ML20071H202
| ML20071H202 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 04/23/1981 |
| From: | Knight J Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20071F735 | List: |
| References | |
| FOIA-93-375 NUDOCS 8105070488 | |
| Download: ML20071H202 (22) | |
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s., *. j APR 2 31981 MEMORANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing, DL FROM:
James P. Knight, Assistant Director for Components & Structures Engineering, DE
SUBJECT:
SER INPUT FOR SHOREHAM TheMechanicalEngineeringBranchhascompletedare0iewoftheShorehamFSAR through Amendment 38. The enclosed SER sections have been updated and discuss the following open issues which must be resolved prior to OL issuance:-
1.
SER Section 3.9.1.1 - The piping preoperational vibration test program description in the FSAR must include essential small bore instrument lines and safety related snubbers.
2.
SER Section 3.9.1.4 - Analysis methods under Loss of Coolant Accident loadings must address the effects of annulus pressurization on reactor vessel supports and internals. Thir was discussed with and agre6J to by the BWR Owners Group.
3.
SER Section 3.9.2.1 - The results of the fatigue analysis of ASME Class 2 & 3 downcomers and safety relief valve discharge piping must be documented and the final Mark II pool dynamic loads must be reconciled with the load values used in the structural analysis of mechanical components and equipment.
i 4.
SER Section 3.9.2.1/Section 5.2.1 The applicant must specify that his criteria for functional capability of essential piping as described in Appendix E of the Design Assessment Report is equivalent to that which has been approved by the staff for the Mark II Owners. (ie use of NED0 21985) 5.
SER Section 5.2.1 - Feedwater nozzle /sparger and control rod return nozzle modifications must be 'comitted to and described in the FSAR as specified in NUREG-0619.
6.
SER Section 5.2.1 - BWR jet pump hold down beam periodic inspection must be comitted to until a long term fix is recomended by GE and approved by the staff.
7.
SER Section 5.2.2 - Program for inservice testing of pumps and valves must be submitted (Ref. Standard Review Plan Seciton 3.9.6). Also a.comitment must be made by the applicant to leak check pressure isolation valves as specified under Section 5.2.2 of the attached SER input.
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. Resolution of the above open issues will be reported in a supplement to this SER. The Office of Inspection and Enforcement has responsibility for implementation of IE Bulletion 79-02 with respect to Pipe Support Base Plates.
We are not aware of any technical issues which remain open with respect to this item and.unless our support is requested by I&E, we consider this open item as resolved. Additionally, we have contracted with Argonne National Laboratory to perform an independent analysis of a feedwater piping loop in the Shoreham plant. This analysis will verify that this sample piping system meets the applicable ASME code requirements, and provides a check on the applicants ability to correctly model and analyze his piping systems. We will report the results of this analysis in a supplement to this SER. This SER input was provided to the Shoreham projecqanager in draft form on March 26, 1981.
y i
M.
gh Assistant Director for Components & Structures Engineering Division of Engineering
Enclosure:
As stated cc:
R. Vollmer, DE J. Knight, DE R. Tedesco, DL J. Wilson, DL B. Youngolood, DL F. Cherny, DE E. H m.inger, DE H. Ahmed. ANL U. Brammer. f R. Bosnax, DE W. Butler, DSI i
Z. Rosztoczy, DE i
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3.6 Protection Against Dynamic Effects Associated with the Postulated Rupture of Piping We have reviewed the applicant's criteria for classifying piping systems as moderate or high energy systems and for selecting the locations where pipe breaks and leakage cracks are postulated to occur, both inside and outside the containment.
The applicant's criteria provides a level of protection equiva-lent to that provided by Regulatory Guide 1.46, " Protection Against Pipe Whip Inside Containment," and Branch Technical Position MEB 3-1, " Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," for postu-lating pipe breaks inside and outside containment.
The applicant has also provided a summary of results indicating the systems and postulated pipe break 1ccations in the Shoreham Nuclear Power Station.
These results are consistent with their criteria and are acceptable.
The analytical methods and procedures used by the applicant to establish re-straint locations and pipe and restraint interaction are based on acceptable methods as identified in the Standard Review Plan Section 3.6.2.
The pipe whip restraints are designed to withstand the loads resulting from postulated pipe rupture and to remain intact to assure the protection of safety-related structures, systems, and components.
The applicant's design provides for protection against the simultaneous occur-rence of the combined loadings imposed by a safe shutdown earthquake and a concurrent single pipe rupture of the largest pipe at any one of the design break locations.
The applicant's program meets the pipe break criteria in the Standard Review Plan Section 3.6.2 and provides assurance that the following safety conditions and functions are met:
(1) The magnitude of the design basis loss-of-coolant accident cannot be aggra-vated by multiple failures of piping.
(2) The reactor emergency core cooling system can be expected to perform its intended function.
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(3) Structures, systems, and components important to safety will be I
protected.
Based on our reviews of the applicant's design, we conclude that the proposed pipe failure criteria used for systems, components, and structures which'are required to safely shut the plant down and to mitigate the consequences of those postulated piping failures, provides reasonable assurance of their ability to perform this safety function following a failure in high-or moderate-energy
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piping systems.
