.
A-94 b
LLW Franklin Research Center
% e n. %%,
o e
USAS B96.1 (1967) for welded aluminum alloy field-erected storage tanks cannot be used for Class 2 storage tanks since aluminum alloy is not a permitted Class 2 material as listed in Table I-7.0 [1].,However, allumimum alloy can be used for Class 3 storage tanks since aluminum alloys are listed in Table I-8.4, which is currently used for aluminum shell design, and in Tables ND-3852.7-2 through ND-3 852.7-6 for aluminum roof design. A comparison of allowables based on past and current codes is shown in the following table:
Aluminum Specified Min.
Allowable Stress Structures Material Strength Past Current (Type of Stress)
(Temper)
TS/YS (USAS B96.1) (ASME III (1977))
Shell (Tension) 5050 (0) 18.0 ksi/6.0 ksi 4.8 kai 4.0 ksi Shell (Tension) 6061 (T4,T6) 24.0 ksi/ -
7.2 ksi 6.0 kai 18.0 ksi 18.0 ksi Bolts (Tension) 6061 (T6)
Roof Support 19.0 19.0 (Axial Compres-6061 (T6) sion, L/r < 10)
Roof Support (Axial Compres-6061 (T6) 20.4-20.4-ston 10 < L/r < 67) 0.135 L/r 0.113 L/r From this table, it can be concluded that:
1.
shells designed to USAS B96.1 (1967) may be overstressed by as much as 20% compared to current allowables 2.
Dolts designed to USAS 395.1 (1967) satisfy current requirements 3.
roof supports with slenderness ratios up to 10 satisfy current requirements 4.
roof supports with slenderness ratios between 10 and 67 more than satisfy current ecmpression allowables by as much as 13%.
Therefore, aluminum alloy storage tanks built to USAS B96.1 (1967), when evaluated against current requirements:
1.
At temperatures to 100*F.
A-95
a 1.
may not satisfy materials requirements in Table I-7.0 if the tank is a Class 2 component 2.
may be unconservatively designed when compared to current stresc allowables, by as much as 20% for the shell.
In conclusion, 1.
Tanks designed to A/E specification should be carefully compared to current code requirements 2.
Atmospheric tanks designed to ASME III (1965) Class C are likely to satisfy current requirements with regard to allowable tensile stress, but may not satisfy current compression stress requirements. Class C atmospheric tanks currently classified as Code Class 2 may not satisfy the current quality assurance requirements as discussed in Section 4.1.2 of this Appendix.
3.
O to 15 psig tanks designed to ASM3 III (1965) Class C requirements may not satisfy current tensile allowables for biaxial stress fields in which one of the stress components is compression. These tanks should be examined carefully in light of current requirements. Class C (0 to 15 psig) tanks currently classified as Code Class 2 may not satisfy the current quality assurance requirements discussed in Section 4.1.2 of this Appendix.
4.
Atmospheric storage tank roofs designed to API-650 (1964) satisfy current stress allowables.
i 5.
Atmospheric welded steel storage tanks designed to API-650 (1964) may not satisfy current requirements with regard to a.
use of A-7 plate material not currently acceptable b.
shell tensile stresses may exceed current code allowables 6.
Atmospheric storage tanks designed to USAS 396.1 (1967) may not satisfy current requirements.
A-96
e 5.
BASIS FOR SELECTING REQUIREMENTS MOST SIGNIFICANT 70 COMPONENT INTEGRITY The selection of code requirements most significant to component integrity has been based on the experience of the aurhor and colleagues in industry, govsenment, and academia. Codes pertaining to the design and construction of nuclear power plants have been modified and expanded. The changes reflect new "ctate of the art" knowledge, new techniques of f abrication, examination, totting, and methods of achieving quality that have been " filtered" and accspted by the technical community.
It is the author's view that current codas represent a consensus of what is best for achieving both economy of construction and public safety.
Accordingly, changes in stress limits, full radiography requirements, and fatigue evaluation for piping, cs well as more conservative requirements for fracture toughness, have been given special attsntion.
A-97 p-
...J Frankhn Research Center a w on n r.aea u.
4 6.
REFERENCES
- 1. ASME Boiler and Pressure Vessel Code,Section III, " Nuclear Power Plant Components" American Society of Mechanical Engineers,1977 Edition with Addenda through Summer 1978 a.
Division 1 and Division 2 General Requirements b.
Division 1, Subsection NB, Class 1 Components c.
Division 1, Subsection NC, Class 2 Components d.
Division 1, Subsection ND, Class 3 Components e.
Division 1, Appendices
- 2. Title 10 of the Code of Federal Regulations Revised January 1, 1981
- 3. ASME Boiler and Pressure Vessel Code,Section III, " Rules for Construction of Nuclear Vessels" American Society of Mechanical Engineers,1965
~
- 4. Code for Pressure Piping American Society of Mechanical Engineers, 1955 ASA 331.1-1955
- 5. ASME Boiler and Pressure Vessel Code,Section VIII, "Unfired Pressure vessels," and Section I, " Power Bollers" American Society of Mechanical Engineers,1965
- 6. Quality Group Classifications and Standards for Water, Steam, and Radicactive-Waste-Containing Components of Nuclear Power Plants, Rev. 3 NRC, February 1976 Regulatcry Guide 1.26
- 7. Standard Review Plan, Section 3.2.2,
" System Quality Group Classification" NRC, Of fice of Nuclear Reactor Regulation, July 1981 NUREG-0800 S. Standards of Tabular Exchanger Manuf acturers Association, Fourth Edition, 1959 9.
" USA Standard Specification for Welded Aluminum-Alloy Field-Erected Storage Tanks" United States of America Standards Institute, February 1967 USAS 396.1-1967
- 10. " Welded Steel Tanks for Oil Storage," Second Edition American Petroleum Institute, April 1964 API-650 4
A-98 3.LJ Franklin Research Center
% sNre %.
- 11. Final Safety Analysis P.eport for Palisades Plant (3 Volumes)
?
Consumers Power Company, November 5, 1968 USAEC Docket No. 50-255
- 12. Brock, J. E.
"A Temperature Chart and Formulas Useful with USAS B31.7 Code for Thermal Stress in Nuclear Power Piping" Nuclear Engineering and Design, Vol.10, pp. 79-82 Austria: North Holland Publishing Co., 1969
- 13. " Steel Valves" American Society of Mechanical Engineers, 1977 ANSI B16.34-1977
- 14. Hydraulic Institute Standards, Eleventh Edition, 1965
- 15. " Steel Pipe Flanges and Flanged Fittings" American Society of Mechanical Engineers, 1961 ASA B16.5-1961 O
e A-99 4
JU) Franklin Research Center a w as w-m.
_,