ML20054F725
| ML20054F725 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 05/21/1982 |
| From: | Berkowitz L, Gonzalez A, Tikoo S FRANKLIN INSTITUTE |
| To: | Boyle M NRC |
| Shared Package | |
| ML20054F721 | List: |
| References | |
| CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-437, NUDOCS 8206170252 | |
| Download: ML20054F725 (160) | |
Text
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TECHNICAL EVALUATION REPORT QUALITY GROUP CLASSIFICATION OF COMPONENTS AND SYSTEMS (SEP, III-1)
DAIRYLAND POWER COOPERATIVE LACROSSE BOILING WATER REACTOR (LACBWR)
N RC DOCKET NO.
50,409 FRC PROJECT C5257 i
NRC TAC NO.
41600 FRC ASSIGNMENT 17 NRC CONTRACT NO. NRC 03 79-118 FRC TASK - 437 Author:
on alez Franklin Research Centor L. Berkowitz 20th and Race Street FRC Groi>o Leader:
A. Gonzalez Philadelphia, PA 19103 Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:
M. Boyle May 21, 1982 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees. makes any warranty, ex-pressed or implied, or assumes any legal liability or responsibility for any
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third party's use, or the results of such use, of any information, apparatus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
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Franklin Research Center A Division of The Franklin Institute The Benprran Franklin Parhey. PNia. Pa. 19103 (215)448 1000 8206170252 820607 PDR ADOCK 05000409 p
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TER-C5257-437 CDNTENTS Section Title Page 1
INTRODUCTION.
1 i
2 SCOPE OF THE EVAWATION.
2 3
METHOD OF REVIEW.
5 4
QUALITY CLASSIFICATION OF SYSTEMS AND COMPONENTS.
6 5
EVAWATION OF SPECIFIC COMPONENTS 25 5.1 General Requirements 25 5.2 Pressure Vessels 38 5.3 Piping.
40 5.4 Pumps 42 5.5 valves.
44 5.6 Storage Tanks 46 l
6 CONCLUSIONS AND RECOMMENDATIONS.
47 7
REFERENCES 50 l
APPENDIX A - REVID4 OF CODES AND STANDARDS i
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6 Udbd Franklin Research Center A w w The r- %
5 TER-C5 25 7-437 I
FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Connaission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.
Mr. L. Berkowitz contributed to the technical preparation of this report through a subcontract with Innovation Technology, Inc.
O ddU Franklin Research Center s w at m nmen m.aue
i TER-CS 25 7-437 1.
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 Lacrosse Boiling Water Reactor.
This report was prepared by the Franklin Research Center (FRC) under NRC l
Contract No. NRC-0 3-79-118.
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l nklin Research Center A Onamon af The Frenahn Insende
TER-C5257 437 2.
SCOPE OF THE EVALUATION The SEP concerns a review and assessment of the safety of older nuclear
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plants on the basis of current licensing criteria. Topic III-l la 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 Onofre 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, Topic III-1 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].
(Some codes in effect from 1965 to 1973 were also reviewed because they were used when retrofitting components.)
The objective of the present evaluation is to assess the ability of systems and components in the Lacrosse Boiling Water Reactor 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, installation, and testing of tne plant's systems and components to identify significant differences that might affect structural integrity, and (2) assessment of the effect of these differences on tne safety margins of the systems and components.
1.
Plant discussed in this report. ;
Ed Franklin Research Center Acm onan r,.n.em u.
l
TER-C5257-437 The scope of this evaluation is limited by or to the following:
1.
Table of Systems and Components (including updates and revisions) [4),
compiled by the NRC, corrected an( completed by Dairyland Power Cooperative. This table contains the quality group classification, the current code, and the code used for thy listed systems and components when the plant was designed. When the information in the table was incomple *,e, it was completed as well as possible (see Table 4-1).
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 Pressere Piping" [7] and applicable codes cases; USAS B31.7 (1968), "Draf t Code for Pressure Piping, l
Nuclear Power Piping" [8); ASA B16.5 (1961), " Steel Pipe Flanges and Flanged Fittings" [9); ASME 1962 Boiler and Pressure Vessel Code,Section VIII, "Unfired Pressure Vessels" [10) and applicaole code I
cases; ASA B16.9 (1958), " Factory Made Wrought Steel Butt Welding Fittings" [11); ASA B16.10 (1957), " Face-To-Face and End-to-End Dimensions of Ferrous Valves (12).
6.
Current code:
1977 ASME Boiler and Pressure Vessel (B&PV) Code, Sectio.a III, Division 1, to include the General Requirements (articles with "NA" subscript), Subsection NB, NC, and ND, and Appendices, supplemented through tne 1978 Summer Addenda [2].
7.
Quality Group D components are not considered in this evaluation.
8.
Altnough 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 otner SEP topics:
Topic Description III-5. A Effects of Pipe Break on Structures, Systems and Components Inside Containment l
III-5.B Pipe Break Outside Containment III-6 Seismic Design Consideration l
1.
Letter from S. Bajwa to S. Carfagno dated December 10, 1981. 4%
d'MnkHn Research Center a cm or w n.non %.
TER-C5 25 7-437 Topic Description III-7.A Inservice Inspection, Including Prestressed Concrete Containments with Either Grouted or Ungrouted Tendons III-7.B Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria III-7.D Containment Structural Integrity Tests III-9 Support Integrity V-3 overpressurization Protection V-6 Reactor Vessel Integrity V-8 Steam Generator Integrity IX-6 Fire Protection
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TER-C5257-437 3.
METHOD OF REVI1!N I
To accomplish the objective of this evaluation, a review was performed 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. 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 l
outside the scope of this review (designated as "O").
For each significant requirement in the current code, a similar requirement was sought in the older codes. The major older codes for the Lacrosse Boiling Water Reactor are ASA B31.1 (1955) [7], USAS B31.7 (1968) (8], ASA B16.5 (1961) (9], ASA B16.9 (1958) (11], ASA B16.10 (1957) [12], and the 1962 ASME B&PV Code,Section VIII (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.
l Knowledge of the historical development of the codes and the reasons for the changes was an important element in making effective comparisons. A l
literature survey, supported by consultation with experts in the field, helped l
to identify certain changes for special attention, e.g.,
changes in design l
l criteria, analytical methods, load combinations, quality assurance require-ments, fabrication techniques, and testing requirements. b Franklin Research Center a om.an or n. rr a m
TER-C5257-437 4.
QUALITY CLASSIFICATION OF SYSTDtS AND COMPONENTS 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 Lacrosse Boiling Water Reactor, 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.
The following systems are listed in Table 4-1 with their respective components:
Reactor Coolant System Recirculation System Emergency Systems Boron Injection System High Pressure Core Spray System Alternate Core Spray System Building Spray System Manual Depressurization System Safety Relief Valves Reactor Coolant Pressure Boundary Isolation Valves Containment Penetrations Valves and Piping Control Rod Drive Housing Control Rod Drive System Storage Well Cooling System Condensate /Feedwater System Main Steam System Primary Purification System Decay Heat Cooling System Component Cooling Water System Shutdown Condenser System Water Purification System Seal Injection System High Pressure Service Water System Low Pressure Service Water System Structures (for information only, not in the scope of this review).
A dbranklin Research Center acem.ona m r e u.
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TER-C5 25 7-437 Table 4-2(a) lists all Quality Group A components, Table 4-2(b) lists all Quality Group B components, and Table 4-2(c) lists all Quality Group C components. Components in Table 4-2 are grouped as pressure vessels, piping, pumps, valves, and storage tanks. The major code used when the component was built is also provided. Table 4-2(d) provides an index of the abbreviations used for the systems and their definitions.
Additional information on the review procedure for System Quality Group Classification can be obtained from Section 3.2.2 of the Standard Review Plan (6].
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Table 4-1 fg classification of Structures, Systems, and Components 3E Lacrosse Boiling Water Reactor (LACBWR) am FF Th Quality Classification R
Codes anj Codes and Seismic Classification 5 [)
Structures, Systems, Standards Standards Usea Used in g
and Components RG 1.26 (1) in Plant Design RG 1.29 Plant Design (2)
Remarks R
kEAC10R COOLANT SYSTD4 Reactor Vessel ASME III
'ASME VIII (1962)
Category I NAI33' Class 1
^
Reactor Vessel Supports Category I NA
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e Heactor Vessel Internals ASME'III Unknown Category I NA 7
Class 1 Hany ASTM Specifications
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HECIHCIRATION SYSTD4 Piping ASME III ASA B31.1 (1955)
Category I See Notes 4 Class 1 ASA B16.9 (1958)*
and 5 Allis-Chalmers Spec. No.41-585 s Addenda 1
1 1.
ASME III stands for the Boiler and Pressure Vessel Code,Section III, Division I, published by the American-Society of Mechanical Engineers,1977, Edition with Addenda through the summer 1978 Addenda.
2.
Plant was designed to Uniform Building Code,1958 Edition per Safety Analysis Report.
>3 3.
NA indicates that information given on that line of the table is outside the scope of this report.
Q 4.
Plant design meets the intent of USAS B31.7 (1968) according to Reference 4.
[
5.
ASA B16,9 (1958) and ASA B16.10 (1957) were listed in the Licensee's original submittal, but not on the M
referenced specifications. Clarification on this discrepancy should be provided.
DE The edition of the code is an assumption, because this information is not known at this time.
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table 4-1 (Cont.)
1 Es p:0 Quality Classification
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Codes and Codes aM Seismic Classification f g Structures, Systems, Standards Standards Used Used in g
and Caponentr.
hG 1.26 (1) in Plant Design RG 1.29 Plant Design (2)
Remarks G
Valv9s ASME III ASA B16.10 (1957)*
Category I See Ref. 4 Class 1 ASA 816.5 (1961)*
Allis-Chalmers 2
Spec. No.41-576 Pumps ASME III Allis-Chalmers Category I See Ref. 4 Class 1 Spec. No.41-569 ew EMEHGENCY SYSTD4 Boron Injection System Sodium Pentaborate Tank ASME III AWWA D-100 (1959)*
Category I See Ref. 4 Class 2 Allis-Chalmiers Spec. E).41-570 Chemical Mixing Tank ASME III AWWA D-100 (1959)*
Category I See Ref. 4 Class 2 Allis-Chalmers Spec. No.41-570 Pipiry and Valves ASME III ASA B31.1 (1955)
Category 1 See Ref. 4 Class 2 ASA 816.9 (1958)*
High Pressure Core a
Sprev-System
[
Pumps ASME III Allis-Chalmers Category 1 See Ref. 4 Class 2 Spec. No.41-572 M
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Table 4-1 (Cont.)
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> 71 Ouality classification a 5-Codes and Codes and Seismic Classification g
Structures, Systems, Standards Standards Used Used in and Ccuaponents RG 1.26 (1) in Plant Design RG 1.29 Plant Design (2)
Remarks
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5h overhead Storage ASME III Allis-Chalmers Category I Calculations A
Tank Class 2 Spec. No.41-550 used in
.k ASME VIII (1962) tank design 14 Code case 1272N-5 provided in Ref. 4 Piping and Valves ASME III ASA B31.1 (1955)
Category I Class 2 ASA B16.9 (1958)
- ASA B16.10 (1957)*
ASA B16.5 (1961)
- b Core Spray Header ASHE III ASA B31.1 (1955)
Category I y
Class 2 ASA B16.9 (1958)*
ASA B16.10 (1957)*
t ASA B16.5 (1961)
- Alternate Core Spray System Diesel-Driven Pumps ASME III Sargent & Lundy Category I See Ref. 4 Class 2 Spec. No. W-1924 Piping and Valves ASME III ASA B31.1 (1955)
Category I i
Class 2 ASA B16.9 (1958)*
ASA B16.10 (1957)*
)
ASA B16.5 (1961)*
Building Spray System ti Piping and Valves ASME III ASA B31.1 (1955)
Category I h
Class 2 ASA B16.9 (1958)*
ASA B16.10 (1957)*
ASA B16.5 (1961)*
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Table 4-1 (Cont.)
EC Ouality Classification
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Codes and Codes and Seismic Classification Structures, Systems, Standards Standards Used Used in h
and Components RG 1.26 (1) in Plant Design RG 't.29 Plant Design (2)
Remar k s E
Manual Depressurization System (MDS) ls r[r Piping and Valves ASME III ASA B31.1 (1955)
Category I I
g)
Class 1 ASA B16.9 (1958)*
ASA B16.10 (1957)*
ASA B16.5 (1961)*
SAFETY RELIEF VALVES ASME III ASME VIII (1962)(6) Category 2 Class 1 REAC10R CLY3LANT PRESSUkE BOUT 3DARY b
y' Piping from Reactor Vessel ASME III ASA B31.1 (1955)
Category I up to and Including First Class 1 ASA B16.9 (1958)*
Isolation Valve External ASA B16.10 (1957)*
to Containment ASA B16.5 (1961)
- ISOLATION VALVES Valves Other Than ASME III ASA B31.1 (1955)
Category I See Ref. 4 Those Identified Under Class 1 ASA B16.9 (1958)*
Containment Penetrations ASA B16.10 (1957)*
Valves and Piping ASA B16.5 (1961)
- Allis-Chalmers Spec. No.41-576 6.
Safety relief valves are mentioned in Section VIII in reference to functional requirements, not design requirements. It is more likely that ASA B31.1 and code cases would have been used for design of safety relief g
valves.
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Table 4-1 (Cont.)
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g Quality Classification
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Codes and Codes and Seismic Classification g :i Structures, Systems, Standards Standards Used Used in gp and Components RG 1.26 (1) in Plant Desiga RG 1.29 Plant Design (2)
Remarks f'4 0)NTAINHDIT PENETRATIONS ASME III AS A B 31.1 (1955)
Category I 2 Sr VALVES AND PIPING Class 2 ASA B16.9 (1958)*
fp AS A B16.10 (1957)
- 3 ASA B16.5 (1961)*
4 CONTHOL ROD DRIVE ASME III ASME VIII (1962)
Category I HO USING Class 1 00tiTHOL ROD DRIVE ASME III ASME VIII (1962)
Category I See Ref. 4 SYSTUM Class 2 I
[
S1DRAGE WEI.L (DOLING 8
SYSTEM Pumps ASME III Allis-Chalmers Category I See Ref. 4 Class 3 Spec. No.41-563 Heat Exchanger ASME III ASME VIII (1962)
Category I Class 3 Code Case 1270N Piping and Valves ASMC III ASA B31.1 (1955)
Category I
)
ClasJ 3 CONDENSATE /PEEDWATDt s
SYSTHH Piping from Outermost ASME III ASA B31.1 (1955)
Category I See Note 4 Containment Isolation Class 2 Valve up to and Includ-H ing the Shutoff Valve and Connected Piping up 4
to and Including the First Shutoff Valve hj 4
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Table 4-1 (Cont.)
c.
- Tl Quality Classification Codes and Codes and Seismic Classification p;n Structures, Systems, Standards Standards Used Used in and Components HG 1.26 (1) in Plant Design HG 1.29 Plant Design (2)
Remarks h
MAIN STEAM SYSTDt
[n Piping from Outermost ASME III ASA B31.1 (1955)
Category I g&
Containment Isolation class 2 Allis-Chalmers Valve up to Turbine Stop Spec. No.41-565 and Bypass Valves and Connected Piping up to and Including First Valve PRIMARY PURIFICATION ASME III ASA B31.1 (1955)
Non-seismic SYSTEM Class 3 Category I IW W
DhrAY llEAT COOLING SYSTEM lleat Exchanger Tube Side ASME III ASME VIII (1962)
Category I Class 2 Code Case 1270N Shell Side ASME III ASME VIII (1962)
Category I Class 3 Code Case 1270N Ct44EONENT COOLING WATER SYSTD4 Pumps ASME III Allis-Chalmers See Ref. 4 Class 3 Spec. No.41-563 Tank ASME III ASME VIII (1962)
See Ref. 4 8
Class 3 Allis-Chalmers N
Spec. No.41-570 N
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<=-y Table 4-1 (Cont.)
Quality Classification 13 Codes and Codes and Seismic Classification 2,9 Structures, Systems, Standards Standards Used Used in and Components RG 1.26 (1) in Plant Design RG 1.29 Plant Design (2[
Remarks 37 5
Piping and Valves ASME III ASA B31.1 (1955)
Category I Q
Class 3
?.
N Shield Cooling Pumps ASME III Allis-Chalmers See Ref. 4 Class 3 Spec. No.41-563 Shield Cooling Tank ASME III ASME VIII (1962)
See Ref. 4 Class 3 Allis-Chalmers Spec No.41-570 SIEJTDOWN ODNDENSER ASME III ASA B31.1 (1955)
Category I 8
SYSTEM Class 2 WATEM PURIFICATION SYSTEM Liquid Haste Demineraliser ASME III Allis-Chalmers Non-seismic See Ref. 4 Pump Class 3 Spec. No.41-563 Category I Spent Resin Tank ASME III ASME VIII (1962)
Non-seismic See Ref. 4 Class 3 Allis-Chalmers Category I Spec. No.41-570 SEAI. INJECTION SYSTEN H
Pumps ASME III Allis-Chalmers Non-seismic See Ref. 4 Class 3 Spec. No.41-572 Category I Ma w
9
E Table 4-1 (Cont.)
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Ouality Classification gg Codes and Codes and Seismic Classification
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g 5-Structures, Systems, Standards Standards Used g
and Components RG 1.26 (1) in Plant Design RG 1.29 Plant Design (2)
Remarks Used in
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y Surge Tank ASME III ASME VIII (1962)
Non-seismic 5S-See Ref. 4 Class 3 Allis-Chalmers Category I
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Spec. No.41-570 k
Reservoir Tank ASME III ASME VIII (1962)
Non-seismic See Ref. 4 Class 3 Allis-Chalmers Category I Spec. No.41-570 Piping and Valves ASME III ASA E31.1 (1955)
Non-seismic Class 3 Category I h
li!GI PRESSURE SERVICE g
WATER SYSTDt Auxiliary Pumps ASME III Sargent & Lundy Category I See Ref. 4 Class 3 Spec. No. N-19_
Pipirvg and Valves ASME III ASA B31.1 (1955)
Category I Class 3 IIM PRESSURE SERVICE LIME III ASA B31.1 (1955)
Category I WATER SYSTEM Class 3 STRth.W.RES Containment Building Category I NA H
Containment Shell and ASMS III ASME II, VIII, pategory I Penetrations NC and IX (1962)
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,3 TEble 4-1 (Cont.)
59 53
- r g' n Quality classification
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Codes and Codes and Seismic Classification k
Structures, Systems, Standards Standards Used Used in and Components RG 1.26 (1) in Plant Design RG 1.29 Plant Design (2)
Remarks Stack Non-seismic NA Crib House Category I NA (Water Intake Structure) 8 l
H Turbine Building Category I NA i
(Portions Housing Class 1 Equipment)
Control Room Category I MA Waste Treatment Non-seismic NA Dullding I
Spent Fuel Storage Well Category I NA New Fuel Storage Racks Category I NA H
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TER-C5 25 7-437 Table 4-2(a)
Quality Group A Components (1) 7 Code: ASME III-Class 1(2)
Pressure Vessels Code Control Rod Drive Housing (CRDE)
ASME VIII (1962)
Piping Recirculation System Piping (BCS)
ASA B31.1 (1955)
ASA B16.9 (195 8)
- Allis-Chalmers Spec. No.41-585 (3)
Manual Depressurization System ASA B31.1 (1955)
Piping (MDS)
ASA B16.9 (1958)*
ASA B16.5 (1961)*
ASA B16.10 (195 7)
- Piping from Reactor Vessel ASA B31.1 (1955) up to First Isolation ASA B16.10 (1957)*
Valve External to Containment ( RCPB)
ASA B16.9 (195 8)
- ASA B16.5 (1961)
- Pumps Recirculation System Pumps (RCS)
Allis-Chalmers Spec. No.41-569 Valves Recirculation System Valves (BCS)
ASA B16.10 (1957)*
ASA B16.5 (1961)*
Allis-Chalmers Spec. No.41-576 1.
Refer to Table 4-2(d) for abbreviations.
2.
ASME III-Class 1 stands for American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NB, 1977 Edition and Addenda through the Summer 1978 Addenda.
3.
Plant design meets t he intent of USAS B31.7 (1968) according to Reference 4.
- The edition of the code is an assumption because this information is not known at this time.
p '!0bJ Franklin Resear.ch Center a w as w rr n..m
I 4-l TER-C5257-437 Table 4-2 (a) (Cont. )
)
Valves (Cont.)
Code Manual Depressurizatica ASA B31.1 (1955)
System Valves (MDS)
ASA B16.9 (1958)
- ASA B16.10 (1957)*
ASA B16.5 (1961)*
ASME VIII (1962) (4)
Valves Encountered from Reactor Vessel ASA B31.1 (1955) up to and Including First Isolation ASA B16.9 (1958)*
Valve External to Containment (RCPB)
ASA B16.10 (1957)
- ASA B16.5 (1961)*
Isolation Valves Other than Those ASA B31.1 (1955)
Icentified Under Containment ASA B16.9 (1958)
- Penetrations' Valves and Piping (IV)
ASA B16.10 (1957)*
ASA B16.5 (1961)
- Allis-Chalmers Spec. No.41-576 Storage Tanks ( Atmospheric and 0-15 psig)
None 4.
Safety relief valves are mentioned in Section VIII in reference to f
functional requirements, not design requirement.
It is more likely that ASA B31.1 and code cases would have been used for design of the safety relief valves.
nklin Research Center A Dween of The Feenda innaeute
~
TER-C5 25 7-437 Table 4-2(b)
Quality Group B Components (1)
Code: ASME III-Class 2 (2)
Pressure Vessel
~
Code Sodium Pentaborate Tank (BIS)
ANNA D-100 (1959)
- Allis-Chalmers Spec. No.41-570 Overhead Storage Tank (HPCSS)
ASME VIII (1962)
Code Case 1272N-5 Allis-Chalmers Spec. No.41-550 Decay Heat Cooling System ASME VIII (1962)
Heat Exchanger - Tube Side (DHG)
Code Case 1270N Piping Boron Injection Sy.3 tem Piping (BIS)
ASA B31.1 (1955)
ASA B16.9 (195 8)
- High Pressure Core Spray System Piping (HPCSS)
ASA B31.1 (1955)
ASA B16.10 (1957)*
l ASA B16.9 (195 8)
- ASA B16.5 (1961)*
l Core Spray Header (HPCSS)
ASA B31.1 (1955)
ASA B16.9 (195 8)
- ASA B16.5 (1961)*
ASA B16.10 (195 7)
- Alternate Core Spray System Piping (ACSS)
ASA B31.1 (1955)
ASA B16.9 (195 8)
- ASA B16.10 (195 7)
- ASA B16.5 (1961)*
l 1.
Refer to Table 4.2(d) for abbreviations.
2.
ASME III-Class 2 stands for American Society of Mechanical Engineers, Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NC, 1977 Edition and Addenda through the Summer 1978 Addenda.
As UOhbranklin Research Center l
A % w N rr.noa me
TER-C5257-437 Table 4-2 (b) (Cont. )
Piping (Cont. )
Code Building Sprav System Piping (BSS)
ASA B31.1 (1955)
ASA B16.9 (1958)
- ASA B16.10 (1957)*
ASA B16.5 '1961)
- Containment Penetration Piping (CP)
ASA B31.1 (1955)
ASA B16.9 (1958)
- ASA B16.10 (1957)*
ASA B16.5 (1961)*
Control Rod Drive System (CRDS)
ASME VIII (1962)
Piping from Outermost Containment ASA B31.1 (1955)
Isolation Valve up to the Shutoff Valve and Connected Piping up to First Shutoff Valve (C/EW)
Main Steam System Piping from ASA B31.1 (1955)
Outermost Containment Isolation Valve up Allis-Chalmers to Turbine Stop and Bypass Valves and Spec. No.41-585 Connected Piping up to First Valve (MSS)
Shutdown Concenser System Piping (SCS )
ASA B31.1 (1955)
Pumps High Pressure Core Spray System Pumps (HPCSS)
Allis-Chalmers Spec. No.41-572 Alternate Core Spray System Sargent & Lundy Diesel-Driven Pumps (ACSS)
Spec. No. W-1924 Valves Boron Injection System Valves (BIS)
ASA B31.1 (1955)
ASA B16.9 (1958)*
High Pressure Core Spray System Valves (HPCSS)
ASA B31.1 (1955)
ASA B16.9 (1958)
- ASA B16.10 (1957)*
ASA B16.5 (1961)* dlhranklin Research Center
~ ~ ~.
=
TER-C5 25 7-437 Table 4-2(b) (Cont.)
Valves (Cont.)