The applicant's criteria comply with Standard Review Plan Section 3.6.2 and satisfy the applicable portions of General Design Criterion 4.
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g 3.9 Mechanical Systems and Components 3.9.1 Dynamic Systems Analysis and Testing l
A
- 3. 9.1.1 Piping Vibration Operational Testing Program The applicant has agreed to perform a preoperational piping vibrational and dynamic effects test program on the following piping systems including their restraints, components and supports:
a.
All ASME Class 1, 2 and 3 piping systems.
b.
All high energy piping systems outside containment.
c.
All seismic Category I portions of moderate energy piping systems outside containment.
This test program, which is in accordance with Standard Review Plan Section 3.9.2, will provide assurance that the piping and piping restraints have been designed to withstand vibrational dynamic effects.
These effects are due to valve closures, pump trips, and operating modes associated with the design basis transients.
The tests, as planned, will develop loads similar to those experienced during reactor operation.
A commitment to proceed with such a program constitutes an acceptable design basis for fulfillment of the mechanical engineering requirements of General Design Criteria 14 and 15.
It is the staff's position that all essential safety-related instrumentation lines should be included in the vibration monitoring program during pre-operational or start-up testing.
We require that either a visual or instrumented inspec-tion (as appropriate) be conducted to identify any excessive vibration that will result in fatigue failure.
We require that, prior to the issuance of an operating license, the applicant provide a list of all safety-related small bore piping and instrumentation lines that will be included in the initial test vibration monitoring program.
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Our position is that the essential instrumentation lines to be inspected should include (but are not limited to) the following:
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a.
Reactor pressure vessel level indicator instrumentation lines (used for monitoring both steam and water levels).
b.
Main steam instrumentation lines for monitoring main steam flow (used to actuate main steam isolation valves during high steam flow).
j c.
Reactor core isolation cooling (RCIC) instrumentation lines on the RCIC i
steam line outside containment (used to monitor high steam flow and actuate isolation).
d.
Control rod drive lines inside containment (not normally pressurized but required for scram).
e.
High pressure coolant injection (HPCI) instrumentation lines on HPCI steam line outside containment (used to monitor high steam flow and actuate isola-tion).
- 3. 9.1. 2 Due to a long history of problems dealing with inoperable and incor-rectly installed snubbers, and due to the potential safety significance of failed snubbers in safety-related systems and components, we require that main-i tenance records for snubbers be documented with a formal test procedure to be performed as part of the piping preoperationa vibration test program.
- 3. 9.1. 3 Preoperational Test Program for Reactor Internals The reactor internals for Shoreham are substa tially the stae as those of the prototype BWR/4 plant, the James A. FitzPatrick Nuclear Power Plant.
Conse-quently, the Shoreham reactor internals will be tested in accordance with the provisions of Regulatory Guide 1.20 Revision, " Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational and Initial Startup Testing,"
for non prototype plants.
The testing will include operation of the recircula-tion system at flows up to 100% of rated flow with the internals installed, followed by a visual inspection for evidence of vibration, wear, or loose parts.
In addition to the information provided in the Final Safety Analysis Report, the applicant has referenced General Electric Company Topical Report NEDE 24057-P,
" Assessment of Reactor Internals Vibration in BWR/4 and BWR/5 Plants," which 3-4
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presents results of the prototype tests and contains additional information on the confirmatory inspection program.
Our review has established that the above topical report is acceptable for BWR/4 plants such as Shoreham.
This vibration assessment program will constitute an acceptable basis for verifying the design adequacy of these internals under test loading c.oi.oltions comparable to those that will be experienced during operation. The combination of tests and post-test inspection will provide adequate assurance that the reactor internals will, during their service lifetime, withstand the flow-induced vibrations of reactor operation without loss of structural integrity.
Satisfactory completion of the preoperational vibration assessment program constitutes an acceptable basis for demonstrating that the design of the reactor internals conforms to the recom-mendations of Regulatory Guide 1.20 and satisfies the applicable requirements of General Design Criteria 1 and 4.
- 3. 9.1. 4 Analysis Methods Under Loss of Coolant Accident Loadings The applicant has performed a dynamic system analysis of the reactor vessel, its internals, and supports for combined responses due to a postulated loss-of-coolant accident (LOCA) and safe shutdown earthquake (SSE) but did not include the response resulting from related drywell annulus pressurization effects.
We require that the effects of annulus pressurization be included in this analy-sis and the results documented in the FSAR prior to issuance of an operating license.
Stresses resulting from the combined responses of the loss-of-coolant accident, including the annulus pressurization effects and the safe shutdown earthquake, should not result in stress distributions which exceed the ASME Code Level D Service Limit (formerly referred to as the Faulted Stress Limit).
The applicant has agreed to include the effects of annulus pressurization in its analysis methods and to address specifically those issues which were raised by the staff's Question 112.21.
Unless the applicant reports results which exceed the faulted limits, our findings are as follows:
The applicant has performed a system dynamic analysis due to peak loads result-ing from postulated LOCA and SSE events. The responses due to these loads were combined by the Square Root of the Sum of Squares (SRSS) method to determine i
i the structural integrity of the reactor coolant pressure boundary and those systems which are used to limit the consequences of an accident or to shut the 3-5 J
plant down.
For the reasons stated in Section 3.9.2 of this safety evaluation report for ASME Class 2 and 3 components, we have also accepted the use of the SRSS method of combining dynamic responses for ASME Class 1 components, supports, and reactor internals for the Shoreham plant.
The applicant has evaluated all related load effects on the reactor vessel, its internals, and supports, which result from a postulated LOCA, and has demon-strated that structural integrity will be maintained and that the plant can be safely shut down.
3.9.2 ASME Code Class 2 and 3 Components 3.9.2.1 Design, Load Combinations and Stress Limits The applicant has stated in the Design Assessment Report that it has made use of the SRSS method for combining peak dynamic responses.
The use of the SRSS method for combining peak dynamic responses due to the loss-of-coolant accident and safe shutdown earthquake has been accepted by us in NUREG-0484, Rev. 1,
" Methodology for Combining Dynamic Responses." The use of the SRSS a.ethod for combining dynamic responses to other loads in BWR Mark II plants has been approved by us under certain conditions This approval was in a letter from J. R. Miller (NRC) to G. G. Sherwood (GE), dated 6/19/80, concerning our review of the GE Topical Report NEDE-24010-P, " Technical Bases for the Use of the Square Root of the Sum of the Squares (SRSS) Method for Combining Dyaamic Loads for Mark II Plants." We have determined that Shoreham meets the conditions set by this letter.
Therefore, the use of the SRSS technique for Shoreham is i
acceptable.
The requirement to perform a Class 1 fatigue analysis for the Class 2 and 3 downcomers and safety relief valve piping was discussed generically with the Mark II Owners Group and is considered generically resolved.
Each plant in this group is required to perform such an analysis.
The fatigue analysis of the ASME Class 2 and 3 downcomers and safety relief valve discharge piping is being completed by the applicant using the ASME Class 1 fatigue rules.
Unless the applicant reports results which do not meet the Code acceptance criteria, we consider this approach acceptable.
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Subsequent to the final determination of acceptable values for the Mark II pool dynamic loads, the applicant will reconcile the loads used for the structural analysis with those found acceptable.
Unless the acceptable load values exceed those used in the structural analysis, we consider this approach acceptable.
The applicant has described in its plant Design Assessment Report (DAR) the criteria which were used in evaluating the functional capability of essential piping.
Prior to issuance of an operating license we require that the appli-cant specify that this criteria, which is described in Appendix E of the
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Shoreham DAR, is equivalent to that which we have approved for the Mark II owners Group in GE Topical Report NE00-21985 " Piping Functional Capability Criteria." On receipt of this commitment, we will conclude that under emerg-ency and faulted plant conditions, ASME Code Class 1, 2, and 3 piping in essential systems (i.e., those which are required to shut the plant down and maintain it in a shutdown condition) will retain sufficient dimensional stabil-ity to deliver rated flow under the most severe loads.
Subject to resolution of the open items discussed above, our findings are as follows.
All General Electric-supplied seismic Category I pressure retaining components, j
and their supports, outside of the reactor coolant pressure boundary, are de-signed to sustain normal loads, anticipated transients, the operating basis earthquake, and the safe shutdown earthquake within stress limits consistent with Regulatory Guide 1.48, " Design Limits and Loading Combinations." The design basis combir.3tions of these loadings were applied to the design of the i
safety-related ASME Code Class 2 and 3 pressure-retaining components in sys-tems classified as seismic Category I.
This design approach provides assurance that in the event:
(a) an earthquake should occur at the site or (b) an upset, emergency, or faulted plant transient should occur during normal plant opera-tion, the resulting combined stresses imposed on the components would not exceed the allowable design stress and strain limits for the materials of construction.
Limiting the stresses under such loading combinations provides a basis for the design of the system components to withstand the most adverse combinations of loading events without gross loss of structural integrity.
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g-3.9.2.2 Operability of Active Pumps and Valves We have informed the applicant that design by analysis of active pumps and valves, to assure structural integrity, may not necessarily guarantee oper-ability under faulted condition loadings.
We have concluded that there may be a need for retesting of selected vital appurtenances to the pumps and valves at the Shoreham facility.
The necessity for such additional component testing will be established based on an inspection by the Equipment Qualification Branch of installed equipment at the plant site.
The balance of the applicant's program for component operability provides assur-ance that active ASME Class 2 and 3 components can withstand postulated seismic loads in combination with other loads without loss of structural integrity and can perform their active function (i.e., valve closure or opening or pump opera-tion) when a safe plant shutdown is to be effected, or the consequences of an accident are to be mitigated.
The applicant's component operability assurance program, plus the completion of any additional component testing which may be required, constitutes an acceptable basis for meeting the applicable require-ments of General Design Criterion 1.
The criteria used for the design and installation of overpressure protection i
devices, namely safety / relief valves, in ASME Code Class 2 systems are set forth in Regulatory Guide 1.67, " Installation of Overpressure Protection Devices." The applicant's criteria meet Regulatory Guide 1.67 and provide assurance that, under discharging conditions, the resulting stresses are not expected to exceed the allowable design stress and strain limits for the mate-rials of construction.
Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a basis for the design of the system components to withstand these loads without loss of structural integrity or impairment of their function.