Code Alternate Core Spray System Valves (ACSS)
ASA B31.1 (1955)
ASA B16.9 (195 8)
- ASA B16.10 (195 7)
- ASA B16.5 (1961)*
l Building Spray System Valves (BSS)
ASA B31.1 (1955)
ASA B16.9 (1958)
- ASA B16.10 (195 7)
- ASA B16.5 (1961)*
Containment Penetration Valves (CP)
ASA B31.1 (1955)
ASA B16.9 (195 8)
- ASA B16.10 (195 7)
- ASA B16.5 (1961)*
Condensate /Feedwater System ASA B31.1 (1955)
Valves from Outermost Containment Isolation Valve up to and Including all Shutoff Valves (C/Df)
Valves from Outermost Containment ASA B31.1 (1955)
(
Isolation Valve up to Turbine Stop and Bypass Valves and Including First Valve (MSS)
Shutdown Condenser System Valves (SCS)
ASA B31.1 (1955)
Storage Tanks ( Atmospheric and 0-15 psig)
Chemical Mixing Tank (BIS)
AWWA D-10 0 (195 9)
- Allis-Chalmers Spec. No.41-570 l
pm N ranklin Research Centes
% an.r- -
TER-C5 25 7-437 Table 4-2 (c)
Quality Group C Components (1)
Cod e r ASME III-Class 3 (2)
I Pressure Vessels Code Storage Well Cooling System ASME VIII (1962)
Heat Exchanger (SWCS)
Code Case 1270N Decay Heat Cooling System ASME VIII (1962)
Heat Exchanger - Shell Side (DHG)
Code Case 1270d Component Cooling Water System Tank (CCNS)
ASME VIII (1962)
Allis-Chalmers Spec. No.41-570 Shield Cooling Tank (CCWS)
ASME VIII (1962)
Allis-Chalmers Spec. No.41-570 Spent Resin Tank (WPS)
ASME VIII (1962)
Allis-Chalmers Spec No.41-570 Seal Injection Surge Tank (SIS)
ASME VIII (1962)
Allis-Chalmers Spec. No. 41-5 70 Seal Injection Reservoir Tank (SIS)
ASME VIII (1962)
Allis-Chalmers Spec. No.41-570 Piping Piping Associated with ASA B31.1 (1955)
Storage Well Cooling System (SWCS)
Primary Purification System Piping (PPS)
ASA B31.1 (1955)
Component Cooling Water System Piping (CCWS)
ASA B31.1 (1955)
Seal Injection System Piping (SIS)
ASA B31.1 (1955) 1.
See Table 4.2(d) for abbreviations.
2.
ASME III-Class 3 stands for American Society of Mechnical Engineers, Boiler and Prescure Vessel Code,Section III, Division 1, Subsection ND, 1977 Edition and Addenda through the Summer 1978 Addenda.
A !dU Franklin Research Center a omen a m r,.anan m ui.
_- _. ~.. -.
TER-CS 25 7-437 Table 4.2 (c) (Cont.)
T Piping (Cont. )
Code High Pressure Service Water System ASA B31.1 (1955)
Piping (HPSWS)
Low Pressure Service Water System Piping (LPSWS)
ASA B31.1. (1955)
Pumps Storage Well Cooling System Pumps (SWCS)
Allis-Chalmers Spec. No. 41-5 63 Component Cooling Water Pumps (CCWS)
Allis-Chalmers Spec. No.41-563 Shield Cooling Pumps (CCWS)
Allis-Chalmers Spec. No. 41-5 63 Liquid Waste Demineralizer Pump (WPS)
Allis-Chalmers Spec No. 41-5 63 Seal Injection System Pumps (SIS)
Allis-Chalmers Spec. No. 41-5 72 l
High Pressure Service Water System Sargent & Lundy Auxiliary Pumps (HPSWS)
Spec. No. W-1924 i
Valves Storage Well Cooling System Valves (SWCS)
ASA B31.1 (1955)
Primary Purification System Valves (PPS)
ASA B31.1 (1955)
Component Cooling Water System Valves (CCWS)
ASA B31.1 (1955)
Seal Injection System Valves (SIS)
ASA B 31.1 (1955)
High Pressure Service Water System ASA B31.1 (1955)
Valves (HPSWS)
Low Pressure Service Water ASA B31.1 (1955)
System Valves (LPSWS)
Storace Tanks (Atmospheric and 0-15 psig)
None A Ubbranklin Research Center A Cheman of The Frennan insatuse
TER-C5 25 7-437 Table 4-2 (d)
Index of Abbreviations for Systems Abbreviations Definitions, ACS S Alternate Core Spray System BIS Boron Injection System BSS Building Spray System CCNS Component Cooling Water System C/?W Condensate /Feedwater System CP Containment Penetration CRDH Control Rod Drive Housing CRDS Control Rod Drive System DHCS Decay Heat Cooling System HPCSS High Pressure Core Spray System HPSWS High Pressure Service Water System IV Isolation Valves LPSWS Iow Pressure Service Water System MDS Manual Depressurization System MSS Main Steam System PPS Primary Purification System RCE Reactor Coolant Pressure Boundary RCS Recirculation System SCS Shutdown Condenser System SIS Seal Injection System SRV Safety Relief Valves SWCS Storage Well Cooling System WPS Water Purification System p Udj Franklin Research Center s om at m vrmen m.aua
~ _ _ -.. _
l TER-C5257-437 i
5.
EVALUATION CF SPECIFIC COMPONENTS 5.1 GENERAL REQUIREMENTS The purpose of this section is to evaluate, for the specific components of the Lacrosse Boiling Water Reactor, how the general code requirements of the current code affect the safety margin to which these components were l
originally designed.
General code requirements are those requirements that apply to all the components discussed in this report (i.e., piping, pressure *iessels, valves, pumps, and tanks).
The following topics were identified in Section 4.1 of Appendix A to be general requiremants that have changed from older codes to the current codes fracture roughness, quality assurance,
' quality group classification, and code stress limits. They will be discussed herein.
5.1.1 Fracture Toughness 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 if it is necessary or not to impact-test the material used for each specific component of the Lacrosse Boiling Water Reactor. The results of this evaluation are compiled in Table 1
5-1.
Data on nil ductility transition temperature (T NDT materials can be found in References 13,14, and 15.
Of 72 items reviewed in Table 5-1:
o Twenty-three items (32%) do not require impact testing o Material used was not specified for 25 components (35%)
o More data are required in order to assess 24 components (33%).
1.
Although discussed in this report, quality assurance is outside the scope of the SEP according to a letter from S. Bajwa to S. P. Carfagno dated December 10, 1981.
-_ch_ Center nkjin Resear l
c-E Table 5-1 Fg]v'
>n Review of Fracture Toughnesa Requirements taCrosse nosiing water Reac: tor itaCawRi
!E
- e. 3N Structures, Systems, Quality Group Impact Test Beason for h
and Components Classification Material Required?
Exemption (1)
Remarks N
RECIROHATION SYSTD4
[
- h 16" and 20" Outside Class A Carbon steel Insufficient No information Diameter (O.D.) Reactor AS1H A212-B data on Tg3y Circulating Pump Suction bare material available Piping from Beactor (piping)
Circulating Water Outlet Connections (Including Stainless steel No Se 16" O.D. Cross-Connec-AS1H A351-CF8M tions) to (and (valves) g Including) Valves cn FRV-101 and FRV-103 8
in Each Case l
16" and 20" O.D.
Class A Carbon steel Insufficient No information Reactor Water Circu-AS1H A212-B data on Tg>g lating Pump Discharge bare material available
{
Piping f rom (and (piping)
Including) Valves FRV-102, FRV-104, Stainless steel No Se FRV-105, and FRV-106 AS1H A351-CF8M to Reactor Circula-(valves) ting Water Inlet Connections (Includ-ing 16" O.D. Cross-Connections) in Each Case 4
1.
Refer to Tables A4-4 through A4-6 of Appendix A for explanation of exemptions.
g b
Y 4
W 4
.i
Table 5-1 (Cont.)
3
/
Do 2E s 5' Structures, Systems, Quality Group Impact Test Reason for an.1 Components Classification Material Required?
Exemp ion (1)
Remarks 20" Reactor Water Class A Carbon steel Insufficient M2 information fy
- r Circulating Pump AS M A212-8 data on TNOT Q
Suction Piping from bare material available R
(but not Including)
Valves FRV-101 and FRV-10) to Circula-ting Pump Inlet connection 20" Reactor Water Class A Carbon steel Insufficient No information Circulating Pump AS W A212-B data on THDT Discharge Piping bare material available N
from Circulating Y
Pump Outlet Con-nection to (but not Including)
Valves FRC-102, FRC-104, FRC-105, and FRV-106 Vents and Drains from Class A Carbon steel Insufficient Size not Reactor Circulating AS m A212-B data specified Pumps and from the Above bare material Piping (Previous Two Items) to and Including the Second Valve in Each Case Balance of Piping Class A Carbon steel Insufficient Size not ASM A212-B data specified g
bare material 1
Rotary Cone Type Class A Cast Steel Insufficient N0 information on Valves AS m 216 data
% T available M
Y.
a La3 4
.I
l#n IF'
> n h' l
g Table 5-1 (Cont.)
j 3
E a3g Structures, Systems, Quality Group Impact Test Reason for e
q and Components Classification Material Required?
Exemption (1)
Remarks heirculation System Class A 54 chrome Insufficient No information on Pumps - Casing steel data THDT available aR 4
utEHGENCY SYSTEMS Boron Injection System Sodium Pentaborate Tank Class B Stainless No Se Allis-Chalmer s l
steel Spec. No.41-570 l
Type 304 I
l
'd Chemical Mixing Tank Class B Carbon steel Insufficient Material thickness I
AS'IM A285 data not given Piping and Valves Class B Not given Not discussed in FSAR 1
1 liigh Pressure Core l
Spray System l
l Pumps - Casing Class B tut given Allis-Chalmers I
Spec. No.41-572 overhead Storage Tank Class B Carbon steel Insufficient No information on ASni A201 data THDT available Grade "B" Piping and Valves Class B tut givesa Not discussed in nu d
Core Spray lleaders Class B Not given Not discussed in Y.
Lab 4
.I
i Table 5-1 (Cont.)
m.N
~
"a P
[h cs Structures, Systems, Quality Group Impact Test Reason for
[
and Components Classification Material Required?
Exempt ion (1)
Remarks asN Alternate Core
!l $
S Spray System
!S Diesel-Driven Pumps -
Class B Cast iron Insufficient Size and type of Q
Casing data steel not given Piping and Valves Class B Not given Ibt discussed in FSAR huilding Spray Systen Piping and Valves Class B Not given Not discussed in s
FSAR w
I mnual Depressurization System i
Piping and Valves Class A Not given W L mentioned in FSAR 1
SAFETY RELIEF VAINk$
Class A Not given Not mentioned in FSAR REAC10R (DOLANT PRESSURE BOUNDARY (HCPB)
Piping from Reactor Class A Not given Not mentioned in Vessel up to and Includ-FSAR ing First Isolation Valve External to Containment H
ISOLATION VALVES Valves Other than Those Class A Not given Not mentioned in g
Identified Under FSAR q
Containment Penetrations
{
Valves and Piping w
i 4
.e
l i
Table 5-1 (Con t. )
c=.
t=2 ET Ps }/
>n Structures, Syste=s, cuality Group rapact Test Re. son for lh and Components Classification Material Required?
Eweaption (1)
Remarks L3 2,7 Itotary Cone Type Valves Class A Cast steel Insufficient Ib information on Ty AS'IN A216 data TNDT available R
HW
- r CuttrAINMENT PENETHATIONS Class B tbt given Ibt mentioned in
- ?.
VAINES Ato PIPING FSAR I4 OltfrROL ROD DRIVE Class A Not given Not discussed in HOllSING FSAR CONTHOL ROD DRIVE Class B tut given Not discussed in SYSTEN FSAR g
S1DRACE WEf,L COOLING W
SYSTEN g,
Pumps - Casing Class C Stainless No 8e steel 316 lieat Exchangers Class C Not given Not discussed in FSAR Piping and Valves Class C Not given Not discussed in FSAR (DHDENSATE/FEEIMATER SYSTIN Piping from Outermost Class B lbt given Not discussed in Containaent Isolation FSAR Valve up to and Includ-ing the Shutoff Valve and Connected Piping up to and Including i
the First Shutoff Valve g
Y
.W 4
.--.a O
J
{E' Table 5-1 (Cont.)
f]E s
B Structures, Systems, Quality Group Impact Test Reason for and Components Classification Material Required?
_ Exemption (1)
Remarks J
gg MAIN STEAM SYSTM
$3 Main Steam Piping to Class B Austenitic No 8e Allis-Chalmers gg Main Turbine-Generating stainless steel Spec. No.41-585
- g Unit, from Reactor AS1M A376 (pipe)
Steam Outlets up to and Including Valve MSV-101 Austenitic Ho 8e Allis-Chalmers and MSV-101 Bypass stainless steel Spec. No.41-585 Valve A351, Grade CF8M (gate, globe, angle, and check valves 1/2"
{
and larger) w 8
Stainless steel No 8e No information on A182, Grade F316 TNDT available.
(globe, angle, Allis-Chalmers and check Spec. No.41-585 valves - 2* and smaller)
Main Steam Piping to Class B Austenitic No Se Allis-Chalmers Reactor Shutdown stainless steel Spec. No.41-585 Condenser, from Reactor ASDt A176 (pipe)
Steam outlets Complete to a Point Outside the Austenitic No 8e Allis-Chalmers Biological Shield stainless steel,
Spec. No.41-585 A351 Grade CF8M, (gate, globe, angle, and check g
valves 1/2" and larger)
Stainless steel No 8e No information on y
A182, Grade F316 TNDT available.
9 (globe, angle, Allis-Chalmers 1
and check valves Spec. No.41-585 l
- 2* and smaller) i l
i
1 l
Table 5-1 (Cont.)
11h f$
5 EE Structures, Systems, Quality Group Impact Test Reason for and Components classification Material Required?
Fuesption (1)
Remarks u@
8e Allis-Chalmers g
Reactor Vent and Drain Class B Austenitic No
[8 Lines and Vents and stainless steel Spec. No.41-585 i
Drains from the Above AS M A376 (pipe) e
- g Piping (Two Previous g
- i Itess) to and Including Austenitic Ho Se Allis-Chalmers the Second Valve, in Each stainless steel Spec. No.41-585 Case, also to a Point A351 Grade CF8N just Outside of the (gate, globe, Biological Shield angle, and check valves 1/2" and larger) 1 Stainless steel No Se No information on U
A182, Grade F316 TNDT available.
8
, globe, arqle, Allis-Chalmers
(
and check Spec No.41-585 valves - 2" and smaller)
Reactor and Shutdown Class B Austenitic No 8e Allin-Chalmers condenser Tube Side stainless steel Spec. No.41-585 Water Level Indicator ASM A376 (pipe)
Piping, Hydrostatic Test Connections, Calorimeter Austenitic No 8e Allis-Chalmers Connection to Main Steam stainless steel Spec. No.41-585 Line and All Other A351 Grade CF8M Miscell&neous Lines (gate, globe, Connected with angle, and check Piping Mentioned valves 1/2" Above or with Reactor, and larger) g Where such Lines are not Covered Elsewhere Stainless steel No Se No information on All Within the Biologi-A182, Grade F316 TNDT available.
cal Shield and/or to a (globe, angle, Allis-Chalmers y
Point Outside of the and check Spec. No.41-585 J
Biological Shield valves - 2" 1
smaller) y'
. 4
4 s
E='
>,i k Table 5-1 (Cont.)
5 ?E EL 3 N
Structures, Systems, Quality Group Impact Test Season for t
and Components classification Material Required ?
Exemption (1)
Remarks R
c I
Main Steam Piping to Class B Austenitic No 8e Allis-Chalmers h
Reactor Shutdown Con-stainless steel Spec. No.41-585 k
denser, from a Point AS1M A335 (piping)
Outside the Biological Shield Complete to Stainless steel Insufficient No information on i
Shutdown Condenser, AS1M A217 (gate, data Tuor available.
l Including the Associ-globe, angle, Allis-Chalmers ated Safety Valve Con-and check valves Spec. No.41-585 nections up to and 1/2*and Including the Safety larger)
Valve, in Each Case U
Stainless steel Insufficient No information on i
8 A182, Grade F11 data TNDT available.
l (globe, angle, Allis-Chalmers and check valves Spec. No.41-585
- 2" and smaller)
Main Steam Piping Class B Austenitic No 8e Allis-Chalmers to Main Turbine-stainless steel Spec. No.41-585 Generating Unit, from ASTM A335 (piping) 1 (but not Including)
Valve MSV-101 and Stainless steel Insufficient No information on i
MSV-101 Bypass Valve A217 (gate, globe, data TNDT available.
to a Point Outside of angle, and check Allis-Chalmers the Containment Vessel, valves 1/2" Spec. No.41-585 as Indicated on the and larger)
Drawings Stainless steel Insufficient No information on 3
A182, Grade F11 data TMDT available.
g (globe, angle, Allis-Chalmers 1
and check valves Spec. No.41-585 lA
- 2" and smaller)
N Y
au
.I
![-;
Table 5-1 (Cont.)
rr D
ce
[
Structures, Systems, Quality Group Impact Test Reason for s 5' and Ccaponents Classification Material Required?
Exemption (1)
Remarks N
?$
Vents and Drains from Class in Austenitic Ho 8e Allis-Chalmers R
the Above Piping (Two stainless steel Spec. No.41-585
,:r Previous Items) to and ASIM A335 (piping)
[Q Including the Second R ?.
Valve, in each case Stainless steel Insufficient Ib information on 2
A217 (gate, globe, data got available.
angle, and check Allis-Chalmers valves 1/2" Spec. No.41-585 and larger)
Stainless steel Insufficient No information on A182, Grade Fil da ta
%gy available.
e (globe, angle, Allis-Chalmers W
and check valves Spec. No.41-585
- 2" and smaller)
All Other Miscellaneous Class B Austenitic tb 8e Allis-Chalmers Lines Connected with stainless steel Spec. No.41-585 Piping Items C and D ASTM A335 (piping)
Above and not Covered Elsewhere in These Specifications PHIMARY E41RIFICATION Class C Not given Not discussed in SYSTEM FSAR DEX'AY llEAT (DOLING SYSTEN lleat Exchanger H
Tube Side Class B tut given Not discussed in N
FSAR Shell Side Class C Not given Not discussed in su y
.s.
W 4
'l
'='
Table 5-1 (Cont.)
/
5E SLguctures, Systems, Quality Group Impact Test Heason for E
and Components Classification Material Required?
Exe mpt ion (1)
Remarks J
ij $
CCailONk24T (DOLING WATER
[
SYSTEM In gg Pump Casing Class C Cast steel Insuf ficies.t Size and type of g
data steel not given Tank Class C Carbon eteel Insufficient Size and type of data steel not given Piping and Valves Class C Not given Shield Cooling Pump Class C Austenitic W
8e 1
Casing stainless steel u'
Type 316 shield Cooling Tank Class C Carbon steel
. Insufficient Size and type of data steel not given SIRJ11X443 CONDENSER Class B Not given Not discussed in SirSTtN FSAR WATD4 PURIPICATION SYSTEM Liquid Waste Deminera-Class C Austenitic W
8e lizer Pucp Casing stainless steel 31o Spent Hesin Tank Class C Stainless steel No 8e Type 304 8
N SEAL INJECFION SYSTEN
[
un Pumps Class C Not given Not discussed in U
Nu y
Las 4
e e f
lL aa
>r
/
n 2E e5 M
E'$
Table 5-1 (Cont.)
4
$n7 Structures, Systems, Quality Group f(*
and com;unents Impact Test Reason for s ?.
classification Material Required?
Exemption (1)
Bemarks E
Surge Tank Class C Carbon steel Insufficient Size and type of data steel not given
/
Reservoir Tank Class C Carbon steel Insufficledt Stae and type of w,
data g
.. steel not givent Piping and Valves Class C I
ht given J
llot discussed in w
s
^
as I
FSAR HIQ1 PHESSURE SERVICE y-WATER SYSTEM Auxiliary Pumps Class C Cast iron Ins af ficient Sise and type of data steel not given Piping and Valves Class C Not given blot discussed in FSAR I4W l'RESSURE SERVIG Class C WA'IER SYSTEM Nt given Not discussed in PSAR H
Y c
m.
M i
4
+1
-C
I s
TER-C5 25 7-437 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 c, odes [7, 8, 9,10].
Quality assurance requirements for the design, and construction, fabrication, and installation of components at the Lacrosse plant are addressed in Section 4.4 of the PSAR [5].
5.1.3 Quality Group 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 [7] or in the ASME B&PV Code,Section VIII, 1962 Edition
[10]. USAS B31.7 [8] classified piping as Class I, II, or III.
The ASME D&PV Code Section VIII [10] in conjunction with Code Case 1270N classifies vessels in two different categories, primary vessels and secondary vessels.
Prinary vessels were defined as vessels which contain reactor coolant and are equivalent to current Class 1 vessels.
Secondary vessels do not' contain reactor coolant and are not subject to irradiation and are equivalent to current Class 2 and 3 vessels.
Note in Table 4-2(a) that the control rod drive housing, which is currently a Class 1 pressure vessel, was designed to Section VIII requirenents, - but Code Case 1270N was not invoked.
Note in Table 4-2(b) that the decay hest cooling system heat exchanger (tube side) and the overhead storage tank, which are currently Class 2 pressure vessels, were designed to Section VIII with Code Case 1270N invoked for the decay heat cooling '"atem heat exchanger only. The sodium pentaborate tank, a Class 2 pressurs s
- sal, was not designed to Section VIII requirements.
Similarly, in Table 4-2 (c),
all current Class 3 pressure vessels -- storage well cooling system heat 1.
Although discussed in this report, quality assurance is outside the scope of the SEP as indicated in a letter dated December 10, 1981 from S. Bajwa to S. P. Carfagno.
A !!Obd Frank!!n Research Certer A Dmean of The Frenasen ktseewie
TER-C5 25 7-437 exchanger, decay heat cooling system heat exchanger (shell side), component cooling water system tank, shield cooling tank, spent resin tank, seal injection surge tank, and seal injection reservoir tank --
were designed to Section VIII.
Code Case 1270N was invoked for the storage well cooling system heat exchanger and the shell side of the decay heat cooling system heat e xchang er.
CurrentlIy classified Class 1, 2, and 3 pressure vessels (secondary vessels) should be evaluated against t.crent radiography requirements (see discussion on full radiography in Section 5.2 of this report). Class 1 pressure vessels sho'ild be evaluated against current fatigue analysis requirements (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.
Design based on the old piping code [7] and the ASME B&PV Code,Section VIII [10), was based on the maximum normal stress theory of failure as compared to design based on the maximum shear stress theory of failure of the current code [2] for Class 1 components.
The maximum shear stress theory currently used is advantageous for analysis because it is less conservative and it facilitates a more precise fatigue analysis.
The current code for Class 2 and Class 3 components uses the same theory of failure as past codes.
Consideration of stress limits for equivalents of service levels C and D is not mentioned in the Lacrosse 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.
5.2 PRESSURE VESSELS As discussed in Appendix A, Section 4.3, major differences between current requ irements [2] and old requirements [10] for tha construction of pressure vessels appear in four areas:
fracture toughness, quality group classifica-tion, design, and full radiography requirements.
& bil Frank!!n Research Center A Dewson of The Frenen insoluse r
TER-C5257-437 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 stress limits and consideration of service leve]
C and D loading conditions. This topic is addressed in Section 5.1.4 of this report.
Full radiograph requirements for pressure vessels are discussed in Section 4.3 of Appendix A.
The conclusion to be drawn is that, in general, past full radiography requirements were dependent on material of construction, total plate thickness of the vessels, code cases invoked for design purpose, and whether or not the vessel contained lethal radioactive substances.
Section VIII required that all joints whose material thickness exceeded 1-1/2 inches be fully radiographed. All longitudinal and circumferential welded joints with material thickness less than 1-1/2 inches should be fully radiographed as described in Paragraphs UCS-57, VHA-33, and UCL-35 of Reference 10.
Presently classified Class 1 vessels were referred to as primary vessels (containing radioactive lethal substances) when constructed to Section VIII (10] and Code Case 1270N.
All longitudinal and circumferential welded joints of primary vessels were fully radiographed when designed to Section VIII (10]
with Code Cases 1270N and 1273N invoked.
Information regarding the radiography requirements imposed on all the control rod drive housing should be provided for review against current radiography requirements (see Table 4-2 (a)). A primary vessel (Code Case 1270N) designed to Section VIII requirements with Code Case 1273N invoked would meet current full radiography requirements.
Fatigue analysis was not required for vessels built according to Section VIII.
All vessels currently classified as Class 1 should be evaluated for current fatigue analysis requirements. Discussion on current fatigue analysis requirements is provided in Section 5.4 of this report.
For the control rod drive housing, the Licensee should provide the following:
proof that the five conditions outlined in Section 5.4 of this report a.
were met and, therefore, analysis for cyclic loading is not required, or %dd Franklin Research Center
- cm on er n. re.a.n m.au.
c l
TER-C5257-437 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.
Vessels built according to Section VIII [10] with Code Case 1270N invoked and formerly classified as secondary vessels (not containing radioactive lethal substances) are currently categorized as Class 2 and 3 vessels.
All longitudinal and circumferential welded joints of secondary vessels with material thickness exceeding 1-1/2 inches were fully radiographed when designed to Section VIII [10]. All longitudinal and circumferential welded joints with material thickness less than 1-1/2 inches should be fully radiographed as described in Paragraphs UCS-57, VHA-33, and UCL-35 of Reference 10.
Information regarding the radiography requirements imposed on the all Class 2 and 3 vessels should be provided for review against current radiography requirements.
5.3 PIPING In addition to the general requirements previously discussed, the follow-ing items are considered in the design of Class 1 piping for fatigue stresses based on the current code [2] but were not considered or were considered differently in the past code [7]:
o Gross discontinuities in the piping systems are accounted for Loading due to the thermal gradient through the thickness of the pipe o
is considered o Indices used in calculating secondary ctresses 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 is concerned ar.d are discussed in detail in Section 4.2 of Appendix A.
When considering gross discontinuities of piping systems, two loading cases can prove to be potentially unconservative designs when evaluated to 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.
A Uddd Franklin Research Center 8 0~~w a ru veme mmu,
TER-CS257-437 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.
Vessels built according to Section VIII (10] with Code Case 1270N invoked and formerly classified as secondary vessels (not containing radioactive lethal substances) are currently categorized as Class 2 and 3 vessels. All longitudinal and circumferential welded joints of secondery vessels with material tnicxness exceeding 1-1/2 inches were fully radiographed when designed to Section VIII (10]. All longitudinal and circumferential welded joints with material thickness less than 1-1/2 inches should be fully radiographed as described in Paragraphs DC5-57, VHA-33, and UCL-35 of Reference 10.
Information regarding tne radiography requirements imposed on the all Class 2 and 3 vessels should be provided for review against current radiography requirements.
5.3 PIPING In addition to the general requirements previously discussed, the follow-ing items are considered in the design of Class 1 piping for fatigue stresses based on the current code [2] but were not considered or were considered cifferently in the past code (7):
o Gross discontinuities in the piping systems are accounted for o Loading due to the thermal gradient through the thickness of the pipe is considered o Indices useo 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 is concerned and are discussed in detail in Section 4.2 of Appendix A.
When considering gross discontinuities of piping systems, two loading cases can prove to be potentially unconservative designs when evaluated to 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.
-4 0 -
M' Franklin Resear.ch Center J
m on or rh. r,.a. a%.
l
TER-C5257-437 These examples are baseo on Palisades specifications [161, Stresses for both examples indicate that no problem exists. Calculations similar to those presented in examples 1 and 2, Section 4.2 of Appendix A, applicable to the Lacrosse plant design parameters should be provided in order to assess the I
impact on the usage factor of gross discontinuities in recirculation system piping, manual depressurization system piping, and piping from the reactor vessel to the first isolation valve external to the containment.
Discussion of coce cases used in conjunction with ASA B31.1 (1955) is provided in Section 4.2 of the appendix 4 In general, piping designed to ASA B31.1 (1955) (7) with Code Case N-1 invoked had safety requirements imposed for loss of radioactive fluid.
If Code Case N-7 I was invoked for piping made from austenitic stainless steel, full radiography requirements and allowable stress values up to a temperature of 650*F would meet current requirements. Code Case N-9 provided guidelines for centrifugally. cast austenitic steel for nuclear service and required fully radiographed welded I
joints. Code Case N-10 permitted the use of cast austenitic butt-welded fittings for nuclear service and required full radiography. Stress allowables for Code Cases N-9 and N-10 meet current stress allowables for a temperature range up to 650*F.
According tc a study conducted by the Licensee (4), the recirculation system piping meets the design criteria of USAS B31.7 (8].
Piping systems designed to USAS B31.7 do not comply with current coce (21 requirements for the following reasons:
Fracture toughness requirements were not considered (see Section G.1.1) o Stress limits for equivalent service levels C and D conditions were o
not specified (see Section 5.1.4)
Stress indices for some cases were lower than current code o
specifications.
1.
Mechanical Engineering, August 1962 (Code Cases N-1, N-7), December 1960 and October 1964 (N-9), and April 1960 (N-10).
1 43m
~
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TER-C5257-437 Piping systems designed to USAS B31.7 comply with current requirements as follows:
o Piping is classified as Class I, II, or III
~
o Shear theory of failure is used The formula "for peak stress intensity ranges for Class I piping is o
used, and local and secondary stress indices are considered o Full radiography is required for circumferential and longitudinal butt welds for Class I and II piping.
Recirculation system piping should be reviewed for the differences outlined and evaluated against current requirements.
For the remaining Class 1 and 2 piping systems designed to ASA B31.1 (1955), information on the radiography requirements imposed should be provided in order to determine if the systems meet the current requirements.
5.4 PUMPS Table 4-2(a) shows that Class 1 pumps (recirculation system pumps) were designed to Allis-Chalmers Specification No.41-569 (4].
Table 4-2(b) shows that Class 2 pumps were designed to Sargent & Lundy Specification W-1924 and Allis-Chalmers Specification No.41-572 [4].
Table 4-2(c) shows that Class 3 pumps were designed to Allis-Chalmers Specification Nos.11-563 and 41-572 and Sargent & Lundy Specification W-1924 (4].
No information was provided regarding design codes usec for any of the plant pumps listed in Table 4-2.
The Licensee should provide additional design information with reference to the current requirements imposed on pumps.
Items to be reviewed regarding pumps are general requirements as discussed in Section 5.1 of this report and full radiography requirements discussed in Section 5.2 of this report, and tne fatigue analysis discussed herein.
Information on the radiography requirements impcsed on the welds of Class 1 and 2 pumps listed in Tables 4-2(a) and 4-2 (b) should be provided and compared with current requirements given in Section 4.2 of Appendix A. 4 dd Franklin Research C. enter 4 o m.ona m. r m a m. n,.
TER-CS257-437 The recirculation system are pumps currently classified as Class 1 pumps. Class 1 requirements specify fatigue analysis if a set of conditions are not met (see NB-3222-4 (d) of Reference 2).
If any one of the following conditions is not met, the recirculation system pumps should be analyzed for cyclic loads:
(1)
Pressure Fluctuations:
the specified full range of pressure fluctuations during normal service does not exceed:
(1/3) (Design Pressure) (S /S )
a m where:
Sa = alternating stress from fatigue curves corresponding to the number of pressure fluctuations Sm = allowable stress intensity at the service temperature (2)
Atmospheric to Service Pressure Cycle N2 < N(3S )
m where:
N2 = the maximum number of atmospheric to service pressure cycles N(3S ) = number of cycles from design fatigue curve for Sa " 3Sm m
(3)
Temperature differences between adjacent points, i.e., two points along (1/2)the meridian of a vessel, nozzle, or flange closer than 2 ( Rt) where P. is the mean radius and t is the mean thickness between the two points:
l i <,S /(2Es) (i = 1,2) aT a
where ATi = temperature differences between two adjacent points i = 1:
startup and shutdown Ldj Franklin Research Center 4 o m.on or n. rr.n.e w..u.
TER-C5257-437 i = 2: normal service E = modulus of elasticity at mean temperature between points a = instantaneous coefficient of expansion, mean value (see Table I-5.0 of Reference 2)
S, = alter,nating stress from uesign fatigue curve corresponding to the numoer 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 tnan S/(2EG) where S is the endurance limit, i.e.,
6 the value of S from the fatigue curve at 10 cycles.
a (4)
Temperature difference - dissimilar materials - see paragraph NB-3222. 4 (d) (4) of Reference 2 (5)
Meenanical loads - stresses due to mechanical load fluctuations (excluding pressure) such as pipe loads on nozzles less than the value of S from tne design fatigue curve corresponding to the a
number of load fluctuations.
The Licensee should provide the followings 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 nine pumps reviewed in this report, none was designed to Section VIII of the ASME B&PV Code.
Information is needed on the pumps that were 7ot designed to ASME B&PV Code, in order to determine if they meet current standards. Codes, code classes, editions, code cases, design calculations, and radiography requirements should be provided for all pumps in the Lacrosse I
plant.
5.5 VALVES Major differences between current requirements (2) and past requirements (7) for valves are discussed in Section 4.5 of Appendix A.
Class 1 valves designed in accordance with past requirements should be adequate wnen Judged by current standards except for: Jb Franklin Research Center 4o-.ono m.r e %
]
l TER-CS257-437 1.
fracture toughness requirements 2.
stress limits might not be satisfied for valves that differ significantly from the body shapes described in the current code
[2, 17]
3.
stress limits for service level C might 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 Lacrosse plant:
1.
See Table 5-1 for the fracture toughness requirements evaluation.
2.
Compare actual body shape of valves with body shape rules of Section NB-3544 [2].
If significantly different, the Licensee should provida 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 was necessary.
If this is true, the previous item 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 Lacrosse plant (see Table 4-2(b) and 4-2 (c) ) should be compared with the current pressure-temperature rating [17].
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, Code Case N-1 in combination with cedes cases N-7, N-9, or N-10 (
to Reference 7 required full radiography for circumferential and longitudinal welds. If these code cases were applied, then current full radiography requirements are met.
Using Table 4-1, the Licensee should provide information indicating which code case were invoked.
1.
Mecnanical Engineering, August 1962 (Code Case N-1, N-7), December 1960 and October 1964 (N-9), and April 1960 (N-10). MJ Franklin Research Center A em,an en N rr n ma,,
TER-C5257-437 ASA B16.9 (1958) [11] used in fittings design provides overall dimensions, tolerances, and markings for drought carbon-and alloy-steel factory-made welding fittings.
It refers to ASA B31.1 [7] for design requirements.
ASA B16.10 (1957) [12] used in valve design provides face-to-face and end-to-end dimensions for ferrous valves of various types and ferrous buttwelding end valves.
ASA B16.9 [11] and ASA B16.10 (12] do not provide design guidance for fittings or valves. Valves and fittings built to these standards should be.
evaluated against the current requirements [2,17].
5.6 STORAGE TANKS As discussed in Section 4.7 of Appendix A, atmospheric storage tanks designed to ASME B&PV Code, VIII (1962), should be evaluated to determine if the current compressive stress requirements are met.
As discussed in Section 4.7 of Appendix A, welded aluminum alloy tank shells (permitted for Class 3 atorage tanks only) may be overstressed by as much as 184, and roof support bolts may not satisfy current material requirements. Storage tanks designed to American Water Works Association (AWWA) Standard D100,1959 Edition (18] should be evaluated to determine if they meet current compressive stress requirements. One storage tank was reviewed in this report, and it was designed to AWWA D100 [18).
-ec31s dud Franklin Research Center a w orn.r,.n.en m u.
l TER-C5257-437 6.
CONCLUSIONS AND RECOMMENDATIONS A comparison of the design standards in effect during the design and construction of the Lacrosse Boiling Water Reactor against current standards indicates differences in the following areas:
fracture toughness requirements, quality assurance requirements,I I 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 limits 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.
Recommendations are given in Section 5 of this report with regard to the necessity for addition.=1 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 recommendations is as follows:
1.
Fracture toughness - Seventy-two items in Table 5-1 were reviewed to determine if impact testing was required. From the information in this table, it can be concluded:
o Twenty-three items (32%) do not require impact testing o Material used was not specified for 25 components (35%)
o More data are required in order to assess 24 components (331).
2.
Full radiography requirements - Information should be provided on the radiography requirements imposed on Class 1 vessels not designed as primary vessels (Code Case 1270N) and for which Code Case 1273N was not invoked.
Information on the radiography requirements imposed on Class 2 and 3 vessels for which Code Case 1273N was not invoked and with welded joint thicknesses less than 1-1/2 in should be provided.
Piping systems designed to ASA B31.1 [7] and Code Case N-1, in combination with Code Cases N-7, N-9, and N-10, meet current full 1.
Although discussed in this report, quality assurance is outside the scope of the SEP as indicated in the December 10, 1981 letter from S. Bajwa to S. P. Carfagno. UUUd Franklin Research Center A om.on or m r,.non m.m.,
TER-CS257-437 radiography requirements.
Information on the radiography requirements imposed on Class 1 and 2 piping and valves designed only to ASA B31.1 (1955) should be provided in order to determine if these components meet current radiography requirements.
If the piping systems were designed to USAS B31.7 (1968) [8], full radiography requirements were met.
Information on the radiography requirements imposed on Class 1 and 2 pumps should be provided in order to determine whetner the pumps meet current radiography requirements.
Tables 4-2 (a), (b), and (c) should be used in providing the required information.
3.
Pressure vessels - Information regarding the radiography requirements imposed on all Class 1, 2, and 3 pressure vessels is required.
Additional data on the materials of construction are needed for the fracture toughness evaluation. Proof of compliance with current fatigue analysis requirements for Class 1 vessels (Table 4-2(a))
should be provided per the discussion in Section 5.4 of this report.
4.
Piping - In addition to the impact testing and full radiography requirements previously discussed, calculations similar to those presented in examples 1 and 2, Section 4.2 of Appendix A, applicable to the Lacrosse plant design parameters, should be provided in order to assess the impact on the usage factor of gross discontinuities in Class 1 piping systems (Table 4-2(a)) for a medium and large number of cyclic loans.
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 evaluate if they meet current body shape and pressure-temperature rating requirements as discussed in Section 5.5 of this report. Valves designed to ASA B16.9 and ASA B16.10 should be evaluated against the current requirements (2, 17].
6.
Pumps - None of the nine pumps reviewed in this report was designed to Section VIII of the ASME B&PV Code.
Information is needed on these pumps in order to evaluate if the current requirements are met. Codes, code classes, editions, codes cases, design calculations, and radiography requirements should be provided for all pumps in the Lacrosse plant. Proof of compliance with current fatigue analysis requirements, as discussed in Section 5.4 of this re por t, for current Class 1 pumps (the recirculation system pumps) should be provided.
7.
Storage tanks - (i) Atmospheric s torage tanks should be evaluated to determine if they meet current compressive stress requirements; (ii)
O to 15 psig storage tanks should be evaluated to determine if they meet current tensile allowables for biaxial stress field conditions; (iii) additional information and calculations for storage tanks i,dd Franklin Research Center A Cennion ci The Freman ensoeute
TER-C5257-437 designed to AWWA D100 should be provided to determine if they meet current standards. One storage tank was reviewed in this report, and it was designed to AWWA D100 requirements.
8.
Missing information - The following information, which is incomplete or missing from Table 4-1 or Tables 4-2(a), (b), and (c) of this report, should be provided:
1.
information on codes, class, and code cases used in the design of 9 out of 59 components (Table 4-2) ii. any specifications or calculations used in designing pumps, valves, and tanks that may assist in conducting this evaluation 111. confirmation of assumptions made regarding code editions (see Table 4-1) iv.
provision of fatigue analysis calculations for Class 1 piping systems similar to examples based on the Palisades Specifications given in Section 4.2 of Appendix A v.
clarification or additional information on notes 5 and 6 in Table 4-1. Ubd Franklin Research Center a w a n. ramuu.
TER-C5257-437 7.
REFERENCES 1.
NRC
" Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants" Revision 3, February 1976 Regulatory Guide 1.26 2.
American Society of Mechanical Engineers
" Boiler and Pressure Vessel Code,"Section III, Division 1 New York 1977 Edition and addenda through Summer 1978 3.
Title 10 of the Code of Federal Regulations Section 50.55a, " Codes and Standards" Revised January 1,1981 4.
F. Linder (DPC)
Iatter to D. M. Crutchfield (NRC) with Table of Systems and Components, dated August 6,1981 and received at NRC on Augus t 14, 1981; F. Linder (DPC), Letter to D. M. Crutchfield (NRC) with additional information on Table of Systems and Components, dated March 9,1982 and received at NRC on March 18, 1982; M. Boyle (NRC), Telephone Memorandum to A. Gonzalez (FRC), March 28, 1982 5.
Final Safety Analysis Report for Lacrosse Boiling Water Reactor (3 Volumes)
Docketed USAEC, October 9, 1974 Docket No. 50-409 6.
NRC Standard Review Plan Section 3.2.2, " System Quality Group Classification" office of Nuclear Reactor Regulation NUREG-75/087 7.
American Standards Association l
" Code for Pressure Piping" Published by the American Society of Mechanical Engineers,1955 l
ASA B31.1-1955 -
8.
American Society of Mechanical Engineers "Draf t Code for Pressure Piping, Nuclear Power Piping" USAS B31.7, February 1968 i
2.
American Standards Association
" Steel Pipe Flange and Flanged Fittings" ASA B16.5-1961 American Society of Mechanical Engineers,1961
( ~
_nklin Rese_ arch Center
TER-C5257-437 10.
American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section VIII, "Unfired Pressure Vessels" 1962 7
11.
American Standards Association
" Factory Made Wrought Steel Buttwelding Fittings" ASA B16.9-1958 American Society of Mechanical ingineers, 1958 12.
American Standards Association
" Face-to-Face and End-to-End Dimensions of Ferrous Valves" ASA B16.10-1957 American Society of Mechanical Engineers 13.
Snaider, R.
P., Hodge, J. M.
Levin, H. A. and Zudans, J. J.
" Potential for Low Fracture Toughness and Lamellar Tearing on PWR Steam Generator and Reactor Coolant Pump Supports" Published for Comment, October 1979 NUREG-0577 14.
Electric Power Research Institute
" Nuclear Pressure Vessel Steel Dats Base" Prepared by Fracture Control Corporation Palo Alto, CA:
December 1978 NP-933, Research Project 886-1 15.
American Society for Metals j
"Me tals Handbook" Metals Park, Ohio: Ninth Edition,1979 16.
Final S-fety Analysis Report for Consumers Power Company, Palisades Plant (3 volumes)
Docketed USAEC, November 5,1968 Docket No. 50-255 17.
American National Standards Institute
" Steel Valves" l
American Society of Mechanical Engineers,1977 l
American Water Works Association "AWWA Standard for Welded Steel Elevated Tanks, Standpipes and Reservoirs for Water Storage" New York:
1959 AWWA D100 l
l l
l f O
ranklin Re?estch Center a won on rh. Fr.n u.o in u.
n
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l l
APPENDIX A 1
1 REVIEW OF CODES AND STANDARDS APPLICABLE TO LACROSSE, SAN ONOFRE, AND HADDAM NECK PLANTS l
l l
4 Franklin Research Center A Division of The Franklin Institute l
The Benprnen Franklin Parkway. PNia.. Pa 19103 (215)448 '000 l
CDNTENTS Section Ti ti,e_
Pege 1
INTRODUCTION A-1
'4 2
SUMMARY
OF RESULTS OF CODE COMPARISON 2.1 General A-4
- 2. 2 Piping.
A-4 2.3 Pressure Vessels A-4 2.4 Pumps.
A-4 2.5 Valves.
A-5 2.6 Heat Exchangers A-5 2.7 Storage Tanks.
A-5 3
CONCLUSIONS AND RECOMMENDATIONS.
A-24 4
COMPARISON OF SIGNIFICANT CURRENT (DDE REQUIRD4ENTS AND PAST REQUIREMENTS A-26 4.1 General Requirements A-26 4.2 Piping.
A-57 4.3 Pressure Vessels A-83 4.4 Pumps A-91 4.5 Valves.
A-93 4.6 Heat Exchangers A-98 4.7 Storage Tanks.
A-100 5
BASIS EUR SELECTING REQUIRTMENTs MOST SIGNIFICANT TO COMPONENT INTEGRITY.
A-104 6
REFERENCES.
A-105 iii A-w Ubd Franklin Research Center 4 % or Th. r.m m
T 1.
INTRODUCTION The purpose of this appendix is to compare the code currently used in the design, faorication, 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 II Sunsner 1978 Addenda (1, 2]. The three major older codes being compared against the current code are the B&PV Code,Section VIII,1962 Edition [3] for vessels and the " Code for Pressure Piping," American Standard Association B31.1, 1955 Edition (4] and/or the B&PV Code,Section I,1962 Edition [5] for piping.
Table Al-1 groups the SEP plants according to the major codes used to design them.
In order to take advantage of the similarities in each group, this appendix applies only to the Group II plants: Lacrosse, San Onofre, and Hadoam Necx plants.
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 major identified differences are discussed in Section 4.
The scope 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 generators, and supports are outside the scope of this appendix, as is the seismic classification of systems and components.
All these subjects are addressed in 1.
Ibgether witn Code Cases 1270N, 1271N, 1272N, and 1273N for vessels and/or Code Cases N-1, N-2, N-4, N-7, N-9, N-10, N-ll, and N-12 for piping, when invoked.
I l
A-1 M!! Franklin Research Center A Dmmon of The Frenahn insenAs L
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.I I
~
I i
2.
Ietter from S. Bajwa to S. Carfagno dated December 10, 1981.
A-2 nklin Rese- ~_ arch. Center
Table Al-1 Major Codes and Standards Used in Design of Systems and Components of SEP Plants Commercial Plant Operation Maior Codes Group I (1969-1971)
Palisades Dec. 1971 1.
ASME III (1965)
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 Group II (1968)
Lacrosse Nov. 1969 1.
ASME I & VIII (1962) and Code Cases 1270N, 1271N, 127 2N, and 1273N San Onofre Jan. 1968 2.
ASA B31.1 (1955) and Code Cases N-1, N-2, N-4, N-7, N-9, N-10, N-ll, and N-12 Haddam Neck Jan. 1968 Group III (1961-1963)
Big Rock Point March 1963 1.
ASME I & VIII (1959) 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 nklin Research Center
~ ~ -. -.
2.
SUMMARY
OF RESULTS OF CODE COMPARISON 2.1 GENERAL The current code requirements for the construction of nuclear power plant components (1) are outlined in Table A2-1.
For each article or subarticle, the applicability to Code Class 1, 2, or 3, correspondi.ag 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 tne current and past code requirements for the materials, design, fabrication, examination, and testing of piping systems and components for nuclear power plants. The past codes for piping are the B31.1 (1955) power piping code and Section I of ASME B&PV Code (1962) [5],
" Power Boilers," for piping Detween the reactor vessel up to and including the valve or valves required by ASME I.
A comparison of significant past and current piping requirements may be found in Sections 4.1 and 4.2 of this appendix.
2.3 PRESSURE VESSELS Table A2-3 compares the current code [1] requirements for the materials, design, fabrication, examination, and testing of pressure vessels for nuclear power plants against ASME B&VP Code Section VIII (1962) (3].
A comparison of significa,nt past and current pressure vessel requirements may be found in Section 4.3 of this appendix.
2.4 PUMPS See Section 4.4 of this appendix.
l A-4 dd Franklin Research Center A Ns.on of The Frennha insatute
2.5 VALVES See Section 4.5 of this appendix.
2.6 HEAT EXCHANGERS Heat exchangers were usually designed to the ASME Boiler and Pressure vessel Code,Section VIII [3], which is discussed in Sections 2.3 and 4.3 of this appendix, and passibly to the Standards of the Tubular Exchanger Manufacturers Association (TEMA), 1959 Edition [8].
Discussions regarding TEMA may be found in Section 4.6 of this appendix.
2.7 STORAGE TANKS Storage tanks that must withstand pressures above atmospheric may have been designed to the ASME Boiler and Pressure Vessel Code,Section VIII,1962 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 tanks were also designed to the American Petroleum Institute (API)
Standard 650 (10), 1964 Ed ition. USAS B96.1 and API-650 are discussed in Section 4.7 of this appendix, A-5 g4 UdNnklin Research Center A Dmeson of The Frannan wwotute
Table A2-1 g
Current Code Requirements [1]
14
[h )3
- 71 Article
- I or Class Class Class Signi-gg Subarticle Description 1
2 3
ficant Remarks R3 N
NA-1000 SCOPE OF SECTION III A
A A
e 1
NA-2000 CLASSIFICATION OF COHPONENTS A
A A
A kh NA-3000 RESPONSIBILITIES AND DUTIES A
A A
a.
NA-4000 QUALITY ASSURANCE NA-4100 Quality Assurance Requirements A
A NA A
NA-5000 INSPECTION NA-5100 General Requirements for Authorized A
A A
A Inspection Agencies and Inspectors NA-5200 Duties of Inspectors A
A A
A
{
NA-6000 QUALITY CONTROL SYSTEMS EUR CLASS 3 os CONSTRUCTION NA-6100 General Requirements NA NA A
A NA-6200 Organization and Responsibilities NA NA A
A NA-6300 Control of Operations NA NA A
A NA-6400 Records and Forms NA NA A
A NA-8000 CERTIFICATES OF AUTHORIZATION, A
A A
NAMEPLATES, STAMPING, ANO REPOHTS 1000 INTROOUCTION 1100 Scope A
A A
A 2000 MATERIAL 2100 General A
A A
A 2200 Material Test Coupons and Specimens A
A A
for Ferritic Steel H.sterials A
Addressed in the Code for the specified 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 addressed in the Code for the specified class.