The applicant's complia'nce with Regulatory Guide 1.67 constitutes an acceptable basis for meeting the applicable requirements of General Design Criterion 15.
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5.2 Integrity of Reactor Coolant Pressure Boundary
- 5. 2.1 Design of Reactor Coolant Pressure Boundary Components The applicant has provided design load combinations and associated stress and deformation limits used for reactor coolant pressure boundary Code Class 1 components and their supports.
These limits provide assurance that:
(1) in the event an earthquake should occur at the site, or (2) other system upset, i
emergency, or faulted conditions should develop, the resulting combined stresses imposed on the system components will not exceed the allowable design stress and strain limits for the materials of construction.
Limiting the stresses and strains under such loading combinations to ASME Code Class 1 limits will provide a basis for the design of the system components for the most adverse loadings postulated to occur during the service lifetime without loss of the system's structural integrity.
The applicant has identified to the staff the active components within the l
reactor coolant pressure boundary which must operate to safely shut the plant down and maintain it in a safe condition in the event of a safe shutdown earthquake or design basis accident.
The applicant has utilized an oper-ability assurance program, in addition to stress and deformation limits, to qualify active valves.
This program includes valve testing, or a combination of tests and predictive analysis, supplemented by seismic qualification of valve operator systems.
This combination of tests and analyses provides assurance that active components:
(1) will withstand the imposed loads asso-ciated with normal, upset, emergency, and faulted plant conditions without loss of structural integrity, and (2) will perform the " active" function under conditions comparable to those expected when safe plant operation or shutdown is to be effected, or the consequences of a seismic transient or of an acci-dent are to be mitigated.
Pending satisfactory resolution of additional component testing as discussed Section 3.9.2.2 o'f this report, the applicant's component operability assur-ance program constitutes an acceptable basis for implementing the requirements
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of General Design Criterion 1 as related to the operability of ASME Code Class 1 active components.
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i The criteria used in the design and mounting of the safety and relief valves of the ASME Code Class 1 systems will provide adequate assurance that, under discharging cunditions, the resulting stresses are expected not to exceed the allowable design stress and strain limits for the materials of construction.
Limiting the stresses under the loading combinations associated with the actuation of these pressure relief devices provides a basis for the design of the system components to withstand these loads without loss of structural integrity and impairment of the overpressure protection function.
The cri-teria used for the design and installation of overpressure relief devices in ASME Code Class 1 systems are consistent with Regulatory Guide 1.67,
" Installation of Overpressure Protection Devices," and Section III of the ASME Code and constitute an acceptable basis for meeting the applicable requirements of General Design Criteria 1, 2, 4, 14, and 15.
Detailed information on the new feedwater nozzle and associated sparger design l
is provided in the General Electric Topical Report NEDE-21821-A " Boiling Water Reactor Feedwater Nozzle Sparger Final Report." For the control rod drive return nozzle, General Electric recomended elimination or capping of the nozzle and rerouting the flow.
Befo;e an operating license can be issued we require that the applicant commit to the General Electric recommended modifica-tions with respect to feedwater and control rod return nozzles as specified in the above topical report and as discussed and approved by the staff in NUREG-0619, "BWR Feedwater Nozzle and Control Rod Return Line Nozzle Cracking." We will discuss the resolution of thi; issue in a supplement to this report.
i Cracking has been observed in BWR jet pump 'holddown beams.
The applicant has addressed this issue and has committed to reduce the preload from 30,000 to 25,000 pounds force as recommended by General Electric. We will require that in addition to this pre-load reduction, the applicant also commit to periodic visual and ultrasonic inspections as specified for other similar plants, and that this commitment be documented in the Jet Pump Technical Specifications.
5.2.2 Inservice Testing of Pumps and Valves We require that prior to the issuance of the operating license the applicant submit his program for inservice testing of pumps and valves.
Inservice 5-2
inspection of ASME Code Class 1, 2, and 3 rumps and valves, is to be performed in accordance with Section XI of the ASME Code and applicable Addenda as required by 10 CFR, 50.55a(g), except where specific relief has been granted.
One area of concern is the periodic leak testing of pressure isolation valves.
There are several safety systems connected to the reactor coolant pressure boundary that have design pressures below the rated reactor coolant system pressure.
In order to protect these systems from overpressure, two or more isolation valves are placed in series to form a pressure boundary interface between the high and low pressure systems.
The leaktight integrity of these valves must be insured by periodic leak testing to prevent an inter system LOCA and the possible overpressurization of the low pressure systems.
The applicant's response to Quastion 212.106 is not satisfactory.
We wili require that pressure isolation valves for the low pressure /high pressure core spray, residual heat removal, and reactor coolant isolation cooling systems be categorized as Category A or AC.
Pressure isolation valves are required to be Category A or AC and to meet the appropriate valve leak rate test requirements of IWV-3420 of Section XI of the ASME Code. The allowable leakage rate shall not exceed 1.0 gallon per minute for each valve as stated in the Standard Technical Specifications.
The applicant will be required to meet leak test requirements as specified in the Standard Technical Specifications NUREG-0123 Revision 3 - 3/4.4.3.