Article number in current Code will be preceded by Nu for Class I component, NC for Class 2 component, and ND for Class 3 cosaponent.
8
l E
Table A2-1 (Cont.)
l's Article
- gg g 5-or Class Class Class Signi-
- c Subarticle Description 1
2 3
ficant Remarks g
m0 l$
2300 Fracture Toughness Requirement A
A A
A d@
5 for Material o
2400 Welding and Brazing A
A A
A 2500 Examination and Repair of Pressure A
A A
4 Retaining Materials 2600 m terial k nufacturers' Quality A
A A
A System Program 2700 Dimensional Standard A
A A
3000 DESIGN 3100 General A
A A
A 3200 Desfgn by Analysis (C1. 1); Alternate A A
NA A
"..aign Rules for Vessels (C1. 2)
- p 8
3300 Jessel Design A
A A
A 3400 Pump Design A
A A
A 3500 Valve Design A
A A
A 3600 Piping Design A
A A
A 3700 Electrical and Mechanical Penetration NA A
A A
Assemblies 3800 Design of Atmospheric Storage Tanks NA A'
A A
3900 0-15 psi (0-103 kPa) Storage Tank NA A
A A
Design 4000 FABRICATION AND INSTALLATIOt3 4100 General A
A A
4200 Forming, Fitting, and Aligning A
A A
4300 Welding Qualifications A
A A
A 4400 Rules Governing Making, Examining, A
A A
and Repairing Welds 4500 Brating A
A A
4600 Heat Treatment A
A A
~~
l _2 Ed' Table A2-1 (Cont.)
- f f3 Article
- 5Ei or 3
Class Class Class Signi-Suba r t icle Description 1
2 3
ficant Remarks
,3 g[g 4700 Mechanical Joints A
A A
g[
4800 Expansion Joints NA A
A f
5000 EXAMINATION g
$100 General Requirements A
A A
A 5200 Hequired Examination of Welds A
A A
A (Cl. 1)3 Examination of Welds (C1. 2 and C1. 3) 5300 Acceptance Standard A
A A
A 5400 Final Examination of items (C1. 1)
A NA A
A Spot Examination of Welded Joints (C1. 3) 5500 Qualifications of Nondestructive A
A A
A I
Examination Personnel DD 5600 NA NA NA 5700 Examination Requirement of NA A
A Expansion Joints 6000 TESTING 6100 General A
A A
6200 Hydrostatic A
A A
6300 Pneumatic A
A A
6400 Pressure Test Gages A
A A
l 6500 Atmospheric and 0-15 psig NA A
A Storage Tanks 6600 Hydrostatic Testing of Vessels NA A
NA Designed to NC-3200 6700 Pneumatic Testing of Vessels NA A
NA Designed to NC-3200 1
6800 6900 Proof Tests to Establish NA A
A Design Pressure O
e 9
Viw hfj Table A2-1 (Cont.)
i Article
- p so or Class Class Class Signi-ik Suba r t icle De scr iption 1
2 3
ficant Remarks 4m h
7000 PROTECTION AGAINST OVERPRESSUkB
.kg r) 7100 General A
A A
7200 Definitions Applicable to A
A A
4 Overpressure Protection Devices 7300 Overpressure Protection Report A
A NA (CL. 133 Analysis (C1. 2) 1400 Relieving Capacity Requirements A
A A
and Acceptable Types of Overpressure Protection Devices 7500 Set Pressures of Pressure Relief A
A A
Devices 2
7600 Operating Design Requirements for A
A A
j3 Pressure Relief Valves 7700 Requirements for Nonteclosing A
A A
Pressure Relief Devices 7800 Certification Requirements A
A A
7900 Marking, Stamping, and Reports A
A A
8000 NAMEPLATES, STAMPING, AND REPCRTS 8100 Ge ne ral A
A A
MANDA10kY APPENDICES I
Design Stress Intensity values, A
A A
A Allowable Stresses, Material Properties, and Design Fatigue Curves II Experisental Stress Analysis A
A A
111 Basis for Establishing Design A
A A
A Stress Intensity Values and Allowable Stress Values
euE
'[.,_
g A
h/
[g Table A2-1 (Cont.)
E3 23P Article *
?$
or Class Class Class Signi-hk subarticle Description 1
2 3
ficant Remarks y(:r i
)
IV Approval of New Materials Under A
A A
y.
the ASME Boller and Pressure Vessel Code for Section III Application V
Certificate Holder's Data Report A
A A
Forms and Application Forms for Certificates of Authorization for Use of Code Symbol Stampa VI Rounded Indications Charts A
A A
VII Charts for Determining Shell A
A A
Thickness of Cylindrical and T
Spherical Components Under
((
External Pressure XI Rules for Bolted Flange NA A
A Connections for Class 2 and 3 Components and Class HC Vessels XII Design Considerations for Bolted A
A A
A Flange Connections XIII Design Based on Stress Analysis NA A
NA for Vessels Designed in Accordance with NC-3200 XIV Design Based on Fatigue Analysis NA A
NA for Vessels Designed in Accordance with NC-3200 XVI Nondestructive Examination A
A A
O Methods Applicable to Core Support Structures XVII Design of Linear Type Supports by A
A A
O Linear Elastic and Plastic Analysis I
i
.f
i t' _Z n
[h v'
> m [/
Eh 1
IG to N
'N jm g3
- r a
kA
{g Table A2-1 (Cont.)
<n Article
- or Class Class Class Signi-Subarticle Description 1
2 3
ficant Remarks NONNANDA1 DRY APPENDICES A
A NA NA B
Owner's Design Specif1 cation A
A A
- p, C
Certificate 11 older's Stress Report A
NA NA D
Nonmandatory Preheat Procedures A
A A
P E
Minimum Bolt Cross-Sectional Area A
NA NA F
Rules for Evaluation of Level D
.A A
A A
Service Limits G
Protection Agaicst Nonductile Failure A A
A A
H Capacity Conversions for Class 3 NA NA A
Safety Valves J
Owner's Design Specifications for A
NA NA O
Core Support Structure K
Recommended Maximum Deviations and A
A A
O Tolerances for Component Supports
.I
s g>
Ybq h Table A2-2 gg Comparison of B31.1 (1955) [4] Against ASME Section III (1977) [1]
R3N Article
- Corresponding
?N or Class Class Class Signi-Article in i
f=$
Sutu r t icle Description 1
2 3
ficant B31.1 (1955)
Remarks lh NA-1000 SCOPE OF SECTION III A
A A
38 NA-2000 CLASSIFICATION OF COMPONENTS A
A A
A Not Addressed NA-3000 RESPONSIBILITIES AND DUTIES A
A A
NA-4000 QUALITY ASSURANCE NA-4100 Quality Assurance Requirements A
A NA A
Not Addressed NA-5000 INSPECTION f
NA-5100 General Requirements for Authorized A
A A
A Not Addressed
[
Inspection Agencics and Inspectors NA-5200 Duties of Inspectors A
A A
A Not Addressed NA-6000 QUALITY CONTROL SYSTEMS FOR CLASS 3 CONSTRUCTION NA-6100 General Requirements HA NA A
A Not Addressed NA-6200 Organization and Responsibilities NA NA A
A Not Addressed NA-6300 Control of Operations NA NA A
A Not Addressed NA-6400 Records and Forms NA NA A
A Not Addressed NA-8000 CERTIFICATES OF AUTIlORIZATION, A
A A
NAMEPLATES, STAMPING, AND REPORTS A Addressed in the Code for the specified class or considered significant for this review.
- Not considered significant for this review.
O Outside the scope of this review.
HA Not applicable to this review or not addressed in the Code for the specifled class.
Article number in current Code will be preceded by N8 for Class 1 component, NC for Class 2 component, and ND for Class 3 component.
t
M, C 73 Table A2-2 (Cont.)
Article
- Corresponding ym or Class Class Class Signi-Article in
- 3 Subarticle IAascription 1
2 3
ficant B31.1 (1955)
Remarks
,S h
1000 INTRODUCTION g'n 1100 Scope A
A A
A 101, Table 22, 5j Note 2 4
2000 MATERIAL 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 Requirement A
A A
A Not Addressed for Material 2400 Welding and Brazing A
A A
A Sect. 6: Chapter h
4 and Appendices W
2500 Examination and Fepair of Pressure A
A A
Retaining Materials 2600 Material Manufacturers' Quality A
A A
A Not Addressed System Program 2700 Dimensional Standard 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 Vessels (C1. 2) 3300 Vessel Design A
A A
A NA 3400 Pump Design A
A A
A NA 3500 Valve Design A
A A
A 107,108,124, 129,134,139 3600 Piping Design A
A A
A Sect. 1 3700 Electrical and Mechanical Penetration NA A
A A
NA Assemblies 3800 Design of Atmospheric Storage Tanks NA A
A A
NA 3900 0-15 pai (0-103 kPa) Storage Tank NA A
A A
NA Design
.t
lC UP Table A2-2 (Cont.)
kE Atticle*
Corresponding or Class Class Class Signi-Article in J
Subarticle Description 1
2 3
ficant B31.1 (1955)
Remarks 50
$h 4000 FABRICATION AND INSTALLATION Sect. 6 Jn 4100 General A
A A
Not Addressed
{@
4200 Forming, Fitting, and Aligning A
A A
4300 Welding Qualifications A
A A
Appendix A to Sect. 6 4400 Rules Governing Naking, Examining, A
A A
and Repairing Welds 1500 Brazing A
A A
4600 lieat Treatment A
A A
4700 Hechanical Joints A
A A
Chapter 2 of Sect. 6
- p 4800 Expansion Joints NA A
A Not Addressed i
5 5000 EXAMINATION 5100 General Requirements A
A A
A Not Addressed
$200 Required Examination of Welds A
A A
A Not Addressed (C1. 1)3 Examination of Welds (C1. 2 and C1. 3) 5300 Acceptance Standard A
A A
A Not Addressed 5400 Final Examination of Items (C1.1);
A NA A
A Not Addressed Spot Examination of Wulded Joints ici. 3) 5500 Qualifications of Nondestructive A
A A
A Not Addressed Examination Personnel 5600 NA NA NA 5700 Examination Requirements of NA A
A Expansion Joints 6000 TESTING 6100 General A
A A
6200 Ilydrostatic A
A A
6300 Pneumatic A
A A
O
dl 5i lI I
'J 1-~ ' s Taole A2-2 (Cont.)
" '1 /
Article
- Corresponding gy or Class Class Class Signi-Atticle in i
Subarticle Description 1
2 3
ficant B31.1 (1955)
Remarks po O
f*g 6400 Pressure Test Gages A
A A
2O 6500 Atmospheric and 0-15 peig NA A
A f
ln Storage Tanks 6600 Ilydrostatic Testing of Vessels NA A
NA Q
Designed to NC-3200 6700 Pneumatic Testing of vessels NA A
NA Designed to NC-3200 6800 6900 Proof Tests to Establish NA A
A NA Design Pressure 7000 PROTECTION AGAINST OVERPRESSURE 7100 General A
A A
NA h
7200 Definitions Applicable to A
A A
NA Ovespressure Protection Devices 7300 Overpressure Protection Report A
A NA NA (C1. 1): Analysis (C1. 2) 7400 Relievir>J Capacity Requirements A
A A
NA and Acceptable Types of Overpressure Protection Devices 7500 Set Pressures of Pressure Relief A
A A
7' Devices 7600 Operating Design Requirements for A
A A
NA Pressure Relief Valves 7700 Requirements for Nonreclosing A
A A
HA Pressure Relief Devices 7800 Certification Requirements A
A A
NA 7900 Marking, Stamping, and Reports A
A A
NA 8000 NAMEPI).TES, STAMPING, AND RE. ORTS 8100 General A
A A
t
Table A2-2 (Cont. )
<= /
17 Article
- Em Corresponding or Class Class Class Signi-Article in lh subarticle Description 1
2 3
ficant B31.1 (1955)
Remarks s:2N MANDATORY APPENDICES n
f$
I Design Stress Intensity Values, A
A A
A Tables 1 and 2, s#
Allowable Stresses, Material Sect. 1 fh Properties, and Design Fatigue Curves II Experimental Stress Analysis A
A A
III Basis for Establishing Design A
A A
A Not Addressed Stress Intensity Values and Allowable Stress Values IV Approval of New Materials Under A
A A
the ASME Boller and Pressure Vessel Code for Section III Application V
Certificate Holder's Data Report A
A A
p e
Forms and Application Forms for Certificates of Authorization for Use of Code Symbol Stamps VI Rounded Indications Charts A
A A
VII Charts for Determining Shell A
A A
122 Thickness of Cylindrical and Spherical Components Under External Pressure XI Hules for Bolted Flange NA A
A 106,111,138, Connections for Class 2 and 3 143 Components and Class MC Vessels XII Design Considerations for Bolted A
A A
A Flange Connections XIII Design Based on Stress Analysis HA A
NA Not Addressed for Vessels Designed in Accordance with NC-3200 XIV Design Based on Fatigue Analysis NA A
NA NA i
for Vessels Designed in Accordance with NC-3200 1
4
~
I l
If, c=
7R )t/
!E as
}k 30 Table A2-2 (Cont.)
y IS
=S Article
- Corresponding
- 7
/
or Class Class Class Signi-Article in g$
5 Subarticle De script ion 1
2 3
ficant B31.1 (1955)
Remarks 2
XVI Nondestructive Exasination A
A A
O NA Nethods Applicable to Core i
Support Structures XVII Design of Linear Type Supports by A
A A
O NA Linear Elastic and Plastic Analysis NONNANDATORY APPENDICES A
A NA NA h
B Owner's Design Specification A
A A
C Certificate Holaer's Stress Report A
NA NA Nonnandatory Pr eheat Procedures A
A A
a E
Minimum Bolt Cross-Sectional Area A
NA NA F
Hules for Evaluation o.' Level D A
A A
A l
Service Limits l
G Protection Against Nonductile Failure A
A A
A H
Capacity Conversions for Class 3 NA NA A
Safety Valves J
Owner's Design Specifications for A
NA NA O
NA Core Support Structure K
Hecommended Naximum Deviations and A
A A
O NA Tolerances for Component Supports I
Table A2-3 Comparison of ASME VIII (1962) [3] with ASNE !!! (1977) Ill Article *
[
Corresponding or Class Class Class Signi-Article in a 5-Subarticle Description 1
2 3
ficant ASME VIII (1962)
Remarks S
q NA-1000 SCOPE OF SBCTION III t.
A A
A
$h NA-2000 In CLASSIFICATION OF COMPONENTS A
A A
A NA 5k NA-3000 RESPONSIBILITIES AND DUTIES 3
A A
A NA-4000 QUALITY ASSURANCE NA-4100 Quality Assurance Requirements A
A NA A
NA NA-5000 INSPECTION NA-5100 General Requirements for Authorized A
A A
A UG-90 Inspection Agencies and Inspectors NA-5 200 Duties of Inspectors A
A A
A UG-90 (c) s UG-93 b
NA-6000 QUALITY CONTROL SYSTEMS FOR CLASS 3 CONSTRUCTION NA-6100 General Requirements NA NA A
A NA NA-6200 Organization and Responsibilities NA NA A
A NA NA-6300 Control of Operations NA NA A
A NA NA-6400 Records and Forms NA NA A
A NA NA-8000 CERTIFICATES OF AUTHORIZATION, A
A A
UG-ll5 through NAMEPLATES, STAMFING, AND REPORTS UG-120 1000 INTRODUCTION 1100 Scope A
A A
A U-l A
Addressed in the Code for the specified 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 addressed in the Code for the specified class Article number in current Code will be preceded t,y NB for Class 1 component, NC for Class 2 co 1
Class 3 component.
mponent, and ND for
}
1 4
Table A2-3 (Cont.)
e=>
<=>
p Article
- lg or Corresponding Class Class Class Signi-Article in g Ei
_Subarticle Description 1
2 3
ficant ASME VIII (1962)
Remarks Fa?
yg 2000
{S R
2100 Ge neral A
A A
A DG-5 2200 Material Test Coupons and Specimens A
A A
gl}
for Ferritic Steel Materials 3
2200 Fracture Tbughness Requirement A
A A
A UG-84 3
for Material 2400 Welding and Brazing A
A A
A UW E US l
2500 Examination and Repair of Pressure A
A A
}
Retaining Materials 2600 Material Manufacturers' Quality A
A A
A UG-93 i
System Program 2703 Dimensional Standard A
A A
T 3006 DESIGN rd 3100 General A
A A
A NA 3200 Design by Analysis (C1. 1): Alternate A A
NA A
Design Rules for Vessels (C1. 2) 3300 Vessel Design A
A A
A UW-8, UF-12, UG-16 through UG-55 3400 Pump Design A
A A
A NA 3500 Valve Design A
A A
A NA 3600 Piping Design A
A A
A NA 3700 Electrical and Mechanical Penetration NA A
A A
NA Assemblies 3800 Design of Atmospheric Storage Tanks NA A
A A
NA 3900 0-15 ps! (0-103 kPa) Storage Tank NA A
A A
NA De sign t
4000 FABRICATION AND INSTALLATION
{
4100 General A
A A
A U3-75 l
4200 Forming, Fitting, and Aligning A
A A
I 4300 Welding Qualifications A
A A
A UW-28, UW-29 4400 Rules Governing Making, Examining, A
A A
and Repairing Welds 4500 Brazing A
A A
4600 Heat Treatment A
A A
l i
I%
Lf' Table A2-3 (Cont.)
74
$E Article
- Conresponding or Class Class Class Signi-Article in 3
Subarticle Description 1
2 3
ficant ASME VIII (1962)
Remarks e
p g3 4700 Hechanical Joints A
A A
UR-19 4800 Expansion Joints NA A
A NA la e
5000 EXAMINATION j
5100 General Requirements A
A A
A UG-90 5200 Required Examination of Welds A
A A
A UW-46 (C1. 1): Examination of Welds (C1. 2 and C1. 3) 5300 Acceptance Standard A
A A
A UW-51 (m) 5400 Final Examination of Items A
NA A
A UG-99 (g), UW-5 0 UG-99 (g) requires (C1. 11 : Spot Examination of inspection after
- p Welded Joints (C1. 3) hydrostatic bu2 C es L
not apecify liquid o
penetrant or magnetic particle inspections UW-50 requires LPE or magnetic particlo inspection before pneumatic testing.
UG-91 gives requirements for qualification of inspectors, but not NDE personnel 5500 Qualifications of Hondestructive A
A A
A NA Examination Personnel 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
++3 c
g'h Table A2-3 (Cont.)
er
':tg Article
- Corresponding or Class Class Class Signi-Article in In@
Subarticle Description 1
2 3
ficant ASME VIII 11962)
Remarks E*
~ ?.4 6400 Pressure Test Gages A
A A
6500 Atmospheric and 0-15 psig NA A
A Storage Tanks 6600 Hydrostatic Testing of Vessels HA A
NA Designed to NC-3200 6800 Pneumatic Testing of Vessels NA A
NA Designed to NC-3200 6800 6900 Proof Tests to Establish NA A
A A
UG-101 b
Design Pressure w
7000 PROTECTION AGAINST OVERPRESSUR8 7100 General A
A A
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 pequirements A
A A
and Acceptable Types of Overpressure Protection Devices 7500 Set Pressures of Pressure Relief A
A A
Devices 7600 Operating Design Requirements for A
A A
Pressure Relief Valves 7700 Requirements for Nonreclosing A
A A
Pressure Relief Devices 7800 Certification Requirements A
A A
7900 Marking, Stamping, and Reports A
A A
8000 NAMEPLATES, STAMPlNG, AND REPORTS 8100 General A
A A
- t
lli
<a fa Table A2-3 (Cont.)
5E to g 30
,Q Article
- e Corresponding j$
or Class class Class Signi-Article in 3h Subarticle Description 1
2 3
ficant ASME VIII (1962)
Remarks f(1 g$
MANDATORY APPENDICES I
tesign Stress Intensity Values, A
A A
A Subsection C Fatigue Curves Allowable Stresses, Material not included in Properties, and Design Fatigue Sect. VIII Curves II Experimental Stress Analysis A
A A
III Basis for Establishing Design A
A A
A Appendices P&Q Stress Intensity Values and p
Allowable Stress values
[
IV Approval of New Naterials Under A
A A
N the ASME Boiler and Pressure Vessel Code for Section III Application V
Certificate' Holder's Data Report A
A A
Forms and Application Forms for Certificates of Authorization for Use of Code Gymbol Stamps VI Rounded Indications Charts A
A A
VII Charts for Determining Shell A
A A
UG-28 & Appendix V Thickness of Cylindrical and Spherical Components Under External Pressure XI Rules for Bolted Flange NA A
A Appendix II Connections for Class 2 and 3 Components and Class HC Vessels XII Design Considerations for Bolted A
A A
A NA Flange Connections XIII Design Dased on Stress Analysis NA A
NA for Vessels Designed in Accordance with NC-3200 XIV Design Based on Fatigue Analysis NA A
NA for Vessels Designed in Accordance with NC-3200
^3
%n]
Is st 3 N
'Ify Table A2-3 (Cont.)
'S[n 3$
Article
- Corresponding 4
or Class Class Class Signi-Article in Subarticle Description 1
2 3
ficant ASNE VIII (1962)
Remarks XVI Nondestructive Examination A
A A
O Nthods Applicable to Core Support Structures XVII Design of Linear Type Supports by A
A A
O Linear Elastic and Plastic Analysis N0tatANDA10RY APPENDICES A
A NA NA B
Owner's Design Specification A
A A
C Certificate Holder's Stress Report A
PA NA I
D Nonmandatory Preheat Procedures A
i A
E Minimum Bolt Cross-Section Area A
NA NA P
Rules for Evaluation of Level D A
A A
A NA Service Limits l
G Protection Against Nonductile Pailure A A
A A
NA H
Capacity Conversions for Class 3 NA NA A
l Safety Valves
{
J Owner's Design Specifications for A
NA NA O
t Core Support Structure i
K Recommended Maximum Deviations and A
A A
O Tolerances for Component Supports l
l t
i
3.
CONCLUSIONS AND RECOMMENDATIONS j
Nuclear components and systems for SEP " Group II" plants may have been designed in accordance with the following codes:(1) 1.
ASME I (1962) - piping and valves 2.
ASME VIII (1962) - vessels 3.
B31.1 (1955) - piping and valves 4.
TEMA (1959). - heat exchangers 5.
ASA B16.5 (1961) - steel pipe flanges and flanged fittings 6.
Hydraulic Institute Standards (1965) - pumps 7.
USAS B96.1 (1967) - aluminum field erected storage tanks 8.
API 650 (1964) - welded steel tanks for oil storage 9.
B16.10 (1957) - valves and B16.9 (1958) - fittings 10.
USAS B31.1 (1967), AbEI B31.1 (1973), USAS B31.7 (1968 Draf t),
Draf t ASME Code for Pumps and Valves for Nuclear Power (1968).
Current requirements are contained in the following:
11.
ASME III (1977) - Div. 1 nuclear components 12.
ANSI B16.34 (1977) - steel valves.
The following broad conclusions can be made regarding components built to past codes and evaluated against current requirements:
1.
Components currently classified as Class 3 would satisfy basic current requirements, except for full radiography requirements for welded vessel joints less than 1-1/2 in thick for some materials as noted in Section 4.3 of this appendix.
2.
Components currently classified as Class 1 or Class 2 may not satisfy current fracture toughness and full radiography requirements.
3.
Piping currently classified as Class 1 satisfies current requirements except possibly high cycle fatigue, fracture toughness, and full radiography requirements.
Piping currently classified as Class 2 may not satisfy current fracture toughness and full radiography requirements.
4.
Vessels and pumps currently classified as Class 1 may not satisfy current fatigue analysis requirements.
5.
Vessels pre -tously classified as " primary vessels" by Code Case 1270N would curre.ntly be categorized as Class 1 vessels. Vessels previously classified as " secondary vessels" by Code Case 1270N may currently be regarded as Class 2 or Class 3 vessels.
1 1.
Modified for nuclear components by code cases for vessels and piping when invoked.
A-24 dd Franklin Research Center A Omsnm af The Frannha insoMe
The following is recommended:
1.
Component materials should be evaluated for fracture toughness as described in Section 4.1.1 of this appendix.
2.
Standard class rated valves should be carefully checked against
~l current pressure-temperature ratings.
3.
Atmospheric and 0 to 15 psig storage tanks should be carefully reviewed against current requirements.
4.
Unless Code Case N-1 together with either N-2, N-7, N-9, or N-10 has been invoked when designing to B31.1 requirements, Class 1 and 2 piping should be checked to see if full radiography of welded joints was specified.
5.
Past full radiography past requirements for Class 2 and Class 3 welded vessel joints less than 1 1/2 in thick should be reviewed in light of Section 4.3 of this appendix.
6.
Currently classified Class 1 vessels and pumps should be evaluated for fatigue analysis requirements.
t l
A-25 nklin Research Center
4.
COMPARISON OF SIGNIFICANT CURRENT CODE REQUIREMENTS AND PAST REQUIREMENTS 4.1 GENERAL REQUIREMENTS Section 4.1 compares the significant general requirements of the current
~
code [1] with past requiremen.s.
In addition, where feasible, an approach is formulated which facilitates the review of nuclear components and systema 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 Requirements Class 1 Components The current code requires that pressure-retaining materials for Class 1 components shall be impact tested to determine T
- by the Charpy 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, pumps, and valves with nominal pipe size 6 in or less; austenitic stainless steels; and non-ferrous materials.
Drop weight tests are not required for martensitic high alloy chromium (Series 4xx) ano precipitation-hardening steels listed in Appendix I (le); however, other requirements of NB-2332 (lb] do apply.
Class 2 Components Pressure-retaining materials for Class 2 components are required to be impact tested with exceptions as outlined for Class 1 components. Also exempted are commonly used plate, forging, and casting materials listed in Table NC-2311(a)-1 of Reference lc when used in Class 2 components whose lowest service temperature (LST)
- exceeds the tabulated nil ductility transition temperature (TNDT) ya east the thickness-dependent value A,
- See Table A4-1 for definitions of commonly used terms and symbols.
5 A-26 4'h AMnklin Research. Center s o--. a n. n
Table A4-1 Definition of Commonly Used Fracture Toughness Terms and Symbols Symbol Definition TNDT A temperature at or above the nil ductility temperavire 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 l
initiated only after 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 end cv tested in accordance with SA-370 Charpy V-Notch testing exhibit at least 35 mils lateral expansion and not less than 50 ft-lb absorbed energy.
l 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.
l l
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=J Franklin Research Center m.onorrh.r.n.u.o m u,
determined from the curve in Figure NC-2311(a)-1 from Referencs lc.
For convenience, the table and the figure are reproduced as Trble 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 recuired 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 (1c]
applies for these steels in all thicknesses. For nominal wall thicknesses greater daan 2.5 in, the required C values shall be 40 mils lateral y
expansion.
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.
The evaluation of materials based on past codes for which fracture tougnness 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 components 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.
g A-28 Franklin Research Center A Omton of The Freeman meatute
l l
l Table A4-2 TABLE NC-2311(a)-1 EXEMPTIONS FROM IMPACT TESTING UNDER NC-2311(a)-8 Material Material' Condition 8 T.or,8 deg. F SA-537-Class 1 N
-30 SA-516-Grade 70 '
Q&T
-10 SA-516-Grade 70 N
0 SA-508-C! ass 1 Q&T
+10 SA-533-Grade B Q&T
+10 SA-299*
N
+20 SA-216, Grades Q&T
+30 WCB,WCC SA-36 (Plate)
+40 SA-508-Class 2 Q&T M0 NOTES:
(1) These materials are exempt from toughr.ess testing when A or LST T.oris above the curve in Fig. N C-2311(a)-1, for the thickness as defined in NC-2331 or NC-2332.
(2) Material Condition letters refer to:
N - Normalize Q & T - Quench and Temper HR - Hot Rolled (3) These values for T., were established from data on heavy section steet (thickness greater than 2% in.). Values for sec*Jons less than 2% in. thick are held constant until addiJonal data is obtained.
(4) Materials made to a fine grain melting practice.
A-29 L.bu Franklin Research Center ao - em. n e. mum
th.
e.>
dv"{.
11 aw
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- k 120 1
8Q b ?.
Permisiltale Lowest Survice Tempereture - Typ 7 + A rs' 100 66 80
~
8
/
r y
40
/
30 20 0
6/8 2
4 6
8 10 12 14 Thickness. In.
FIG. NC-2311(a)-1 DETERMINATION OF PERMISSIBLE LOWEST SERVICE TEMPERATURE Figure A4-1 I
Table A4-3 TABLE ND.23111 EXEMP1 IONS FROM IMPACT TESTING UNDER ND.2311(a)(8)
Lowest Service Temperature for the Thickness Shown Over % in.
Over % in.
Over 1 in.
Over 1% in.
to % in., ind.
to 1 in., incl to 1% in., inct to 2% in., inct Material (Over 16 mm (Over 19 mm (Over 25 mm (Over 38 mm Material Condition' to 19 mm, incl) to 25 mm, incl) to 38 mm, incl) to 64 mm, inct)
SA.516 Grade 70 N
-30 F (-34 C)
-20 F (-29 C) 0 F (-18 C) 0 F (-18 C)
SA.537 C! ass 1 N
-40 F (-40 C)
-30 F (-34 C)
-30 F (-34 C)
-30 F (-34 C)
SA 516 Grade 70 Q&T (2)
(2)
(2)
-10 F (-23 C)
SA.508 Class 1
.Q&T (2)
(2)
(2) 10 F (-12 C)
SA 508 Class 2 Q&T (2)
(2)
(2) 40 F (4 C)
SA.533 Grace B' Q&T (2)
(2)
(2) 10 F (-12 C)
C! ass 1 SA 216 Grades Q&T (2)
(2)
(2) 30 F (-1 C)
N (2)
(2)
(2) 20 F (-7 C)
NOTES:
(1) Material Condition letters ref er to:
N - Normalize Q & T - Quench and Temper (2) The lowest service temperature shown in the column "Over 1% in to 2% in." may he used for these thicknesses.
(3) Material made to a fine gram melting practice.
A-31 g
JZU Frankiin Research Center A Drnesan of The Frannan inschAe
1 Table A4-4 Evaluation for Fracture Toughness of Pressure-Retaining Material for Class 1 Component / System Nuclear Power Plant I
FSAR Page I.
Component / System Data 1.
Description of Component / System:
2.
Material Description and Thickness: P No.
3.
Design Temperature:
'F 4.
Design Pressure:
psi 5.
Lowest Service Temperature (
(LST) :
'F 6.
Pressure at LST:
psi 7.
Fracture Toughness Requirement? Yes No II.
Evaluation 8.
Material is exempt ( 'I from impact testing because:
(a)
Nominal thicxness 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, 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 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), is likewise exempt from impact testing. However, exemption 9 does not exempt either the weld metal (NB-2430) or the welding procedure qualification (NB-4335) from impact testing. See paragraph NB-2431 of Reference Ib.
3.
The current code does not exempt Class 1 ccmponents from impact testing on the basis of tabulated TNDT and A values as it does Class 2 components.
Item 9 is not an exempcion listed in paragraph NB-2311 but a conservative adaptation of NC-2311(a) (8) for Class 2 components to facilitate the SEP review.
A-32 De UM Franklin Research Center A Densson of The Frannan inseaste
l l
i l
Table A4-4 (Cont. )
(e)
Austenitic stainless steel (f)
Non-ferrous material I
9.
Fracture toughness of material appears does not eopear to be adequ' ate on the basis of the following evaluation:
(a) for material other than bolting and up to 2-1/2 in thick J
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 (b) for material other than bolting in excess of 2-1/2 in thick:
RTNDT =
- F (Ref erence used(4) :
)
- and,
- F which exceeds 120*F (LST - RTNDT)
=
which does not exceed 120*F 10.
For bolting material in excess of 1-in diameter, reference da ta (4) has been available has not been available and found to satisfy not satisfy the requirements of NB-2333 (4(b)]
11.
Fracture toughness cannot be evaluated because of insufficient information.
12.
Material is not exempt from impact testing.
NOTE:
4.
When using references other than the current code to obtain TNDT and RTNDT, be sure that the data have been obtained from specimens whose condition matches the material being evaluated (e.g., normalized or quenched and tempered) and that have designation such as "SA-516 Gr. 70".
nklin Rese
~ ~_ arch _ Center
l l
i l
l l
Table A4-4 (Cont. )
III.
Conclusions Fracture toughness appears to be adequate.
Adequacy of fracture toughness not established; request supplemental test data and supporting documents.
~
l Welding material is is not the basis of foregoing data and Note 2. exempt from impact' testing on l
l l
l
\\
l 1
l l
i A
A-34 JU'1Enklin Research Center a % w w Fran mau.
~.
Table A4-5 Evaluation for Fracture Touanness of Pressure-Retaining Material for Class 2 Component / System Nuclear Power Plant I
FSAR Page I.
Component / System Data 1.
Description of Component / System:
2.
Material Description and Thickness: P No.
3.
Design Temperature
'F-4.
Design Pressure:
' psi 5.
Lowest Service Temperature I I (LST):
'F 6.
Pressure at LST:
psi 7.
Fracture Toughness Requirement? Yes No II.
Evaluation 8.
Material is exempt ( ' from impact testing because:
1 (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, pumps, and valves, nominal pipe size of 6-in diameter or smaller (e)
Austenitic stainless steel (f) Non-ferrous material liUTES:
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. However, 8 (g) exemption does not exempt either the weld metal (NC-2430) or the weld procedure qualification (NC-4335) fecm impact testing. See paragraph NC-2431 of Reference Ic.
A-35 000 nklin Research Center A Dnem of The Frannhn ipseue
Table A4-5 (Cont.)
(g)
LST of material listed in Table NC-2311(a)-1 (see Table A 4-2) exceeds TNDT by at least "A" (A depends on thickness). (2)
- F (from FSAR)
I 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 insufficient information.
10.
Material is not exempt from impact testing.
III.
Conclusions 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-36 O
UUb!.Enklin Research Center A Deus.on of The Frannan insonde
Table A4-6 Evaluation for Fracture Toughness of Pressure-Retaining Material for Class 3 Component / System Nuclear Power Plant FSAR Page I.
Component / System Data 1.
Description of Component / System:
2.
Material Description and Thickness: P No.
3.
Design Temperature:
"F 4.
Design Pressure psi 5.
Lowest Service Temperature (LST) :
'F 6.
Pressure at LST:
poi 7.
Fracture Toughness Requirement? Yes No II.
Evaluation 8.
Ma ter'
.xempt( I 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)
Pipes, fittings, pumps, and valves, nominal pipe size of 6-in diameter or smaller (e)
Austenitic stainless steel (f)
Non-ferrous 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].
p 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.
However, exemption 8(g) does not exempt either the weld metal (NC-2430) or the weld procedure qualification (NC-4335) from impact testing. See paragraph NC-2431 of Reference'ld.
A-37 d Franklin Research Center 4 busson of The Frannhn insatute
f,2 i
Table A4-6. (Cont.)
(g)
LST equals or exceeds TNDP in Table NC-2311(a)-1 for the material and thickness being evaluated. (2)
(h)
LST exceeds 100'F.
9.
Fracture toughness cannot be et aluated because of insufficient information.
.- 10. Material is not exempt from impact testing.
III. Conclusionb Fracture toughness appears to be adequate.
Adequacy of fracture toughnese 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
A-38 Udd Franklin Research Center A Onas on of The Frenda maaeme
~
From Figure A4-1, A = 48' for material _3.75 in thicks (LST - TNDT) = 13 0 ' - O ' = 13 0 'F > 4 8 'F = A so that this material, if it were a Class 2 or 3 component, would be exempt from Lapact 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 suf ficient to require the fracture mechanics 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 examples (LST - TNDT) = 130*F > 100'F so that the material for the Palisades primary coolant piping is considered adequate.
4.1.2 Quality Assurance Requirements (1)
The current code [1] requires that activities in connection with the 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 I
accordance with Article NCA-4000 for Class 1 and 2 items, and in accordance wi th NCA-4135 ( I and NCA-8122 I ' for Class 3 items. The 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.
1.
Quality assurance requirements have been determined to be outside the scope of SEP Topic III-l according to the letter from S. Bajwa to S. Carfagno 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 Summer 1978 Addenda to ASME III (1977) General Requirements.
A-39 g
EUj FranMin Research Center w aonevmonm.au, c
. _ -. = _..-
4 1
J For 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
. an organization with sufficient authority, freedom, and independence a.
from cost and schedule considerations to:
1.
identify quality problems t
I 2.
initiate, recommend, or provide solutions 3.
verify implementation of solutions 4.
limit and control further work on nonconforming items 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 notification of the authorized inspection agency of significant c.
changes in the program d.
control of the design to assure compliance with the design specification of Section III i
design review and chec41ng by individuals or groups other than those e.
who performed the original design f.
documentation for procurement of materials and subcontracted services requiring compliance with Section III g.
document control with provisions for review of changes i
i h.
identification and traceability of materials 1
l i.
the control of construction processes
- j. examination, testing, and inspections verifying the quality of work by persons independent from supervisors i.amediately responsible for the work being inspected, and using measuring and test equipment j
calibrated against measure ent standards traceable to national standards (where such standards exist) at intervals sufficient to maintain accuracy within necessary limits nklin Research Center A On.esan of N 'rensen aramee
k.
proper handling, storage, shipping, and preservation of materials and component.s 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 m.
prompt identification and corrective action of significant conditions adverse to quality, with documented measures to preclude repetition maintenance of quality assurance records as specified in NCA-4134.17 n.
of Reference 1, including maintaining for the life of the plant as a minimum, the followings a permanent record file, certified design and construction specifications, drawings and repcrts, data reports, certified stress reports, certified as built drawings, material test reports, non-destructive examination reports, and test treatment reports a comprehensive system of planned and periodic audits with o.
documentation of results, follow-up action, and re-audit of deficient areas.
Class 3 items are to be designed and consttucted in accordance with the quality control requirements of NCA-4135 of Reference 1, which includes an organization chart which reflects the actual organization a.
b.
a quality control system suitable to the complexity of the work and size of the organization per sons who perform quality control functions with suf ficient c.
responsiblity, authority, and independence to implement the quality control system, identify problems, and initiate, recommend, and provide 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 Nacional Board of Boiler and Pressure Vessel Inspectors or a representative of the facility will participate in the evaluation.
If jurisdictional laws do not require inspection or permit inspection personnel to participate in the evaluation of the quality control system, then i
I A-41 g
C Franklin Research Center ao.m.oaw78.ne nr w.
,?
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 I
system was not required by past codes, many quality assurance features were requirea.
Although the program or system was not specifically required, neverthe-less, construction organizations typically did operate under "in-house" quality assurance programs which provided for the inspection, testing, and surveillance of ccaponents and construction activities.
Design 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.
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 recommended:
- Appendix X to the " Reactor Safety Study - An Assessment of Accident Risks in U.S. Connercial Nuclear Power Plants," WASH-1400, USAEC, Draf t August 1974.
A-42 M* fs*",2F,:Tghgnw w
1.
a design review of the deficient area with design change recommenuations 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 class ified as Class 1, 2, 3, MC, or CS.
Class MC*and CS are for metal containment vessels and core support structures 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)
Components 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 systems, post-accident heat removal systems, post-accident fission product removal b.
reactor shutdown or residual heat removal systems c.
EWR main steam components described in Reference 2:
o main steam line from second isolation valve to turbine stop valve o main steam line branch lines to first valve o main turbine bypass line to bypass valve o first valve in branch lines connected to either main steam lines or turbine bypass lines d.
PWR steam generator steam and feedwater systems up to and including outermost containment isolation valves and connected A-43 UUhb nklin Research Center
% a Th. rr= = a m.
piping up to and including the first valve that is normally closed or capable of automatic closure during normal reactor i
operation e.
systems connected to the reactor coolant pressure boundary not capable of being isolated from 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 cesignated Class 2, but are part of:
1.
cooling water and auxiliary feedwater systems important to safety, such as emergency core cooling or post-accident heat removal 2.
cooling water and seal water systems that are designed for functioning of components important to safety, such as cooling water systems for reactor coolant pumps, diesels, and control room 3.
systems connected to the reactor coolant pressure boundary that are
- apable of being isolated from tne boundary by two valves normally closed or capable of automatic closure during normal operation 4.
systems not previously defined, other than radioactive waste management systems that contain or may contain radioactive material, and whose postulated failure would potentially result in of f-site doses that exceed 0.5 rem.
Comparison with Past Codes The past B31.1 (1955) piping code and ASME I (1962) do not designate quality classes for piping or valves. Comparison of the component classifica-tion designations in the FSAR with the standards previously described for each SEP plant is required before a comparison with current code requirements can be initiated.
_nklin Rese_ arch._ Center
Piping built to ASME I (1962 Edition) and for which Code Class -1270N was invoked would currently be designated as Class 1 piping.
Piping built to B31.1 (1955) for which Code Case N-1 was invoked would be classified in one of the two following categories:
- 1. Nuclear piping designed to contain a fluid "whose loss from the system could result in a radiation hazard either to the plant personnel or to the general public"
- 2. Conventional steam and service non-nuclear piping.
Nuclear piping can currently be designated as Class 1 or Class 2.
Code Cases N-7, N-9, and N-10, when invoked for nuclear piping, together with ASA B31.1 requirements, have been evaluated in Section 4.2 of this appendix to determine whether piping built to these requirements codes satisfies current Class 1 or Class 2 requirements.
Past ASME B&VP Code Sections I and VIII, 1962 editions, did not classify vessels for nuclear service. Code Case 1270N, when invoked, did classify vessels for nuclear service as follows:
a.
Reactor vessel (outside scope of this study) b.
Primary vessels, other than reactor and containment vessels designed to contain reactor coolant c.
Secondary vessels which do not contain reactor coolant or are otherwise subject to irradiation I
d.
Containment vessels (outside scope of this study)
Nete that a vessel previously designated as a primary vessel under Code Case 1270N would currently be designated as a Class 1 vessel. A vessel previously designated as a secondary vessel could be designated as a Class 2 or Class 3 vessel under current rules.
Secondary vessels currently designated as Class 2 or Class 3 should be carefully reviewed for possible non-compliance with current full radiography requirements.
A-45
.ef21 dUNnklin Research Center
1 i
4.1.4 Code Stress Limits Strength Theories Past codes [3, 4, 5), 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 commonly accepted that both the maximum shear y
stress theory (Tresca criterion) 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 Misea criterion is more accurate than the maximum shear stress tneory, the present code [1] uses the maximum shear stress theory of strength for Class 1 components because (1) it is more conservative, (2) it IN is easier to apply, and (3) it facilitates fatigue arslysis. Class 2 and Class 3 components continue to be designed in accordance with the maximum principal stress assumption.
If the principal stresses at a point are at > e2 > 03, then yielding occurs when:
Tmax = (1/2) (or-0) = (1/2) Sy 3
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.
Examplo:
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 o and an axial stress c/2. The three principal stresses in descending magnitude would be c1 = a c2 = (1/2)o c3 = -p According to the current code, the " stress intensity" is:
(c1 - 03) = (0 + p) 1.
Except for Class 2 vessels designed in accordance with the alternative rules of NC-3200.
A-46 nklin Researc
-_h Center
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 comprehensive and detailed determina-tion of stress and strain fields, in botn the elastic and plastic states due I
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:I I 1.
Primary Stresses Any normal or shear stress induced by an imposed load wnich is necessary to satisfy equilibrium oetween the external and internal forces and moments. A primary stress is not self-limiting. The existence of primary stresses in excess of the yield strength across tne thicxness 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 categorizea as:
a.
General Membrane Stress. The average primary stress across a solid section excluding the effects of gross and local discontinuities. The six stress components associated with a primary general membrane stress are symbolized by P
- m b.
Local Membrane 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 pressure. 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 bendingstress across the thickness of 1.
See Figure NB-3222-1 [lb).
A-47 d$b du Franklin Research Center w ao n. mn wu.
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
- b 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 failure due to secondary stresses induced by the application of one load will not occur.
Secondary stresses are further categorized as follows:
a.
Secondary Expansion Stresses.
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 Membrane and Bending Stresy. Occurring at gross structural discontinuities and caused by mechanical loads, pressure, or differential thermal expansion, symbolized by Q.
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.
Coce 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 components 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 I
(YS) at design temperature for ferritic steels. (1)
For austenitic steels, S is the lower of 1/3 (UTS) or 0.9 YS at design temperature or 2/3 (YS) at room temperature.I }
1.
See III-2110(a) of Reference le.
2.
See III-2110 (b) of Reference le.
A-48 ub Franidin Research Center A CWsson of The Frannan insrante
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 [lbi)
Stress Category Limit of Stress Intensity P rimary Stresses Tabulated YS UTS P,
S, i 2/3 (YS) i 1/3 (UTS)
P 1.5 S,
< YS
< 1/2 (UTS)
P
+P b
m 2.
Level A and B Service (Operating and Upset Conditions)
(See Fig. NB-3222-1 (lb])
Stress Catecory Limit of Stress Intensity (a) Expansion Stress Intensity Tabulated YS UTS P, (not for vessels) 3 Sm 12 YS 1 UTS (b) Primary and Secondary (1)
PL+Pb + Pe + Q 3S
$2 YS S 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-3222-4 (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(e) as follows:
a.
Determine the stress difference and the alternating stress intensity, Sa, for each condition of normal service.
1.
3S may be exceeded provided the conditions of NB-3228.3 are satisfied.
m 2.
For cyclic operation.
3.
2S for full range of fluctuation.
a A-49 branklin Research Center m aon m r.enm.au.
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 geometries 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 cycler Ni for a given alterna-ting strest value (Salt i. The alternati r stress determined l
from the analysis should be multiplied b che ratio of the modulus of elasticity given on the design fatigue Jarve divided by the modulus of elasticity used in the analysis before entering the design fatigue curve.
d.
Cumulative usage for multiple stress cycles is be determined from U = Sum of (M /N )
g where M is the expected number of cycles associated with (Salt)i and Ni s the corresponding number of cycles from the design fatigue curve. The cumulative usage factor U shall not exceed 1.
4.
Level C (Energency Conditions)
(See Fig. NB-3224-1 [lb])
Stress Category Limit Tyoe of Analysis Primary Stresses I1)
Pm (pressure and 1.2 S or YS m
Elastic mechanical)
Pm (pressure - only 1.1 S or 0.9 YS I1)
Elastic m
for ferritic material)
PL 1.8 S or 1.5 YS I1) m Elastic 0.8 (collapse load)
Limit PL+Pb 1.8 S or 1.5 YS I1)
Elastic m
0.8 (collapse load)
Limit 4.8 S Triaxial StressesI2) m Secondary / Peak Evaluation not required 1.
Whichever is greater.
2.
Based on sum of primary principal stresses.
A-50 Os Z Franklin Research Center m a m rr n,u m.
Bolting Material Stress Limits - Class 1 Components (NB-3230)
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 I l bolting material stress allowables given in Table I-1.3 of Reference le, which are based on the lower of:
L/3 (YS) at room temperature 1/3 (YS) at design temperature (up to 800*F).
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 )
j[ 2/3 (YS) b avg
- 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 )
ji (YS) b max Fatique Analysis of Bolts Fatigue analysis of bolts is required unless all the conditions of NB-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 A-51 AUU Franklin Research Center
% ae78.pr.nonmuu.
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.
For bolting having less than 100 kai tensile strength, use a fatigue c.
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.
Code Stress Limits Class 2 and Class 3 Components Design Allowable Stress Values Design allowable stress values are given in Table I-7.0 I ' for Class 2 and Class 3 and in Table I-8.0 for Class 3 component materials. These design 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 components 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).
2.
Austenitic Steel Non-Bolting Materials The stress allowable for austenitic steels is the lowest of:
1/4 (UTS at room temperature)
L/4 (UTS at temperature) 2/3 (YS at room temperature) 0.9 (YS at room temperature).
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., tne same as for Class 1 components.
nklin ReSe_ arch _ Center
3.
Bolting Materials Design stress allowables for bolting materials are based on the same criteria as for non-bolting materials, except that for heat-treated bolting materials, the allowable shall be the lower of:
1/5 (UTS at room temperature)
L/4 (YS at room temperature).
Level D (Faulted 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; tnermal 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 limi us on primary stress intensities for level D conditions depend on whether the system has been analyzed elastically or inelastically.
Elastic Syst.em 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 which 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 cesign pressure, in which case the primary stress computed by Equation 9 of NB-3652 should not exceeed 3Sm (2 x YS).
b.
Collapse 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 aaterial, or by experiment.
1.
A load which is in equilibrium with a system of stresses which satisfies i
l equilibrium everywhere, but nowhere exceeds the YS at or below the l
collapse load.
_nklin Rese_ arch C_ enter
c.
Stress Ratio Analysis: which is a pseudo-elastic analysis method utilizing the techniques and curves given in Appendix A-9000 [le), in which the apparent stress (2) is limited to the lesser of 3 S Of m
l 0.7 S except when the methods of A-9000 [le] 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:
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 Analysis:
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.
Stress Ratio Analysis: as described previously.
d.
Plastic I.: stability 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 PI or YS + (SI-YS)/3 where S is the true effective stress associated with plastic I
instability, e.
Strain Limit Load Analvsis:
in which the load P is limited as described in (d) but not to exceed Ps associated with a specified strain limit.
f.