On receipt of the applicant's program for inservice inspection of pumps and valves and his commitment to categorize and leak test pressure isolation valves as discussed above, we will conclude that the applicant's program provides reasonable assurance that the design pressure of low pressure systems will not be exceeded and that the probability of an intersystem LOCA has been reduced in accordance with the requirements of General Design Criterion 55.
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gae 9 3.r. 2 BRANCH TECHNICAL POSITION MEB 3-1 1
POSTULATED RUPTURE LOCATIONS IN FLUID SYSTEM PIPING INSIDE AND OUTSIDE CONTAINMENT l
A.
BACKGROUND This position on pipe rupture postulation is intended to comply with the require-ments of General Design Criteria 4, of Appendix A to 10 CFR Part 50 for the design of nuclear power plant structures and components.
It is recognized that pipe rupture is a rare event which may only occur under unanticipated conditions, such as those which might be caused by possible design, construction, or opera-i I
tion errors; unanticipated loads or unanticipated corrosive environments. Our I
observation of actual piping failures have indicated that they generally occur at high stress and fatigue locations, such as at the terminal ends of a piping system at its connection to the nozzles of a component.
The rules of this position are intended to utilize the available piping design information by I
postulating pipe ruptures at locations having relatively higher potential for failure, such that an adequate and practical level of protection may be achieved.
1 8.
BRANCH TECHhICAL POSITION 1.
High-Energy Fluid Systems Piping a.
Fluid Systems Separated From Essential Systems and Components For the purpose of satisfyin the separation provisions of plant arrangement as specified in
.1.a of Branch Technical Position (BTP)
ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show the effects of postulated piping breaks at any location are isolated or physically remote from essential systems and components.1 At the designer's option, break locations as determined from B.1.c. of this position may be assumed for this purpose.
b.
Fluid System Piping in Containment Penetration Areas Breaks and cracks need not be postulated in those portions of piping from containment wall to and including the inboard or outboard isola-l tion valves provided they meet the requirements of the ASME Code SectionIII,SubarticleNE-1120andthefollowingadditionaldesIgn requirements:
(1) The following design stress and fatigue limits should not be exceeded:
1 For ASME Code,Section III, Class 1 Piping i
i (a) The maximum stress range between any two load sets (including I
the zero load set) should not exceed 2.4 S,, and should be calculated by Eq. (10) in Paragraph NB-3653, ASME Code, J
Section III, for those loads and conditions thereof for l
l 5ystems and components required to shut down the reactor and mitigate the i
1 consequences of a postulated pipe rupture without offsite power.
3.6.2-10 Rev. 1 - July 1981
which level A and level B stress limits have been specified in the system's Design Specification, including an operating basis earthquake (0BE) event transient.
The 5, is design stress intensity as defined in Article NB-3600 of the ASME Code Section III.
If the calculated maximum stress range of Eq. (10) exceeds 2.4 S,, the stress ranges calculated by both Eq. (12) and Eq. (13) in Paragraph NB-3653 should meet the limit of 2.4 Sm-(b) The cumulative usage factor should be less than 0.1.
(c) The maximum stress, as calculated by Eq. (9) in Paragraph NB-3652 under the loadings resulting from a postulated piping failure beyond these portions of piping should not exceed 2.255 except that following a failure outside containment, thepTpebetweentheoutboardisolationvalveandthefirst restraint may be permitted higher stresses provided a plastic hinge is not formed and operability of the valves with such stresses is assured in accordance with the requirements saecified in SRP Section 3.9.3.
Primary loads include those witch are deflection limited by whip restraints.
For ASME Code,Section III, Class 2 Piping (d) The maximum stress ranges as calculated by the sum of Eq. (9) and (70) in Paragraph NC-3652 ASME Code,Section III, con-sideringthoseloadsandconditionsthereofforwhichlevelA and level B stress limits have been specified in the system's Design Specification (i.e., sustained loads, occasional loads, and thermal expansion) including an OBE event should not exceed 0.8(1.2 Sh
- S ).
The S and S are allowable A
h A
stresses at maximum (hot) temperature and allowable stress range for thermal ex)ansion, respectively, as defined in Article NC-3600 of the ASME Code,Section III.
(e) The maximum stress, as calulated by Eq. (9) in Paragraph NC-3652 under the loadings resulting from a postulated piping failure of fluid system piping beyond these portions of piping should not exceed 1.8 S '
h Primary loads include those which are deflection limited.
by whip restraints.
The exceptions permitted in (c) above may also be applied provided that when the piping between the outboard isolation valve and the restraint is constructed in accordance with the Power Piping Code ANSI B31.1 (see ASB 3-1 B.2.c(4), the piping shall either be of seamless construction with full radiography of all circumferential welds, or all longitudinal and circumferential welds shall be fully radiographed.
(2) Welded attachments, for pipe supports or other purposes, to these portions of piping should be avoided except where detailed stress analyses, or tests, are performed to demonstrate compliance with the limits of B.1.b(1).
3.6.2-11 Rev. 1 - July 1981
(a) At terminal ends.2 3
(b) At intermediate locations where the maximum stress range as calculated by Eq. (10) and either (12) or (13) exceeds 2.4 S,.
(c) At intermediate locations where the cumulative usage factor exceeds 0.1.
(d)
If two intermediate locations cannot be determined by (b) and (c) above, two highest stress locations 4 based on Eq. (10) should be selected.
If the piping run has only one change i
or no change of direction, only one intermediate location should be postulated.
As a result of piping reanalysis, the highest stress locations may be shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exist:
(i) Maximum stress ranges or cumulative usage factors exceed the threshold levels in (b) or (c) above.
(ii) Achangeisrequiredinpipeparameterssuchasma,jor differences in pipe size, wall thickness, and routing.
(iii) Breaks at the new highest stress locations are signifi-cantly apart from the original locations and result in consequences to safety-related systems requiring additional safety protection.
In such conditions, the newly determined highest stress locations should be the intermediate break locations.
t 2Extremitiesofpipingrunsthatconnecttostructures, components (e.g.,
vessels, pumps, valves), or pipe anchors that act as rigid constraints to piping motion and thermal expansion.
A branch connection to a main piping run is a terminal end of the branch run, except where the branch run is classified as part of a main run in the stress analysis and is shown to have a significant effect on the main run behavior.
In piping runs which are maintained pressurized during normal plant conditions for only a portion of the run (i.e., up to the f 4st normally closed valve) a terminal end of such runs is the piping connection to this closed valve.
85 tress range under those loads and conditions thereof for which level A and level B stress limits have been specified in the system's Design Specification, including an OBE event per paragraph NB-3653 of the ASME Code,Section III.
l 45 tresses 1nder those loads and conditions thereof for which level A and level B stress limits have been specified in the System's Design Specification, including an OBE event as calculated by Eq. (9) and (10), Paragraph NC/ND-3652 of the ASME Code,Section III.
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3.6.2-13' Rev. 1 July 1981
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d.
The designer should identify each piping run he has considered to postulate the break locations required by B.1.c above.
In complex 1
systems such as those containing arrangements of headers and parallel piping running between headers, the designer should identify and include all such piping within a designated run in order to postulate l
the number of breaks required by these criteria.
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e.
With the exceptions of those portions of piping identified in B.I.b, leakage cracks should be postulated in ASME Code,Section III, Class 1 piping where the stress range by Eq. (10) of Paragraph NB-3653 exceeds
- 1. 2 S and in C1 css 2 and 3 or nonsafety class piping where the stress bythI,sumofEq.(9)and(10)ofParagraphNC/ND3652 exceeds 0.4 (1.2 Su + S ).
Nonsafety class piping which has not been evaluated a
to obtVin sTmilar stress information shall have cracks postulated at locations that result in the most severe environmental consequence.
2.
Moderate-Energy Fluid System Piping a.
Fluid Systems Separated from Essential Systems and Components For the purpose of satisfying the separation provisiens of plant arrangement as specified in B.1.a of BTP ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show that the effects of through-wall leakage cracks at any location in piping designed to seismic and nonseismic standards are isolated or physically remote from essential systems and components.
l b.
Fluid System Piping In Containment Penetration Areas Leakage cracks need not be postulated in those portions of piping from containment wall to and including the inboard or outboard isola-tion valves provided they meet the requirements of the ASME Code,Section III, Subarticle NE-1120, and are designed such that the maxi-mum stress range does not exceed 0.4 (1.2 Sh
- S ) f r ASME Code, t
A Section III, Class 2 piping.
c.
Fluid Systems In Areas Other Than Containment Penetration (1) Through-wall leakage cracks should be postulated in fluid system piping located adjacent to structures, systems or components important to safety, except-(1) where exempted by B.2.b and B.2.d, or (2) where the maximum stress range in these portions of Class 1 piping (ASME Code,Section III) is less than 1.2 5,, and Class 2 or 3 or non-safety class piping is less than 0.4 (1.2 Sh + S )*
A The cracks should be postulated to occur individually at locations that result in the maximum effects from fluid spraying and flooding, with the consequent hazards or environmental conditions developed.
(2) Through-wall leakage cracks should be postulated in fluid system piping designed to nonseismic standards as necessary to satisfy B.3.d of BTP ASB 3-1.
I i
3.6.2-15 Rev. 1 - July 1981 m
+
C.
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piping stiffness as may be demonstrated by inelastic limit analysis (e.
, a plastic hinge in the piping is not developed under loadin (4) The dynamic force of the. jet discharge at the break location should be based on the effective cross-sectional flow area of the pipe and on a calculated fluid pressure as modified by an analytically or experimentally determined thrust coefficient.
Limited pipe displacement at the break location, line restric-tions, flow ifmiters, positive pump-controlled flow, and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge.
(5) Pipe whipping should be assumed to occur in the plane defined by the piping geometry and configuration, and to initiate pipe movement in the direction of the jet reaction.
j b.
Longitudinal Pipe Breaks The following longitudinal breaks should be postulated in high-energy fluid system piping at the locations of the circumferential breaks specified in B.3.a:
(1) Longitudinal breaks in fluid system piping and branch runs should be postulated in nominal pipe sizes 4-inch and larger, except i
where the maximum stress range,4 exceeds the limits specified s
in B.1.c(1) and 8.1.c(2) but the axial stress range is at least 1.5 times the circumferential stress range.
(2) longitudinal breaks need not be postulated at:
(a) Terminal ends.