Inelastic Analysis:
in whien primary stress is limitad as in (a).
Comoarison 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.
The B31.1 (1955) piping code and the ASME Boiler and Pressure Vessel Code, Sections I and VIII (1962) are based on maximum normal stress theory of failure.
2.
The current code for Class 2 and 3 items is based on the same theory of failure as past codes.
2.
Computed value of stress assuming elastic behavior.
A-54 b Franklin Research Center A Onneson of The Fransen insatuse
y 3.
The current code for Class 1 components considers primary as well _as secondary stresses and peak stress categories. The B31.1 (1955) power piping and ASME I (1962) power boiler codes do not consider peak stresses.
4.
Tne current code for Class 2 and 3 vessels considers primary stresses
~
for size selection, as does ASME VIII (1962). (1)
The current code for Class 2 and 3 piping considers primary and secondary stresses, as does the past B31.1 (1955) piping code.
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 evaluated.
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 determination of delta ferrite shall be performed for A-No. 8 weld material (see WW-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 l
typically be used to join chrome-nickel austenitic stainless steels such as i
SA-312 Grade TP 316. The minimum acceptable delta ferrite shall be SFN and results 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.
1.
Unless the vessel is designed in accordance with the alternative NB-3200 rules which are based on primary, secondary, and peak stresses.
A-55 N Franklin Research Center l
4c e n.rr.n m
I
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.
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 Design Considerations for Bolted Flange Connections Appendix XII of the current code (li provides supplementary information 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 ana?ysis of the joint which considers changes in bolt elongation, flange deformation, and gasket load that can take place upon pressurization and that may indicate a required bolt preload greater than 1.5 times the design value. This practice is permitted provided that excessive flange distortion and grosa crushing of the gasket is prevented. Bolt relaxation under high temperatures should also be investigated.
Me thods 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 current considerations of Appendix XII may be useful in evaluating problem connections.
_,gStz, A-56 Jij Franklin Research Center a om.on or m rr.n.u,n em.
~. _
4.2 PIPING The current Class 1 piping design requirements are given in N3-3600 of Reference Ib.
The fundamental differences between current and past require-ments are that:
1.
The currr;t code explicitly considers and evaluates the margin against fatigue damage by a formulation for peak stress which accounts for local as well as gross dis-continuities. The secondary stress indices C in the g
current code are equivalent in principal to the screes intensification factor i of the past code (4]. The current code magnifies gross discontinuity stress by g) multiplying C by a local stress index K.
The past code g
1
' considers the effect of cyclic loading by reducing the allowable 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 shews a plot of the allowable expansios stress based on the past code and labelled past " fatigue" curve super-i= posed against the design fatigue curves for carben, low alloy, and high tensile steels (Fig. I-9.1 of Reference la) of the present codes, labelled current fatigue curve.
The past " fatigue" curve is based on a 70 ksi ultimate tensile strength (UTS) material whose all,cwable stress range, S,(2) is f (1.5)(UTS)/(4) (0.9) where 0.9 accounts for A
the efficiency of a welded joint, and f depends on the nunber 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 number 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.
Assuming that for most piping systems the c.aximum local sr.ress index is not likely to be higher than 5, but higher than 1.4, we conclude from Figure A4-2 that piping systams designed in accordance with the past anu 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.
331.1 (1955) only; ASME I (1962) does not explicitly consider cyclic loads.
2.
SA = f (1.25 Se + 0.25 S ).
Using Se approxi=ately equal to Sh and h
Sh 5.0.9(1/4 UTS) gives SA f,f (1.5(UTS)/4)0.9.
A-57 l
-<1Sds 30hD Franklin Researen Center A Onimen af The Fransen inesame
10*
u
^
';p l
!EE nuh PAST CODE COtiSEllVATIVE' PAST CODE MAY y w y w g
(LESS TilAN 500 CYCLES)
BE tillCollSEllVATIVE pgg S
CODE PROBABLY l
10s Q
-UNCONSERVATIVE --
l K
7 (10)- 25 g
(BETWEEN 500 AND 100.000 CYCLES)
FO'R SION'FICANT l
A g Ot mRFt1T FATKit_lE CllRyg--
MORE IN N
g 100.000 CYCLES) 0
.N K(100)-3 0
N f-- K(500)- 5.0 s%
A 3
j k
K(10*) - 1.4 i
N w' K(1000) - 3.75' 3
i
~
l s'_
,(
10
~
~
K(10*)-1.8 K(10*)-1.0 PAST "FATIGilE" CilRVE -
s'~s 24 30 _
/
02
' s%
20 -
d 7
17 M&
-h
- I
/
10' l
I I IIIU l _ L LI LLij 1
1 I I Lljl i
l l l l l 11 1
i i:gli a
10 10 10s 10*
10s ta.
tJOTE: E - 30 x 10' ksl UTS < 80.0 ksi
--- UTS 115.0-130.0 ksi Interpolate for UTS 80.0-115.0 ksi
~
Figure A4-2.
Ileutgn Fatigue Curves for Carbon, Iew Alloy, and liigh Tensile Steels (For Ifetal Temperatures flot Exceeding 700*F) (Reference 4e)
2.
The current code considers the influence of thermal gradients threugh 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.p 3.
The current secondary stress indices C are either equal or less chan twice the corresponding streds 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.
N3-3653.2 gives a simplified expression for S which conservatively estimates the sum of primary and secondary and peak stresses as follows:
PD D
+
p" 11 2t 22 i
' 2(1.y) K Eal AT l 3
+KCE 3 3 ab
- aa ~ "b b
+ (t y) E3lAT l (1}
2 where:
K,K,K
= local stress indices i
1 3 3 AT = linear portion of thermal gradient through the y
thickness AT, = non-linear portion of ther=al gradient carough the thickness M = resultant range of moment due to service changes g
in temperature l AT l or mechanical loads such as earthquake C,C,C
= secondary stress indices 1 2 3 i
P = range of service pressure v = 0.3 Es = =cdulus of elasticity ti=es the =ean coefficients of ther=al expansion D = outside diameter of the pipe A-59 I
ddu Franklin Research Center sca
.om v en mm.
t = nominal wall thickness I = sectional moment of inertia T,(T ) = range of average temperature on side a(b) of 3
gross structural 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 past piping code (4] sets limits on the first two terms 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:
M.
=1f100)
Internes Messent Thermal Pressere
- t. seeings Laesing P'oing Presorts ans Jaists e,
r, s,
c s
ce c
r, e
e e
Stresent poe.
frome weses or einer essentanutase 0.3 La La*
La Le La L3 La Girtle lustt need Detwess strasgnt psee er betweme pee and butt weesses censonantstaa (a) thsen 0.3 La L18 L3 Le L1 LO 0.5 L1 (b) as weised t>J/16 si, [and Alt $3.2]
G.3 La L2' La La La La a.S L7 (c) as messed t$3/16 in. (or 4/t>G.1]
0.5 La L2' La L4 LS LS 0.5 L7 Girta fluet woes ta socket wese f!ttings, see on flanges, or socket wasang risagse 0.75 2.0 3.0 LS 2.1 2.3 La La 3.0 Lv : J tutt weies se strasgat spees (4) fban OJ La Ls Lg Lg L4 (g
L; (b) as weised t>3/16 ut, 0.5 L1 L ',
La L2 L3 L3 L2 (c) as messee tg3/16 in.
0.5 L4 2.38 La L2 1.3 L2 L2 Tasered tranation joints per NS4425 ape Fig. N5 423318Ma (a) fbse or ne girta wese ceser taan yW 0.3 L2 La 1.1 La L1 (b) as weisse 0.3 L2 La La L3 L7 Scenen connecumns car N5364paa L3 2.3 1.7 La L7 Cwwe seee or tutt weieing eeone one ANSI Bib g. ANSI 516.28 L3 L3' LO 1.3 u
La or MS3 SP44*L6 EIA*'*
Sutt aoseing-toes per ANSI 316.9 or M11 $P*4 mas LO LS 4.0 La L2 0.3 L3 Butt weneng resucers per ANSI 516,9 or M53 $P44 u La e'
La n
L3 0.5 L3 NOTES; 113 (as The venues or. f, mown 'or inese conisenants are felif C es = 0.,un. is not ressor inwi 048c,. ans ae.
aesseeente for corneenents imen we of rewieness not costaspe wue et #, mov ee oosennes Dv enweasvent tne Fester mes Q. Car. anere out of reunenses is estinee es teauneses muse of X, by tne fester #,os 0, nee = 0
,,ane M3 D.,.ee e momensune ousase diaerever of cross sentmen. in.
- s o
- I
- p og, 3
e annunues oussuno ammover of crees sortion. in.
- nonunes went tneennen..n.
er nne,e M
- 2 for foreme steses and nonferrous rnstonese seesse noser a _
- .c.
adevs see
- 3) if tne cross toetsC8e es out of rthares suen that the trees -
won.cnetene adeve secoen es amoromemotory eehotises, an ariseresse vesue of g
- 2.7 for austenme scenes neckr n K. may ne consense tv metiesvang tne taeusasse venues
_..ress mievs. and c n reemeum adeye of T, Dy tne faster #,e; Sy a yeses stresigtft at esogn temooreture. pm (Taeems 3 2.08 D.. - C.,,..
1.5
- Demgn humare. pm 8.e*1*
ag@
0, and t are defines in (al ane fbl.
g 1e0.485 7
i (2) Weees e acoeroenes wect ene recuorements of tnee Svenscoon.
emnere 0,
- aonunae ourmoe daamster, en.
Le6 # dias messe are defined as tnose weeeds tin.cn neue tteen p
- enternes crenewe, aus yound on gotn ensertor and esterior sur1 ace to re:Teowe tuas menernum wasue of creasuas en the lose vesse arregumences and swuus enen9es an. amour eue to evee unest caneeserereone
_ 4 __
Thectnem of nets reinforcorrent (totes g a manusus os esasucry of nieteral at room 'em.
ensees and outsioet saast not e;aceed O tt. No concavity perature, ces
- an tne root s.ue es carrensted. The harsnes comour snast Ciner svmews are ces.aee a dal.
aownwe naw a uoos We swasuras from tangent to I.4A
_Aklif) Researt. Center
-en
Table A4-8-(Cont.)
NB.3000 - DESIGN Tabie N5-M82.*.I mntess of mee er. on tesemus trenssteen s.ee of viene to Por Mg me nesn.nas treneteen earfames yesear enen 7 est, see MD
- V"'s s **'* rs '"*ss * "itone mammes on erunen 7 ens men.
p., y, I est nom.
g W, e vw,,*es',,*as%
- casesene namnuais en run j, T
~
I visuse W,, M,and W,are esseernense as fossemet l
y a
3 game.msn.3 eme.==i.
7 ir u,, one u,, n e tre sea. aseme,e= s. in.a u u,, ane u, neue eet.une as,.orse ugno, ene Jw,,.0. se A
s r is me
,6 J.
eessier ed M, or M.,meere d a r. r s.
i t
PerIsrween _-. _7 of tees tre Ma term of tauseens fel Ae swone is cogned as wenes not presnag me sammes (St. (103. (111. or (13 insel be nosaems tev me reseemmeng n-
._ for Muse esses. At tree wiversmanen ed a paraof tems:
lenensowise tsure wee ise sareigne snee vem a prtn ensee
~"""~""~
,,,4.#,,#
i,
- 03 mie s, e t A us u.
The C,, X,, C,, X,. and K, jnetems snese be me presume Er
-(1034(13 C,e 7.
- C,, -=
of me to.emmews wesens for me aoraptumanas west ane girse sunne. Per examese se tree intersmetten of art b
,,4,, 2, J 9' 8""""8 'em an asasesov toneaseines EausammeIlli C,et,3 Duet woe. C,".e 1.1X1.1*i.21. C, for a prun finee vues C
I, anserummeng a nosepasennes uimme sness tuo taeon as 24.
[33The sureus mesmo yvese are aemassmose onsy to taronen were
. we ssrogne see amen arenen ame nernes to tne suas survene and veneen maae me emoneenes resusrements lea vtr',,,)' Tg and tinuammene of NS4858 anse Fig.Ne4858.1 6 (48 # = aarvue see or messe reene..n.
2, e v#ne 7>
8 r
- mese reenas of creas saseen. wa.
- (De - et/2. sunsre r = nennines==es inschnen (5) The msums et menene, g,,. innes to ootainee from are-Far tennen sannousions per NS4843,se f eestese 3 ascue eierews of me pesag eveenne m -
_ c.msn Ns4872.
e'm T, #,,, ane T, are oetense in rag. NB.3684.1 8 3
w e estween as tne range of enemone somew=g asas.se manng For auseg toes our ANSI SISJ or MS SP 48:
e se nemettee coerenne crese, c'., a mesa reeus os asogneens arenari mae M
T, a nonunas vasa inesanum of osegneens arenen g
g A
armyne rgrees,esee
/
- .,
- mema reeius of congaseos run mee wy T,e nonense ies tn cnness of assigneuse non mee upa nwone se 9ame A
/
(8)Ineses are assuenese to taoores tronetson toonts vean a prin "f
- N "e ' ** * **8, '
2 nuse venne se tne tnen one of tne tronecon.
s 9
C,
- 1.3
- 0.003 (Dyn
- I.51&/rl 1M A gue not yeever tre 2.0 N or W 68e
/
C,*1.4*0.C0410,n
- 10to/tl 3
M,a manione as Powie A
./"
hase not yester snan 2.1 C
- 1.2
- 0.000 (0,W We = v u, ' **s, ' =as, '
g "2
(7) d,,
- 0.75 C,3Due not fees tnan f Q
- ,,
- 0.75 C,, aut noe i*ss enan I.0
&*w'n' Men' C,a
- t#,,,JTel Vap#,,,h"' (Ta/TrJ V,,,,'rg3. but noe
.. =
tems snan 13 Monumms cascunaeue for sowie as wwerseccan of nas ane A
Tn r',,,. T. and r,
are cotines in Fig.
3 twenen center linee Ns. me.18
&My3 T a
- 14 i
C,,
- OJ (#,,,/T,3 8#8 (,*,,,/#,,,J. Due not Isus unen 1.0 r,,
- 2.0 Two eresues se e,,s, snee no e =ma.m or 34
-- 23 4 "r8
'G
' "r2 (s) e, *
.nus aae enen iJ:s,
- 0Js C, r#
e, *..un-e r
- nomme e.se wed m.cene.
/
g,i g
O'
- eene raenas of curwee D.co er eemo
,
- m n co,se.e,
' to - m2
,,si "12 e
!!5 A-63 NUd Franklin Research Center A Q>sineum of The Fm J
t Table A4-8 (Cont.)
Table N54482.:.1 SECT!ON !!!. DIVISION 1 - SUBSECTION NB (914
- 8.,
- 0.75 C e IcJ Resumers us anon r, ane r,,0.10, i
Csa
- C.,
- 457 iA,,,/7,J'd, tnet not les than 10 l
Am
- musa remos of eenagnoses run esse T, e non=nes amas tnicanons of een gnesed 'nen pes C.
- 3
- 4C058 avDe.re i
K,e
- K,re1D
=
1Ma,D, -033 (103 The K ino6am even for fittings per ANSI SMS. ANSI Ce *1 e0.".SeDeJe2 r
818.23.or assa spas assey oney to seemeem f 6tungs ames ne
_. artneruvents. er eene seerensene stree remmee weiere Oeh,, s tne targer of O, A, one 0, A,.
on the teases enereof. For fittings amen tenytuenas tsiset i
weses, tRe K enmens unesse eness tus rmstenosene try the 1.1. for (WJ Reemesse ise seneft r, ansler r, <410, aume noser ac softnee we Nese 2; try I.2 for vunnes not mesesngme a u ; for #use mens' C,
- 1
- 40044Sa' fDW
(11) The serous weesee twen oreeser sareams sensen ensur ies een tieey of a fitang. It es not renseree to taae arte geoeums of g,. g e ggg gga yg,,,,
seems wooeces for tvue ossweg produsts nuen as a too ame a
'*8 easer. er a eso one a pren eurt mee anon tvooed tow tier
,,,,,,g
,,,g,,,,q,,,, g, g, g,y,,
escass for tne caso of curwee peo or bust osaseng neous esamese togemer or foened try a pass of strument snee ce songm of veneen es less enen I onos caemoser. Per tnse (148 The K ineeems eness in (a). !bl. ane ici assey for resumere messfie came tne tirous inges f or tne curwoe snes or tnset streories to tne conneenres poo onen rtume or aa==sese vueuseng esco== rnuss ne mu6eiches try tnet for une prm ouet swtn weies as cefutee.a footnoes (2). Nore that tne esses. Essieses frone tne nuesrect.comen are ene d, ane C',
connserung garvn wune muss mes ce enessas sammreamy for inescen. Thew vesus es to to: #,
- 14.C',
- 0.50.
(al For resumers __-_
to esos seem /tuae pree butt (13 4.s corones as ene maannueve -_-
ensenween as viomme we P6g. N4 42221 A eve of 6 tems enan 2/22 in. may me ames uses proviene tne emanaer neemmsen es sommfies for L,,,
'sericatesse. Por #was woner, seeened we foernese (2. & rney K.
- 1.1 - 41
, but not see toten 1.0 es taman se sere.
V C.s,4i, K,
- 1.1 - 41
. but not seus enen 1.0 4C&m L9 w
venere Q'Cwm,s the smuseer of L,/v"6"" and q
L, /v'D e t,.
y 6
f///////
.na en, IBJ For roeusere _,
_ to pie
^
i f e,y bu
- o. -,,.,, > 2n.s ween a.s mussed geret
. a,i.
l 0.1
't { eg g
g hm
,3 6
km i
s
- 1.2 - 0.2
. bus nas seus mei 1.0 i
rg Op Mude 4
9 4,
- IJ - 48
. but not fem enes 1.0 v 0 seeds
- r.
- neemnal west enwoness, targe one anorejf.VCWm is the enester of L,/vM and L /w G e ',.
f, e aesmnes wese meanness enma and S
O, e aan=nes outsee sameter,large one 0, e neshnas entmos esameter, ames eroe (cf Por roeucers conneetos to poo wie aa messes gartn
.
- cone ance, oog tnatt anses, senere r, or r, 4 0/18 en. or 4, /t, or 4,a, >
g 0.13
- -SJ The w= eses seven n teJ ano (el aserv.4 the fosounrog L.,,
conessons are nies.
K,
- 1.2 - 42
. tnet not less then 1.0 (f1 Cano angse. 4 eees not oncese 50 deg. ane ene vosm soeuser se conconme.
(2J The was enesanese es not tese than r,m inrougmous the La K,
- 15 - 1.5
_. but aoe tens inan 1.0 boey of me recusar, esceos in arie.evunesistory aanneene to the evaanencas earnen ese the enes end.
an** tne tfuceness snas not be W non r.,,,. Weel erwsmeerme r.,
ane t,m are to to ootmaneJ by
"'_Lm/VDwm is tne enesser of L,fyQ,,,,
Eouasion (1). Ns.*s41.1.
l.,/JO, r,.
1:6 1
A A-64 brankHn Research Center i
4 ca a,e Two a:
i Noting that M = Ag and conservatively assuming that a auclear power plant designed in accordance with past codes is such that S
=S and recalling that g
g C :
2 1.9 1, the second term in the expression for S becomes:
p D
f D
\\
22 i
21 b 1*9 if l
= 1.9 Sg (121 2 0.65f 1 K2 (UTS)
(2a)
K)=
2 for ferritic steels 0.63f A,,i K2 (UTS)
(2b)
~
for austen1cic steels Pas t piping codes determine pipe chickness in accordance with the formula (11 PD 2(S + 0.4P) + C (Equation 1. Section 1 of Reference 4),
t =
h s
n' t
where:
P = design press'2re D = outside pipe diameter C = allowance for corrosion S = a 1 wa le stress at temperature h
t = minimum pipe wall thickness When C is small compared to the thickness and 0.4P is sna11 compared to S, the minimum thickness is approximated by PD ea 2S t
Since the actual pipe thickness, e, is not less than c,, we have 1
PD 3 (UTS)(0.9) ferritic steel g 13h"
(
2t f (UTS)(0.9) austenitic steel 1.
Based on y = 0.4 for ferritic materials below 900*F.
A-65 IM M-' \\ TMri@Rkhmitafaicrn
e i
Assuming that the range of service pressure P, is a fraction A of the design g
pressure, we have
?D A ?D 1/4 A (UTS)(0.9) ferritic steels I
g g
3 AS
=
h " 1/5 A (UTS)(0.9) austenitic steels p*
g so that the firs't term in the expression for S may be put in cite form p
fPD fk (UTS) g C1 (0.9) ferritic steel (3a)
C -
=
1 ( 2t j tyA1 (UTS) g C1 (0.9) austeniti.c steel (3b)
Substituting Equations 2a and 3a on Equation 1 and neglecting the fourth term in Equation 1, we obtain:
1 S, = 7 (0. 9) A C (UTS) + 0.65f A,K.,
(UTS) r
- 2. -
1 Es
+ (1_.9) 33 N l ' K I
2 3 2(1-v) l (la) for f erri:1c steels.
Similarly substituting Equations 2b and 3b in Equation 1 and neglecting the four h tern in Equation 1, we obtain:
0.9 S
A, (UIS) K,C1 + 0.63f A.,,
K.,
(.UTS)
=
p 5
+ (1-v) Esl AT l + K (ATl (lb) 3 2G - v) 1 4
for austenitic steels.
These expressions can be further s1=plified by noting from Tables I-3.0 and I-6.0 (le] (Winter 1978 Addenda) that:
3
-6 Ea 27.9 x 10 x 7.3 10 ksi
=
, y) 0.7
.r er c steels
=7 1
-6 Es
, 28.3 x 10 x 9.4 x 10
= 0.380 for austenitic steela (1 - v)
O.i ay A-66 nklin o. arch Cene~
Substituting appropriately in Equations la and lb and multiplying the second term by 1.3 to account for ::iovements of pipe ends attached to equipment, we have:
S = 0.23 A (UTS) K C + 0.85f A K2 (UTS) 7 21
+0.291laTl+0.145K3laTl (la) 2 y
for ferritic steels S = 0.18 A1 (UTS) K C1 1 + 0.82f A.,, K2 (UTS) p
+0.380lATl+0.190K 0
(D}
2 3
1 for austenitic steels where:
7
= (range of service pressure)/(design pressure) = f A y UTS = ultimate te=sile strength of material at 70*F f = stress-reduction factor (see Table A4-7)
A
= (Change in temperature for i ch service cycle] divided by (nardmum operating temperature - 70*F]
=lAT,l/l(T)
-70*Fl o TJLX
(,C,K 'l0T l 'K ' l0T l = previously defined.
7 2 2
3 i
The alternating stress intensity, Sg, is one half of the peak stress intensity, S ; that is:
S
=ls alt 2 p For a given value of alternating stress entresponding to actual n service cycles, the nu=ber of such cycles N allowed may be found from the applicable design g
fatigue curve, Figure I-9.0 [12]. The usage factor for the given n service g
cycles is defined as :
l n,
U ="
i Ng A-67
$$hranklin Research Center
1 i.
The cumulative usage factor, U = IU shall not exceed 1.0 as required by NB-3222.4(e)(5) of Reference Ib.
Equations la and Ib may be used to evaluate Class 1 piping designed in accordance with past code requirements from the viewpoint of present code re-quirements.
Some e:ramples will be used to illustrate use of the formulae.
Examnie 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 chick 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 ::ower 2.
500 reactor trips from 100% power.
F.xnsination of Figure 4-8 of Reference 10'shows the reactor coolant temper-ature as a straight line function of NSSS pcwer.
Considering the hot temperature function, note that this full power T = 594*F and at 0% power T = 532*F.
This indicates that the ta=perature change associated with the reactor trips is 62*F.
For each AT, we shall assu=e that AT = 0.75 aT and AT2 = 0.25 ar.
7 A note accurate determination of AT and aT nay be obtained from Reference y
3 11, so that:
Service Cvele - 1 1 = 15,000 AT of Service Cycle 1 = 62*F n
AT = 0. 75 x 62 = 46.5'F 1
f = 0.8 AT = 0.25 x 6.2 = 15.5*7 l
A-68 O
ilihJ Fgnklin,Research center
l 1
Service Cvele - 2 n
= 500 AT of Service Cycle 2 = 62*F 2
AT = 0.75 x 62 = 46.5'?