(b) At intermediate locations where the criterion for a minimum number of break locations must be satisfied.
l (3) Longitudinal breaks should be assumed to result in an axial split without pipe severance.
Splits should be oriented (but not concurrently) at two diametrically opposed points on the l
piping circumference such that the jet reactions causes out-of-plane bending of the piping configuration.
Alternatively, a si,7qle split may be assumed at the section of highest tensile stress as determined by detailed stress analysis (e.g., finite j
elemer,t analysis).
(4) The dynamic force of the fluid jet discharge should be based on-a circular or elliptical (2D x 1/2D) break area equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as determined for a circumferential break at the same location.
Line restrictions, flow limiters, positive pump-controlled flow, i
and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge.
k 3.6.2-17 Rev. 1 - July 1981'
~
~
which level A and level B stress limits have been specified in the system's Design Specification, including an operating basis earthquake (OBE) event transient.
The S,is design stress intensity as defined in Article NB-3600 of the ASME Code Section III.
If the calculated maximum stress range of Eq. (10) exceeds 2.4 S,, the stress ranges calculated by both Eq. (12) and Eq. (13) in Paragraph NB-3653 should meet the limit of 2.4 Sm-(b) The cumulative usage factor should be less than 0.1.
(c) The maximum stress, as calculated by Eq. (9) in Paragraph NB-3652 under the loadings resulting from a postulated piping failure beyond these portions of piping should not exceed 2.255 except that following a failure outside containment, thepTpebetweentheoutboardisolationvalveandthefirst restraint may be permitted higher stresses provided a plastic hinge is not formed and operability of the valves with such stresses is s sured in accordance with the requirements s)ecified in SRP Section 3.9.3.
Primary loads include those witch are deflection limited by whip restraints.
For ASME Code,Section III, Class 2 Piping (d) The maximum stress ranges as calcuTated by the sum of Eq. (9) and (10) in Paragraph NC-3652 ASME Code,Section III, con-sideringthoseloadsandconditionsthereofforwhichlevelA and level B stress limits have been specified in the system's Design Specification (i.e., sustained loads, occasional loads, and thermal expansion) including an OBE event should i
not exceed 0.8(1.2 Sh
- S ).
The S and S are allowable -
A h
A stresses at maximum (hot) temperature and allowable stress range for thermal expansion, respectively, as defined in Article NC-3600 of the ASME Code,Section III.
(e) The maximum stress as calulated by Eq. (9) in Paragraph NC-3652undertheloadingsresultingfromapostulatedpiping failure of fluid system piping beyond these portions of piping should not exceed 1.8 S
- h Primary loads include those which are deflection limited.
by whip restraints.
The exceptions permitted in (c) above may also be applied provided that when the piping between the outboard isolation valve and the restraint is constructed i
in accordance with the Power Piping Code ANSI B31.1 (see ASB 3-1 B.2.c(4), the piping shall either be of seamless construction with full radiography of all circumferential welds, or all longitudinal and circumferential welds shall be fully radiographed.
(2) Welded attachments, for pipe supports or other purposes, to these f
portions of piping should be avoided except where detailed stress analyses, or tests, are performed to demonstrate compliance with the limits of B.1.b(1).
3.6.2-11 Rev. 1 - July 1981
(3) The number of circumferential and longitudinal piping welds and branch connections should be minimized.
Where guard pipes are used, the enclosed portion of fluid system piping should be seamless construction and without circumferential welds unless specific access provisions are made to permit inservice volumetric examination of the longitudinal and circumferential welds.
(4) The length of these portions of piping should be reduced to the minimum length practical.
(5) The design of pipe anchors or restraints (e.g., connections to containment penetrations and pipe whip restraints) should not require welding directly to the outer surface of the piping (e.g., flued integrally forged pipe fittings may be used) except where such welds are 100 percent volumetrically examinable in service and a detailed stress analysis is performed to demonstrate compliance with the limits of B.1.b(1).
(6) Guard pipes provided for those portions of piping in the contain-ment penetration areas should be constructed in accordance with the rules of Class MC, Subsection NE of the ASME Code,Section III, where the guard pipe is part of the containment boundary.
In addition, the entire guard pipe assembly should be designed to meet the following requirements and tests:
(a) The design pressure and temperature should not be less than the maximum operating pressure and temperature of the enclosed pipe under normal plant conditions.
(b) The design stress limits of Paragraph NE-3131(c) should not be exceeded under the loading associated with containment design pressure and temperature in combination with the safe shutdown earthquake.
(c) Guardpipeassembliesshouldbesubjectedtoasinglepres-sure test at a pressure not less than its design pressure.
(d) Guard pipe assemblies should not prevent the access required to conduct the inservice examination specified in B.1.b.(7).
Inspection ports, if used, should not be located in that portion of the guard pipe through the annulus of dual barrier containment structures.
(7) A 100% volumetric inservice examination of all pipe welds should be conducted during each inspection interval as defined in IWA-2400, ASME Code,Section XI.
c.
Postulation of Pipe Rupture In Areas Other Than Containment Penetration (1) With the exceptions of those portions of piping identified in B.1.b, breaks in Class 1 piping (ASME Code,Section III) should be 1
postulated at the following locations in each piping and branch run:
3.6.2-12 Rev. 1 - July 1981
(a) At terminal ends.2 3
(b) At intermediate locations where the maximum stress range as calculated by Eq. (10) and either (12) or (13) exceeds 2.4 S,.