1 f = 1.0 2 = 0.25 x 62 = 15.5'T AT
~
Elbow Consider an elbow in which the band 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 elbov K1 = 1.0 C = (2R r)/(2(R r} }
= (462.5 - 23.13) /(2 x (231.25 - 23.13)]
= 1.06 K = 1.0, K3 = 1.0 2
Lensitudinal Butt Weld-Straizht Pine A longitudinal butt veld flush in a straight pipe would be a more critical element to investigate since for this element:
K = 1.1, C = 1.0, K
- 1*1' K = 1.1 y
2 3
3 ranch Connections A branched connection which =ay possibly have been used to connect the 12-in Schedule 140 316 stainless steel surge line from pressurizer to the hot leg would have stress indices as follows:
K = 2.0, K = 1.7, K - 2.2, C = 1.5 2
3 1
1 and obviously would be =ost critical. These K and C values are taken from the 7
Sc==er of 1979 Addenda [1].
b) rank 5n Research Center o
Deternination of Usage Factors UTS = 70 ksi (ASTM 516 - Gr. 70)
~
ch For the i service cycle:
(S ) = 0.23A x 70 x g C + 0.85 x 70 f.K A p
2n
+0.291laTl+0,145KlaTl 2
3 1
Assuming that the pressurizer maintains pressure within + 50 psi during these service cycles, then:
100 11 = 2500 = 0.04 so that (S )
= 0.644 K C + 59.5 A fK2+0.291laTl+0.145KlaTl 2
3 1
AT Determination of A for each service evele A
i
=
21
[(T )nax - 70* ]
o (T )
= m=v",um operating temperature = 594*7 AT of Service Cycle 1 = 62*F AT of Service Cycle 2 = 62*F A
= 62/(594 - 70) = 0.12 21 A
= 62/(594 - 70) = 0.12 22 finally (Sg)
= f (S )
A su::snary of the results for each of the two service cycles as it affects the usage of the three ele =ents is given in Tables A4-9 through A4-11.
It is apparent {
from the usage factors calculated in these tables that cu=ulative da= age from l
{
cycles 1 and 2 is negligible.
A-70 nkun Researca center
- - - - - - - - - - - - - - - - - - - - - - ' ~
- ~ ~ ~ ~ ~ ~ '
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, K3 = 1.0 1
1 Ser rice cvele - 1 g = 15,000 f = 0.8 A
= 0.12 21 ai
= 15.5 r aT = 46.5 7 2
1 S =0.644'qC + 59.5 f K A
- 0. 291laT l + 0.145 K laT l = 18.7 kai 2u 2
3 y
S
=1 S
= 9.3 ksi alt 2 p 6
Ny > 10 (See Figure A4-2)
U
= 0.02
=
7 i
Ser rice cvele - 2 n = 500 f = 1.0 A
= 0.12 2
22 aT = 15.5'F AT = 46.5 'F 2
y i
S =0.6447'll+59.5fK.,A.,*0.291laT.,l+0.145Klarl=19.1ksi C
2.
3 1
p
=1 S
S = 9.5 ksi alt 2 p l
6 "2
N2 > 10 (See Figure A4-2) '
U
=-=0 2
2 l
Ut+U2 - 0.02 A-71 l
O
(
J8J Fynklingh Center
=
Table A4-10 Usage Factors Due to Thermal Gradient Through Thickness Frample: Hot Leg of Palisades Pri: nary Coolant Piping Piping Element: Longitudinal Butt Weld-Straight Pine K = 1.1, C = 1.0, K2 " l'1' K ~ 1*1 y
y 3
Serrice Cvele - 1 a = 15,000 f = 0.8 A
= 0.12 4T = 15.5 F 4T
= 46.5 F 2
1 S
0.644 K C + 59.5 f K A221+0.291laTl+0.145K
=
A
= 18.9 ksi 2
3 1
S
=1 S = 9.5 ksi alt 2 p 0
N > 10 (See Figure A4-2)
U
= 0.02
=
1 Service Cvele - 2 n = 500 f = 1.0 1
= 0.12 2
3 AT = 15.5 'F AT = 46.5*F 2
y S
0.644 g C + 59.5 f K A2 22 + 0.291laT l + 0.145 K laT l = 20.5 ksi
=
2 3
1 g,=fS,=10.2ksi S
l 6
2 I
N2 > 10 (See Figure A4-2)
=0 U
+U
= 0.02 2
A-72 O
dudU Franklin Research Center 4 % e n. r-m=
. ~ -. _
Table A4-11 Usage Factors Due to Thermal Gradient Throu2h Thickness Exa=ple: Hot Leg of Palisades Pri=ary Coolant Piping Piping Element: Branch Ccnnection (K and C frca Summer 1979 Addenda (1])
y K = 2.2, C = 1.5, K = 2.0, K3 " 1*
1 1
2 Service Cvele - 1 n = 15,000 f = 0.8 A
= 0.12 21 aT = 15.5 r aT - 46.5 r 2
1 S = 0.644 gC + 59.5 f K A221+0.291laTl 0.145 K laT l = 29.5 ksi 2
3 1
S
=1 S '= 14. 8 'c1 alt 2 p 6
"I Nt > 10 (See Figure A4-2)
U = - = 0. 02 1
Service Cvele - 2
= 500 f = 1.0 A
= 0.12 2
AT = 15. 5 * ?
a2 = 46.5*F 2
S = 0.644 qC + 59.5 f K A 0.291laT l + 0.145 K 1l = 25.2 ksi 2 22 2
3 S
=1 S = 12. 6 'ci alt 2 p 6
N2 = 10 (See Figure A4-2)
U
= 0.0005
=
2 2
= 0.0205 Ut+U2 A-73 O
""UI F pjy"p 3rgrqrq F-pr 3
Example 2 Tne 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 AT i k2i i-ni
('F)
(laTl/(524))
i 1-15,000 (10% to 100% full power) 59' O.113 2-15,000 (50% to 100% full power) 31' O.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 valves k with the value of 0.12 obtained in 21 Example 1, the usage factors associated with the above four additional cycles are negligible.
Comparison With ASME I (1962) Requirements Piping from a reactor vessel up to and including the first isolation valve external to the containment structures could have been designed and built to the following requirements:
a.
ASME I (1962) b.
ASME I (1962) and B31.1 (1955).
If requirement (a) was invoked, expansion stress limits due to cyclic tnermal loading are not specifically imposed. However, ASME I (1962) does require consideration of loads other than working pressure or static head, which ' 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 (1962) criteria should furnish details as to how thermal stresses ware considered.
A-74 "d Franidin Research Ceaer A Cnnsson W The FrerwJin insatute 9
l If requirement (b) was invoked, then paragraph 102(b) of Section I [4]
requires that valves, fittings, and piping for boilers as prescribed in Section 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 Section I.
Nuclear Code Cases N-1, N-2, N-4, N-7, N-9, N-10, N-ll, and N-12, (
when invoked, impose requirements as follows:
1.
Code Case N-1 requires that nuclear piping (for which loss of fluid could result in a radiation hazard) may be designed to B31.1 (1955) supplemtated by the requirements of case interpretations identified by the pre fix "N. "
2.
Code Case N-2 requires that valves used in nuclear power systems:
- a. be of materials recognized by ASA B31.1-1955 and conform to a recognized standard (e.g., ASA B16.5).
- b. meet physical and inspection requirements of Code Case N-10.
- c. have a positive sealing or some provision for stem and brnnet leakoff control.
- d. screwed end valves (in which the thread provides the only seal) are not permitted.
3.
Code Case N-4 permits the temperature limit of 100*F for hydrostatic media to be exceeded.
4.
Code Case N-7 permits the use of nuclear piping made from austenitic stainless steels, provided that:
a.
materials conform to one of the following ASTM specifications:
A376, A358, A312, and A430 for piping; ASTM-A403 for welded fittings; or ASTM-182 for forgings.
1.
Mecnanical Engineering, August 1962 (Code Case N-1, N-7), December 1960 and October 1964 (N-9), April 1960 (N-10), July 1961 (N-2), December 1960 (N-4),
and November 1961 (N-11 and N-12).
A-7 5 aJ Franklin Research Center aom aorm rr m.an.
.,....._-___.,,_m_.
b.
full radiography of longitudinal and circumferential welds is performed; however, fluid penetrant methods are permitted when size or configuration precludes full radiography, or for services at or near atmospheric temperatures up to 212*F provided that piping is tested at 1.5 times the maximum allowable working pressure.
I c.
allowable stress values are used as shown in the following table:
Allowable Stress Values Steel Type 321H 321 Temperature 347H 347 (F*)
304H 304 304L 361H 316 316L 348H 348 309 310
-20 to 100 18750 18750 17500 18750 18750 17500 18750 18750 18750 18750 200 16650 16650 15300 18750 18750 16250 18750 18750 18750 18750 300 15000 15000 13100 17900 17900 14500 17000 17000 17300 18500 400 13650 13650 11000 17500 17500 12000 15800 15800 16700 18200 500 12500 12500 9700 17200 17200 11000 15200 15200 16600 17700 600 11600 11600 9000 17100 17100 10150 14900 14900 16500 17200 650 11200 11200 8750 17050 17050 9800 14850 14850 16450 16900 700 10800 10800 8500 17000 17000 9450 14800 14800 16400 16600 750 10400 10400 8300 16900 16900 9100 14700 14700 16200 16250 803 10000 10000 8100 16750 16750 8800 14550 14550 15700 15700 850 9700 9700-16500 16500 14300 14300 14900 14900 900 9400 9400 16000 16000 14100 14100 13800 13800 950 9100 9100 15100 15100 13850 13850 12500 12500 1000 8800 8800 14000 14000 13500 13500 10500 11000 1050 8500 12200 13100 8500 7100 1100 7500 10400 10300 6500 5000 1150 5750 8500 7600 5000 3600 1200 4500 6800 5000 3800 2500 1250 3250 5300 3300 2900 1450 1300 2450 4000 2200 2300 750 1350 1800 2700 1500 1750 450 i
1400 1400 2000 1200 1300 350 1450 1000 1500 900 900 250 1500 750 1000 750 750 200 i
d.
reheat treat.ng at 1950*F for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per inch of thickness for pipe sections subject to cold or hot formings followed by fluid penetrant testing of all accessible surfaces was performed.
3.
Code Case N-9 allows tne use of centrifugally cast austenitic steel pipe for nuclear service provided that specified chemical and r?chanical properties are satisfied:
inside and outside surfaces shall (1) be machine finished to 250 micro-inch RMS or 225 micro-inch AA or finer; (2) be pressure tested at 1.5 times the rated pressure and fluid penetrant inspected; (3) be fully radiographed; (4) meet the requirements of ASTM E-71 for Class 2 quality casting; and (5) be reheat treated at 1950*F for hot formed sections.
A-76 MJ FrankHn Research Center a om.aa w n. Frnuwamu.
...., - -. _ -. _ _ - - - +..,.
Stress allowaoles should be in accordance with-the following tablet *-
Maximum Allowable Stress Values in Tension, psi ASTM A451 Grade CPPS CPF8M CPF8C Spec. Min.
Tensile 70000 70000 70000 Temperature (*F)
-20 to 100 17500 17500 17500 200 15700 16900 17000 300 14250 16500 15600 400 13100 16300 14200 500 12200 15900 13000 600 11700 15350 12200 650 11500 15000 11900 700 11300 14700 11700 750 11100 14350 11600 800 10900 14000 11500 850 10650 13500 11350 900 10400 13000 11200 950 10100 12350 11100 1000 9850 11700 11100 1050 9600 10600 10900 Note: These stress values are based an a casting quality factor of 1.00.
4.
Code Case N-10 permits the use of cast austenitic steel bo tt welding fittings for nuclear service provided that ASTM Specifications A-351 and ASA B16.9 are augmented by the following requirements:
a.
specified chemistry and mechanical properties shall be satisfied
(
b.
fittings shall be finished to 250 micro-inch RMS or 225 micro-inch AA or finer c.
fittings shall be tested at 1.5 times the rated pressure d.
fittings shall be inspected by the fluid penetrant method and be fully radiographed in satisfaction of the ASTM E-71 requirements for Class 2 quality castings e.
fittings shall be heat treated at 1950*F followed by rapid cooling in air or a liquid medium f.
Stress allowables shall be in accordance with the following table:
- Values are applicaole only af ter October 1964.
l A-77 W
2J Franklin Research Center 4 o.=a w N rrwoon muu.
5
Maximum Allowable Stess Values in Tension, psi ASTM A351 Grade CF8 CF8M CF8C CH2O CK20 Spec. Min.
Tensile 70000 70000 70000 70000 65000 Temperature (F*)
-20 to 100 17500 17500 175(3 17500 16250 200 300
~ 15700 16900 170 A 16100 15300 14250 16500 15600 15150 14900 400 13100 16300 14200 14600 14600 500 12200 15900 13000 14550 14450 600 11700 15350 12200 14450 14450 650 11500 15000 11900 14400 14400 700 11300 14700 11700 14350 14350 750 11100 14350 11600 14300 14300
-800 10900 14000 11500 14150 14150 850 10650 13500 11350 13900 13900 900 10400 13000 11200 13500 13500 950 10100 12350 11100 12500 12500 1000 9850 11700 11000 10500 11000 1050 9600 10600 10900 8500 9750 5.
Code Case N-ll indicates that any sound me'.ns of providing for thermal expansion may be used and the following requirements must be me t:
a.
Must meet requirements of Section 6, Chapter 3 of ASA B31.1-1955.
b.
Material recognized by ASA B31.1-1955.
c.
If sliding or swivel type, have a positive seal or leakoff control.
d.
Provide for thermal expansion due to rapid temperature fluctuations.
6.
Code Case N-12 provides a procedure for qualifying new materials for use in nuclear piping systems. The following subjects are discussed: ASTM identification, alternate identification, creep and stress rupture data, physical properties, heat treatment, hardness measurements, impact strength and transition temperature, radiation and temperature effects, microstructure variations, availability, weldacility, and test results.
The following is concluded:
1.
If ASME I (1962) was used, the Licensee should furnish information regarding how ' expansion thermal stresses were determined. This g
A-78
$dd Franklin Research Center A Drumon of The Frannan msecure
information should be reviewed against current fatigue requirements, especially for services with more than 500 cycles. Fracture toughness should be reviewed against current requirements. See Section 4.1.1.
2.
If ASME I (1962) and B31.1 (1955) were used, the calculations for fatigue evaluation should be reviewed, especially for services with more than 500 cycles. Fracture toughness should be reviewed against current requirements.
See Section 4.1.1.
3.
Piping built to B31.1 (1955) and the code cases should be reviewed for satisfaction of currant fracture toughness requirements.
See Section 4.1.1.
4.
When Code Cases N-1 plus either N-2, N-7, N-9, or N-10 were invoked, current full radiography requirements would be satisfied.
5.
When Code Cases N-1 plus either N-7, N-9, or N-10 were invoked, current stress allowables would be satisfied for a temperature range up to 650*F.
Comparison With USAS B31.1.0-1967 Requirements USAS B31.1-1967 (13] is essentially the same as the power piping portions of ASA B31.1-1955.
The comparison of ASA B31.1-1955 requirements with current requirements would also apply to USAS B31.1-1967 requirements.
l Comparison With ANSI B31.1 (1973) Requirements The ANSI S31.1 (1973) (14] power piping code requirements applicable to I
l this review are essentially the same as the 1955 Code except that the Summer 1973 Addendum modifies the stress intensification i factors for butt welds and l
tillet welds and introduces new factors for 30* taper transition, concentric reducers, and branch connections. Comparison between these factors and half tne C factors (but not less than 1) from ASME III (1977) as shown in Table 2
A4-8 of this appendix indicatus that the i factors are conservative when compared to current values.
ANSI B31.1 (1973) also introduces an equivalent full temperature cycle formula for variaole temperature cycle service. A fatigue evaluation account-1 ing for local discontinuities is not required by either the 1973 power piping code or B31.1 (1955). The fatigue evaluation example and conclusions based on A-79 CU Franklin Research Center A om.ca a m numa m.m.
. - -._.._..._..,_ __,_... ~
.~,
a comparison between the 1955 power piping code and current requirements (see Section 4.2 of this appendix) are also applicable to a comparison of the 1973 power piping code witn current requirements.
Comparison With USAS B31.7 (1968-Draf t) Requirements
~
The following items in the USAS B31.7 1968 Draft Code for " Nuclear Power Piping" (15] are similar to items in the current code:
1.
Piping systems are designed to Class I, II, or III requirements, as given in Subsections 1, 2, or 3 of B31.7 (1968).
2.
The snear dieory of failure with its associated stress intensity concepts and limits for primary, secondary, and peak stress categories for Class I piping are the same.
3.
The formula for peak stress intensity range for Class I piping is the same, and local and secondary stress indices are used in both codes.
4.
Both codes require full radiography fo,r circumferential and longitudinal butt welds for Class I and II piping.
Differences between the USAS B31.7 1968 Draft Code and current requirements are summarized as follows:
1.
Stress indices in USAS B31.7 (reproduced in the table "USAS B31.7 (1968 Draf t) Stress Indices") may in some cases be lower than those currently required. For example, for B1 the stress index for a girth fillet weld-to-socket weld fitting is currently 0.75 (Table A4-8) compared to 0.5 in B31.7.
2.
Fracture toughness (impact testing) requirements are not specified in the older code.
3.
Stress limits for the equivalent service levels C and D (emergency and faulted) conditions are not specified.
In conclusion, piping built to the B31.7 code (15] should be reviewed for the differences noted aoove and evaluated against current requirements.
We lding Requirements Full radiography of welded joints in piping, pumps, and valves as stipulated in past (4, 5] and current codes (1, 16] depends on weld joint A-80 JJJ Franklin Research Center A Cesen of The Franen msname
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.dh Mj Franklin Research Center a cm on or n. rr n m n
category, pipe size, and code class as shown in the table, " Full Radiography Code Requirements for Welded Joints in Piping, Pumps, and Valves."
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
~
enforced.
I Full Radiography Code Requirements for Welded Joints in Piping, Pumps, and Valves Current Codes ASME III (1977)
ANSI B16.34 (1977)
Past Codes (la 2I Description of Class Class ASA B31.1 (1955)
Welded Joint 1
2 3
Standard Special
& ASME I (1962)
A. Longitudinal 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 pipes, pumps, and valves of nominal pipe size exceed-ing 4" as follows (1) Bu tt-wc'.ded Yes Yes No No Yes No (2) Corner-welded Yes Yes No No Yes No full penetration (3) Full penetration les Yes No No Yes No 1.
Applicable also to E 31.1 (1967) and B31.1 (1973).
B31.7 (1968) requires full radiography for circumferential and longitudinal butt welds for Class I and II piping.
2.
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 (1962) requirements.
3.
Except when specified by material specification for piping in excess of 2 in nominal diameter.
4.
Wnen either member thickness exceeds 3/16 in.
A-82 C Franklin Research Center A Dms.on of The Frarmhn insanee 9 _ - _ - - - -. _. -. -,
,.____.__-_-m_.-,,,
4.3 PRESSURE VESSELS The past code requirements for pressure vessels are given in one or more of the following ASME Boiler and Pressure Vessel Codes depending on the SEP nuclear plant group as defined in Table Al-1.
~
Group Pressure Vessel Code I
ASME III (1965)
ASME VIII* (1965)
II ASME VIII* (1962)
III ASME VIII* (1959, 1956)
The current code requirements -[1] and the past ASME VIII (1962) code alfferences are summarized as follows:
g I
Fracture Toughness - Class A Vessels Except for containment vessels, which are covered by Code Case 1272N and outside the scope of this study, impact test requirements for primary vessels (the equivalent of Class 1 vessels) and secondary vessels (the equivalent of current Class 2 or Class 3 vessels) designed and built to ASME VIII (1962) were significantly less severe than current requirements as noted by the following comparison table:
Past(1)
Current Description Requirements Requirements Maximum Temperature of LST o Class 1 - LST-60*F Impact Testing When o Class 2 - LST Required o Class 3 - LST LST Above Whicn Impact
-20*F o Class 1 - None Testing Not Required o Class 2 - See Figure A4-1 o Class 3 - 100*F Specimen Notch Type U or Keyhole V
l Minimum Absorbed Energy 5 to 15 f t-lb o class 1 - 50 ft-lb depending on o Class 2 and 3 - Not l
specimen type specified for thickness less than 2-1/2 in; 50 ft-lb for thickness greater dian 2-1/2 in Minimum Lateral Not specified o Class 1 - 35 mils o Class 2 and 3 - 20 to 40 mils, depending on thickness
- Plus nuclear code cases.
1.
See UCS-66(d), UHA-51, and UG-84 of Reference 3.
A-83
- A NU Franklin Research Center A Dews.on of The f re insetute
It is apparent from the comparison table that current fracture toughness requirements are significantly more sever 2 than past requirements when impact testing is necessary.
Use of Tables A4-4, A4-5, and A4-6 will aid in evaluating material toughness of vessels built to the past code.
Design Requirements Vessels built to ASME VIII (1962) were not classified with regard to quality class. Code" Case 1270N, when invoked, classified nuclear vessels within the scope of this study as follows:
Vessel Tvoe Current Classification Reactor Vessel (outside scope)
Primary Class 1 Secondary Class 2 or Class 3 Containment Vessel (outside scope)
Code Case 1271N deals with modifications to Section I and Section VIII rules for safety requirements for de' vices such as pop-type safety or relief valves, direct reading pressure gages, inspection openings, gage glasses, water columns, gage cocks, and rupture disks.
In general, the code case eliminates the requiremeTts for these devices or provides for the safe containment and disposal of the effluent of such devices if they are inatalled and activated by an accident.
Safety devices other than relief valves are considered outside the scope of this study. Section 4.5 of this appendix reviews the structural integrity requirements of valves; cperational l
requirements were considered outside the scope of this study.
Code Case 1272N dealing with containment vessels and intermediate contain-ment vessels (outside the scope of this study) may nevertheless have been l
invoked for SEP pressure vessels. The provisions of 1272N are briefly summarized as follows:
1.
Stress relieving of containment vessels not inside a heated enclosure is not required provided the vessel material shall conform to ASTM specifications SA-300 and SA-350 for plates and forgings, respectively.
In addition, these materials shall meet the impact test requirements of paragraph VG-84 at LST -30 *F but not less than
-84*F.
In addition, the thickness of shell and head shall not exceed the thickness for which stress relieving is required by UCS-56, except that for P-1 materials stress relieving is not required for thickness of 1-1/4 in to 1-1/2 in, provided a preheat of 200*F is used during welding.
A-84 d%
M Franklin Research Center A Dmmon of The Frarwen hstade
~_
2.
Stress relieving for intermediate containment vessels not containing radioactive materials is not required except as may be required by Section VIII.
3.
The mandatory minimum corrosion allowance provisions of UCS-25 for compressed air service, steam service, or water service are not applicable to containment and interrediate containment vessels.
Code Case 1273N, when invoked, imposed the following additional requirements on primary (Class 1) vessels built to ASME VIII (1962):
1.
Thicknesses shall be no less than that required by the code formulae.
2.
Stresses due to tnermal loads combined with pressure loads cannot exceed 1.5 times allowable working stress, that is 1.5XS.
3.
The maximum allowable bolt design stress may be based on the properties of the heat-treated metal for operating metal temperatures 100*F or more, below the tempering temperature, provided the s
allowable stress does not exceed 1/3 YS at the tempering temperature and the operating metal temperature does not exceed 800*F.
4.
Creep and stress rupturr properties must be considered for long-term exposure at temperature s that will assure adequate safety.
5.
a.
Compensation shall be made for all openings regardless of diameter.
b.
When compensation is totally in the nozzle, the nozzle should be attached by a full penetration weld.
c.
Thermal stresses and external pipe reactions should be considered.
d.
Full penetration welds should be used wherever possible, except where not practicable, such as at closely spaced openings.
6.
All welds are to be fully radiographed except where not practicable, such as at closely spaced openings.
7.
Although no specific rules are provided for corrosion, radiation effects, transient tnermal stresses, mechanical shock, and vibration loads, these factors should be considered to obtain desired vessel life.
8.
Particular consideration should be given to quality of materials, faorication, and inspection.
9.
Cladding thickness in not to be included in code design formulae.
The past codes do not specifically consider loading conditions, other than design, operating, and test.
The FSARs for specific SEP plants may, however, consider the equivalent of emergency and faulted conditions. A A-85 d FrankDn Research Center Aom.onorm n.n am.ou.
e discussion of the evaluation of the FSAR stress limits for these loads against current limits is presented in Section 4.1.4 of this appendix.
l Stress limits for vessels which would currently be classified as Class 2 or 3 are essentially the same as for 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.
Secondary vessels which would currently be classified as Class 2 or Class 3 vessels snould be evaluated against current Class 2 or Class 3 code require-ments, witn 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.
Fatigue Requirements for Pressure Vessels Class 1 vessels designed to the current code are required to be analyzed for cyclic loads unless they can be shown to be exempt from analysis for cyclic service by demonstrating compliance with all the conditions of NB-3222. 4 (d) of Reference Ib as follows:
(1)
Pressure Fluctuations:
the specified full range of pressure fluctuations during normal service does not exceed:
(1/ 3) (Design Pressure) (S /S )
a m where:
Sa = alternating stress from fatigue curves corresponding to the number of pressure fluctuations Sm = allowaole stress intensity at the service temperature (2)
Atmospheric to Service Pressure Cycle N2 < N(3S )
m A-86 g
OUd Franklin Research Center A DMeson of The Fransen inscrure
where:
N2 = tce maximum number of atmospheric to service pressure cycles N(3S ) = number of cycles from design fatigue curve for Sa = 3Sm 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:
Eri i S /(2Ea) (i = 1,2) a 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 a = instantaneous coefficient' of expansion, mean value (see Table I-5.0 of Reference le)
Sa = alternating stress from design fatigue curve corresponding to 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 a
cycles.
(4)
Temperature difference - dissimilar materials - see paragraph NB-3 222. 4 (d) (4) of Reference Ib (5)
Mechanical loads - stresses due to mechanical load fluctuations (excluding pressure) such as pipe loads on nozzles less than the value of Sa from the design fatigue curve corresponding to the nummer of load fluctuations.
Fatigue evaluation was not required for Section VIII vessels.
It is recommended that Section VIII vessels which would be currently categorized as Class 1 (e.g., pressurizer) be reviewed to see if the pressure fluctuation and atmospheric to service pressure cycles, conditions (1) and (2), are satisfied.
The information nee 2ed to see if conditions (1) and (2) are satisfied should be available from the FSAR or other sourc(.
It is also possible to see if condition (3) is satisfied by assuming that the temperature difference between two adjacent points at vessel inlet, outlet, and feedwater nozzles is equal to g
A-87 LiJ' Franklin Research Center
- om.aa or w n.a.u.a
the fluid temperature transients given in the FSAR.
Note that fluid tempera-ture transients would be an upper bound on metal temperature differences between adjacent points.
Accordingly, satisfaction of condition (3) based on fluid temperature transients implies actual satisfaction of condition (3).
Non-satisfaction on fluid temperature transients does not necessarily imply actual non-satisfaction of condition (3) based on actual metal temperature differences which can be determined on the basis of thermal analysis. The following is recommenced:
(1)
Determine'if the limit on pressure fluctuation as defined by condition (1) is satisfied based on FSAR information or other source.
(2)
Determine if the limit on the number of atmospheric to service pressure cycles as defined by condition (2) is satisfied based on FSAR information or other source.
(3)
Determine if temperature difference limits for startup and shutdown and normal service as defined by condition (3), but using fluid temperature transients are satisfied based on FSAR information or other source.
If steps (1) and (2) show non-satisfaction of conditions (1) and/or (2),
tne vessel is not exempt from fatigue evaluation and the licensee should be requested to furnish same for review by NRC.
If steps (1), (2), and (3) show satisfaction of conditions (1), (2), and (3), the licensee should be asked to either demonstrate compliance with conditions (4) and (5) or furnish fatigue evaluation for review by NRC.
If steps (1) and (2) show satisfaction of conditions (1) and (2) but non-satisfaction of condition (3), the licensee should be asked to furnish evidence of compliance of conditions (3) based on metal temperature differences i
cetermined by analysis, as well as satisfaction of conditions (4) and (5), or furnish fatigue evaluation for review by NRC.
l Vessels built to ASME VIII and currently classified as Class 2 or Class 3 do not currently require evaluation for cyclic load conditions.
Welding Requirements Tne table on the following page (Weld Joint Thickness Requiring Full Radiography) provides a comparison between current and past code requirements gg A-88 M Franklin Research Center A c=en a# n. Fr.non in.oto
_ +.
i l
e when radiographic examination of butt-welded joints is mandatory. The values given are tnickness 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 VIII (1962) requirements only and currently classified as Class 1, 2, or 3 vessels may not satisfy the current radiography requirements.
2.
Primary vessels built to ASME VIII (1962) requirements may satisfy current requirements for Class 1 vessels, provided that Code Cases 1270N and 1273N were invoked.
3.
Secondary vessels built to ASME VIII (1962) requirements and currently classified as Class 2 or Class 3 may not satisfy the current radiography requirements for butt-welded joints of thickness less than 1 1/2 inch.
It is concluded that vessels built to past requirements may not satisfy current radiography requirements, depending on materials and whether or not Code Cases 1270N and 1273N were invoked.
Weld Joint Thicknesses Requiring Full Radiograchv P-No.
Current Code Requirements Material Code Class Past Code Requirements Class ification 1
2 3
ASME VIII (1962)(1) 1 0(2) 3/16 in 1 1/4 in o All joints whose material 3
0 3/16 3/4 thickness exceeds 1 1/2-in 4
0 3/16 5/8 5
0 3/16 0
o Lesser thicknesses for carbon 7
0 3/16 5/8 and low alloy steels, hign alloy 8
0 3/16 11/2 steels, and clad plate steels as 9
0 3/16 See Note 3 specified in paragraphs UCS-57, 10 0
3/16 5/8 UHA-33, UCL-35 of Reference 3 11 0
3/16 5/8 which follow.
1.
Vessels containing lethal substances shall have welded joints for materials of all thicknesses fully radiographed.
2.
All tnicKnesses require full radiograpny when "0"
is indicated.
3.
Requirements not specified for this P-No.
A-89 nd Franklin Research Center a w er the Fr.n.an in,.
Carbon and Low Alloy Steels (UCS-57 Radiographic Examination)
"In addition to the requirements in Par. UW-ll, complete radiographic examination is required for each butt-welded joint in vessel built of steel complying with Specifications SA-202, SA-203, SA-204, SA-212, SA-225, SA-299, SA-302 and SA-387 Grades A, B and C at which the plate thickness exceeds 1 in. and for each butt-welded joint in vessels built of steel complying with specifications SA-333 Grade 4, SA-350 Grade LF4, SA-353, SA-357, SA-387, Grades D and E, and SA-410 for all plate thickness.
(See Par. UCSD-19. ) "
High Alloy Steels (UHA-33 Radiographic Examination)
" (a)
The requirements for radiographing prescribed to Pars. UW-11, UW-51, UW-52 shall apply in high-alloy vessels, except as provided in (b).
(See Par. UHA-21).
(b)
Butt-welded joints in vessels constructed of materials conforming to Type 405 welded with straight chromium electrodes, and to Types 410 and 430 welded with any electrode, shall be radiographed in all thicknesses.
Butt-welded joints in vessels constructed of Type 405 materials or of Type 410 with carbon content not to exceed 0.08 percent, welded with electrodes that produce an austentic chromium-nickel weld deposit or a non-hardening nickel chromium-iron deposit snall be radiographed when the plate thickness at the welded joint exceeds 1-1/2 in.
The final radiographs of all straight chromium ferritic welds including major repairs to these welds shall be nade af ter stress-relieving has been performed.
(c)
Butt-welded joints in vessels constructed of austenitic chromium-nickel stainless steels wnich are radiographed because of the thickness requirements of Par. UW-ll, or for lesser thicknesses where the joint efficiency reflects the credit for radiographic examination of Table UW-12, shall be radiographed following post-neating if such is performed."
Clad Steels (UCL-35 Radiographic Examination)
" (a)
General Vessels or parts of vessels constructed of clad plate and those having applied corrosion-res'stant linings shall be radiographed when required by the rules in Pars. UW-ll, and UCS-57.
The plate thickness specified under these rules shall be the total plate thickness for clad construction and the base plate thickness for applied-lining construction.
(b)
Base Plate Weld with Strip Covering. When the base-plate weld in clad or lined construction is protected b'y a covering strip or sheet of corrosion-resistant material applied over the weld in the base plate to complete the cladding or lining, any radiographic examination required by the rules of Pars. UW-ll and UCS-57 may be A-90 At) Franklin Research Center 4 om or The Frer.un in w.
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I made on the completed weld in the base plate before the covering is attached.
(c)
Base Plate Weld Protected by Alloy Weld. When a layer or corrosion-resistant weld metal is used to protect the weld in the base plate from corrosion, radiographic examinations required by the rules in Pars. UW-ll and UCS-57 shall be madu as follows af ter
~
the joint, including the corrosion-resistant layer, is completed:
(1)
On any clad construction in which the total thickness of clad plate is used in the design calculation; (2)
On lined construction, and on clad construction in which the case plate thickness only is used in the design calculations, except as otherwise permitted in (d).
(d)
The required radiographic examination may be made on the completed weld in the base plate before the corrosion-resistant alloy cover weld is deposited provided all of the fallowing requirements are mets (1) The thickness of the base plate at the welded joint is not less than that required by the design calculations (See Par.
UG-16 (c) ) ;
(2)
The weld reinforcement is removed down to the surface which is to be covered, leaving it flush with the adjacent base plate, reasonably smootn, and free from undercutting; (3) The corrosion-resistant alloy weld deposit is not air-nardening ;
(4)
The completed corrosion-resistant weld deposit is examined by spot-radiography as provided in Par UH-52.
Such spot-radiographic examination is to be made only for the detection of possible cracks."
4.4 PUMPS Pumps furnished under the requirements of the Hydraulic Institute Standards (17) were designed to satisfy functional requirements.
Integrity of tne pressure boundary was not covered by this standard. The design of the pump pressure boundary should be evaluated in accordance with the current requirements of NB/NC/ND-3400 (ll.
See Sections 4.1.5 and 4.2 of this appendix for discussion of pump weld-ing requirements. The discussion in Section 4.3 of this appendix, under the subneading, " Fatigue Pequirements for Pressure Vessels," also app).ies to Class 1 pumps.
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Comparison with Requirements from Draf t ASME Code for Pumps and Valves for Nuclear Power Pumps and valves designed and constructed in accordance with the requirements of the Draf t ASME Code for Pumps and Valves for Nuclear Power, 1968 [18] would have been classified as Class I, Class II, or Class III in accordance with the requirements of Section A, B, or C, respectively.
Class I valves and pumps would be designed to the same stress allowables and fatigue limits as currently required [1]. Welding of longitudinal and girth welds and cast products would be 100% radiographed as currently required. Fracture toughness requirements, however, were not as stringent as those currently required. Paragraph 313.4 of Reference 18, " Steel Material for Low Temperature Conditions," does require that such material subjected to metal temperatures below 30*F during operation or testing be evaluated and selected to ensure adequate fracture toughness. Appendix E [18] included in tne 1970 Addenda calls for Snarpy V-notch impact testing if required by the design specifications. Energy absorption is limited to the values given in Table N-421 of the ASME B&PV Code,Section III.
Lcteral expansion must be reported although no limits are set.
These modified fracture toughness cequirements are not as stringent as those currently required.
Design requirements for valves are compatible with current requirements.
However, Class I pump design requirements were not as detailed in the past Code as they are in the current Code.
Section B of the past Code deals with Class II pumps and valver and does require full radiog aphy of welds and cast material. However, design rules are not explicit.
ine past code states "that any design method which has been demonstrz.ed to be satisfactory for specified design conditions may be used."
Section C of the past Code deals with Class III pumps and valves; the requirement permits visual examination of welds unless pipe size is 4 in or greater, in which case random magnetic particle or liquid penetrant examination is required.
In summary, Class I valves designed to the past Code would meet current requirements except possibly current fracture toughness requirements. Class I pumps designed to the past Code, however, should be evaluated against the current design rules.
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Class II and Class III pumps and valves should be evaluated to determiae if current design rules are satisfied.
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 (16],
except that the inside diameter, d, will be the larger of the basic valve body inside diameters in the regica near the welding ends. Class 1 valves may be designeu in accordance with eitner the standard design rules of NB-3530 tnrough NB-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 through the stem packing is not a cause for rejection) at not less than 1.5 times the 100*F rating rounded off to the.7 ext higher 25-psi increment as required by Reference 16, except that valves with a primary pressure rating of less than Class 150 wi?.1 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 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 [lb].
Stress limits for level B loads are based on 110% of operating limits.
Level C pressures are limited to i
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 tne pipe at 500*F.
Primary and secondary stresses for level C loads are based on C = 1.5, Q = 0, and limited to 2.25 S,.
Level D T
loads may be evaluated in accordance with Appendix F [le).
A design report for Class 1 valves will be prepared in accordance with the requirements of NB-3560 [1bl.
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Class 1 valves designed in accordance with the standard rules must satisfy the cody shape rules of NB-3544 which are intended to limit the local stress index to a maximum of 2.0.
Primary and secondary stress intensities may then be calculated by the formulas given in NB-3545.1 anG NB-3545.2 [lb],
respectively, and subject 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.
Class 2 and 3 valves are currently designed to the requirements of suoarticle NC-3500 [lc] and ND-3500 ilo], respectively. 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 provideo 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-3 512. l(b) is satisfied. A shell hydrostatic test satisfying ANSI B16.34 is required. Class 2 and 3 valve stress limits for service 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 B, C, and D shall be as shown in Table A4-12.
openings for auxiliary connections shall satisfy ANSI B16.34 and the c.
reinforcement requirements of NC-3300 and ND-3300.
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 (19]
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Table A4-12 Level B, C, and, D Service Limits for Class 2 and 3 Valves TABLE NC-3521-1 LEVEL 8, C, AND D SERVICE LIMITS Service Umit Stress Umitsw P.8 e,$ 1.1 S Level S (e. or eJ+e,$ 1.65S 1.1 e.$ 1.5 S Level C (e, or e )+e,$ 1.8 5 1.2 6
e,$ 2.0 S Level 0 (e, or o )+e,$ 2.4 5 1.5 g
NOTES:
(1) A casting quality factor of 1 shall be assumed in satisfying these stress limits.
(2)These requirements for the acceptability of valve design are not intended toassure the functional acecuacy of the valve.
(3) Design recuirements listed in this table are not appticable to valve disks, stems, seat rirgs, or ot'ier parts of the valves Wnich are contained within the Confines of the body and bonnet.
(4) These rules do not apply to safety relief valves.
(5) The maximum pressure shall not exceed the tabulated factors :isted under P tirnes the Design Pressure or times the rated pressure at the applicaole sennee temperature, t
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4.
shall be hydrestatically tested as required by Reference 19, i.e.,
1.5 times the in0*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 (16].
As an example, consider a 2500-lb valve designed in accordance with the past code [15].
Body tnickness would be based on Table 33 (19]. Comparison with current requirements may be obtained from Table 3 [16] 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 (19} Table 3 (161 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 requirements with the following possible exceptions:
1.
Fracture toughness requirements may not be satisfied. Evaluate as recommended by Section 4.1 of this appendix.
2.
Valves may not satisfy the primary, secondary and peak stress combination limits if body shape differs significantly from the rules of NB-3544 [lb].
3.
Valves may not satisfy the primary plus secondary stress limit for service level C.
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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 significanc difference 2 are found, the Licensee should be asked to provide calculations and an evaluation based on alternative rules for the ve ve in question, unless it can be shown that the valve has been subje<.:ted to level C conditions and did not have to be replaced.
Design requirements for C wass 1 valves constructed to the Draf t ASME Code.
for Pumps and Valves for Nuclear Power [18] are in compliance with requirements for current Class 1 valves, except for fracture toughness requirements (s ee discussion in Section 4.4 of this appendix).
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 Tational formula-tion as compared to the empirical basis of B16.5.
It is recom:nended 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 vsives 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 from Table A4-13 that the engineered safeguard isolation valves for the PaJ isades plant would satisfy the current standard provided that the valve material was in one of the tabulated material groups other than 1.12, 2.1, or 2.3.
ASA B16.9 (1958) (20] provides overall dimensions, tolerances, and markings for wrought carbon-steel and alloy-steel factory-made welding fittings.
It refers to ASA B31.1 for design requirements.
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Table A4-13 Allowable Working Pressore (1) for a 150 lb Standard Class Valve at 300'F Material Group Allowable Pressure (ps iq) 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 1.10 230 1.11 230 1.12 205 1.13 230 1.1) 230 2.1 205 2.2 215 2.3 175 2.4 210 2.5 225 2.6 220 2.7 220 1.
Based on ANSI B16.34 (1977) (16].
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ASA B16.10 (1957) (21] provides face-to-face and end-to-end dimensions for ferrous valves of various types and ferrous butt-welding end valves.
ASA B16.9 and ASA B16.10 do not provide design guidance for valve or f ittings. Valves and fittings built to these standards should be evaluated against tne current requirements [1,16].
I 4.6 HEAT EXCHANGERS Heat exchangers are currently designed and constructed in accordance with 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 the fol. lowing sections of the current code:
Section Shell Side 3300 Tube Side 3600 Tube Sheet 3300 Shell Flange 3200 (Class 1); Appendix XI (Class 2 and 3).
Heat exchangers designed to ASME VIII (1962) are compared as pressure vessels with current requirements in Section 4.3 of this appendix.
Heat exchangers designed to the standards of the Tubular Exchanger Manuf acturers Association (TEMA) 1959 Edition [8] require that "the individual vessels shall comply with tne 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, icentical with Tables UCS-23 and UCS-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-BW for carbon and low alloy pipe and tubes of welded manufacture; the stress values are one-fourth the specified minimum tensile streng th multiplied by a quality factor of 0.92.
Group II heat exchangers designed to TEMA (19: 9) would be governed by the code requirements of ASME VIII (1962). Comparison of ASME VIII (1962) with current requirements is as follows:
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3 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 VIII (1962) are compared with current requirements in Section 4.3 of this appendix.
4.7 STORAGE TANKS Storage tanks may currently be classified as Class 2 or Class 3 and are designeo 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 s tructures.
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 s tress 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 shoula be at least 3/16 in for tanks of nominal diameter less than 50 f t or L/4 in for tanks of 50 to 120 f t nominal diameter, but not greater than 1 1/2-in thick.
Roofs shall be designed to carry dead load plus a uniform load of at
(
least 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:
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a.
tension - for rolled steel, net section: 20 ksi; full penetration groove welds in thinner plate area:
18 kai.
b.
compression - 20 ksi where lateral deflection is prevented, or as determined from 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.C(c) (1); 20 ksi in tension and compression for unsymmetric members laterally supported at intervals no greater than 13 times the compression snape width; and for other rolled shapes, built-up members, and plate girders: 20 kai in tension and compression as determined by the buckling formulas of NC/ND-3852. 6 (c) ( 4).
d.
shearing - 13.6 kai in fillet, plug, slot, and partial penetration groove welds across throat area, 13 kai on the gross area of beam webs where the aspect ratio (h/t) is less than 60 or:
19.5 1 + (h/t) (2) 7200 0 to 15 psi Storage Tanks Storage tanks which may contain gases or liquids with vapor pressure aoove 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-3 9222.1-1 [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 from 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 Requirements Storage tanks in Group II SEP plants were designed either in accordance with A/E specifications, USAS B96.1 (1967) [9], API-650 (1964) [10], or ASME VIII (1962).
Examination of the ASME VIII (1962) allowable stress values for carbon and low alloy plate steels indicates that the values do not exceed 20 kai except for SA-353 Grade A and B, with allowaole stresses of 22.5 and 23.75 A-101 d5nklin Research Center A cumion c4 m Fr.n mon,.
l
ao k:si, respectively.
ASME VIII (1962) does not consider biaxial stress fields with associated reducticn in tensile allowables. Stress allowables for roofs in Reference 10 are the same as for current atmospheric storage tanka.
A comparison of API-650 (1964) roof design requirements, incl ding stress allowables, shows agreement with current requirements; shell material and I
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 efficiency. 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 pai cased on 0.85 joint ef ficiency, which exceeds the current 12,600 psi allowable.
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, aluminum 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-3852.7-6 for aluminum roof design. A comparison of allowables based on past and current codes is shown in the following table:
(1)
Aluminum Specified Min.
Allowable Stress Structures Material Streng th Past Current (Type of Stress)
(Temper)
TS/YS (USAS B96.1)
( ASME III (1977))
Shell (Tension) 5050 (0) 18.0 ksi/6.0 kai
- 4. 8 ka i 4.0 kai l
Shell (Tension) 6061 (T4,T6) 24.0 ksi/ -
7.2 ksi 6.0 kai Bolts (Tens ion) 6061 (T6) 18.0 kai 18.0 ksi Roof Support (Axial Compres-6061 (T6) 19.0 19.0 sion, L/r < 10)
Roof Support
( Axial Compres-6061 (T6) 20.4-20.4-sion 10 < L/r < 67) 0.135 L/r 0.113 L/r 1.
At temperatures to 100*F.
From this table, it can De concluded that:
1.
shells designed to USAS B96.1 (1967) may be overstressed by as much as 20% compared to current allowables A-102 k A Franklin Research Center A D *= w rw n.m m.u,
Cw, 2.
bolts designed to USAS B96.1 (1967) satisfy current requirements 3.
roof supports with slenderness ratios up to 10 satisfy current i
requirements 4.
roof supports with slenderness ratios between 10 and 67 more than satisfy current compression allowables by as much as 13%.
Therefore, a'.uminum alloy storage tanks built to USAS B96.1 (196'7), when evaluated against current requirements:
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 stress 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 VIII (1962) are likely to satisfy current requirements with regard to allowable tensile stress, but may not satisfy.urrent compression stress requirements.
3.
O to 15 psig tanks designec to ASME VIII (1962) 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.
4.
Atmospheric storage tank roofs designed to API-650 (1964) satisfy current stress allowables.
5.
Atmospneric 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 B96.1 (1967) may not satisfy current requirements.
A-103 dbnklin Research Center
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BASIS FOR SELECTING REQUIREMENTS MOST SIGNIFICANT TO COMPONENT INTEGRITY The selection of code requirements most significant to component integrity he-
'n based on the experience of he author and colleagues in industry, government, and academia. Codes pertaining to the design and construction of nuclear power plants have been modified and expanded. The changes reflect new
" state of the art" knowledge, new techniques of fabrication, examination, testing, and methods of achieving quality that have been " filtered" and accepted by the technical community.
It is the author's view that current codes 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, as well as more conservative requirements for fracture toughness, have been given special attention.
A A-104 du Franklin Research Center hoa on Th. ren e u.
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6.
REFERENCES 1.
ASME Doiler and Pressure Vessel Code Section III, " Nuclear Power Plant Compcaents" New York: American Society of Mechanical Engineers 1977 Edition with Addenda through Summer 1978 Division 1 and Division 2 General Requirements a.
b.
Division 1, Subsection NB, Class 1 Components Division 1,' Subsection NC, Class 2 Components c.
d.
Division 1, Subsection ND, Class 3 Components e.
Division 1, Appendices 2.
Title 10 of tne Code of Federal Regulations Revised January 1,1981 3.
ASME Boiler and Pressure Vessel Code Section VIII, "Unfired Pressure Vessels" American Society of Mechanical Engineers,1962 4.
American Standards Association
" Code for Pressure Piping" ASA B31.1-1955 American Society of Mechanical Engineers,1955 5.
ASME Boiler and Pressure Vessel Code Section I, " Power Boilers" American Society of Mechanical Engineers,1962 6.
" Quality Group Classifications and Standards for Water, Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants," Rev. 3 NRC, February 1976 7.
NUREG-75/087 Standard Review Plan, Section 3.2.2, " System Quality Group Classification" NIC, Office of Nuclear Reactor Regulation 8.
Standards of Tubular Exchanger Manufacturers Association Fourch Edition, 1959 9.
USAS B96 1-1967
" USA Stant3rd Specification for Welded Aluminum-Alloy Field-Erected Storage Tanx3" United States ef America Standards Institute, February 1967 10.
API-650
~
" Welded Steel Tanks for Oil Storage," Second Edition American Petroleum Institute, April 1964 A-105 2 Franklin Research Center a ca.on om. rm %.
11.
Final Safety Analysis Report for Consumers Power Company, Palisades Plant (3 Volumes)
USAEC Docket No. 50-255 November 5, 1968 12.
Brock, J. E.
"A Temperature Chart and Formulas Useful with USAS B31.7 Code for Thermal I
Stress in Nuclear Power Piping" Nuclear Engineering and Design, Vol.10, pp. 79-82 North Holland Publishing Co., Austria,1969 13.
United States of America Standards Institute
" Code for Pressure Piping" USAS B31.1.0-1967 American Society of Mechanical 2ngineers 14.
American National Standards Institute
" Power Piping" ANSI B31.1-1973 American Society of Mechanical Engineers, June 15, 1972 15.
United States of America Standards Institute "Drart Code for Nuclear Power Piping" USAS B31.7 American Society of Mechanical Engineers, February 1968 16.
American National Standards Institute
" Steel Valves" ANSI B16.34-1977 American Society of Mechanical Engineers,1977 17.
Hydraulic Institute Standard:
Eleventh Edition,1965 18.
Draf t ASME Code for Pumps and Valves for Nuclear Power, American Society of Mechnical Engineers, 1968 i
l 19.
American Standards Association l
" Steel Pipe Flanges and Flanged Fittings" ASA B16.5-1961 American Society of Mechanical Engineers,1961 20.
American Standards Association
" Factory-Made Wrought Steel Buttwelding Fittings" ASA B16.9-1958 American Society of Mechanical Engineers,1958 21.
American Standards Association
" Face-to-Face and End-to-End Dimensions of Ferrous Valves" i
ASA B16.10-1957 American Society of Mecnanical Engineers,1957 l
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