(c) At intermediate locations where the cumulative usage factor exceeds 0.1.
(d) If two intermediate locations cannot be determined by (b) and (c) above, two highest stress locations 4 based on Eq. (10) should be selected.
If the piping run has only one change or no change of direction, only one intermediate location should be postulated.
As a result of piping reanalysis, the highest stress locations may be shifted; however, the initially determined intermediate break locations need not be changed unless one of the following conditions exist:
(i) Maximum stress ranges or cumulative usage factors exceed the threshold levels in (b) or (c) above.
(ii) Achangeisrequiredinpipeparameterssuchasmajor differences in pipe size, wall thickness, and routing.
(iii) Breaks at the new highest stress locations are signifi-cantly apart from the original locations and result in consequences to safety-rel ded systems requiring additional safety protection.
In such conditions, the newly determined highest stress locations should be the intermediate break locations.
2 Extremities of piping runs that connect to structures, components (e.g.,
vessels, pumps, valves), or pipe anchors that act as rigid constraints to piping motion and thermal expansion.
A branch connection to a main piping run is a terminal end of the branch run, except where the branch run is classified as part of a main run in the stress analysis and is shown to have a significant effect on the main run behavior.
In piping runs which are maintained pressurized during normal plant conditions for only a portion of the run (i.e., up to the first normally closed valve) a terminal end of such runs is the piping connection to this closed valve.
35 tress range under those loads and conditions thereof for which level A and level B stress limits have been specified in the system's Design Specification, including an OBE event per paragraph NB-3653 of the ASME Code,Section III.
45 tresses under those loads and conditions thereof for which level A and level B stress limits have been specified in the System's Design Specification, including an OBE event as calculated by Eq. (9) and (10), Paragraph NC/ND-3652 of the ASME Code,Section III.
3.6.2-13 '
Rev. 1 - July 1981
(2) With the exceptions of those portions of piping identified in B.I.b, breaks in Class 2 and 3 piping (ASME Code,Section III) should be postulated at the following locations in those portions i
of each piping and branch run:
(a) At terminal ends.
(b) At intermediate locations selected by one of the following criteria:
r (i) At each pipe fitting (e.g., elbow, tee, cross, flange, and nonstandard fitting), welded attachment, and valve.
Where the,'iping contains no fittings, welded attach-ments, or valves, at one location at each extreme of thepipingrunadjacenttotheprotectivestructure.
(ii) At each location where the stresses 4 exceed 0.8 (1.2 Sh + S ) but at not less than two separated A
locations chosen on the basis of highest stress.5 Where the piping consists of a straight run without fittings, welded attachment, or valves, and all stresses are below 0.8 (1.2 S + 5 ), a minimum of one location chosenonthebasisbfhiheststress.
As a result of piping reanalysis, the highest stress locations may be shifted; however, the initially determined intermediate break locations may be used unless one of the appropriate conditions of B.I.c(1)(d) exist.
(3) Breaks in nonnuclear class piping should be postulated at the following locations in each piping or branch run:
(a) At terminal ends of the run if located adjacent to the protective structure.
(b) At each intermediate pipe fitting, welded attachment, and valve.
(4) Applicable to (1), (2) and (3) above:
If a structure separates a high energy line from an essential component, that separating structure should be designed to with-stand the consequences of the pipe break in the high-energy line which produces the greatest effect at the structure irrespective of the fact that the above criteria might not require such a break location to be postulated.
55 elect two locations with at least 10% difference in stress, or if stresses differ by less than 10%, two locations separated by a change of direction of i
the pipe run.
3.6.2-14 Rev. 1 - July 1981
I a
d.
The designer should identify each piping run he has considered to postulate the break locations required by B.1.c above.
In complex systems such as those containing arrangements of headers and parallel piping running between headers, the designer should identify and i
include all such piping within a designated run in order to postulate the number of breaks required by these criteria.
e.
With the exceptions of those portions of piping identified in B.I.b, leakage cracks should be postulated in ASME Code, Section IH, Class 1 piping where the stress range by Eq. (10) of Paragraph NB-3653 exceeds 1.2 S and in Class 2 and 3 or nonsafety class piping where the stress by thI, sum of Eq. (9) and (10) of Paragraph NC/ND 3652 exceeds 0.4 (1.2 Sh + S ).
Nonsafety class piping which has not been evaluated A
to obtain sTmilar stress information shall have cracks postulated at locations that result'in the most severe environmental consequence.
2.
Moderate-Energy Fluid System Piping a.
Fluid Systems Separated from Essential Systems and Components For the purpose of satisfying the separation provisions of plant arrangement as specified in B.I.a of BTP ASB 3-1, a review of the piping layout and plant arrangement drawings should clearly show that the effects of through-wall leakage cracks at any location in piping designed to seismic and nonseismic standards are isolated or physically remote from essential systems and components, b.
Fluid System Piping In Containment Penetration Areas Leakage cracks need not be postulated in those portions of piping from containment wall to and including the inboard or outboard isola-tion valves provided they meet the requirements of the ASME Code,Section III, Subarticle NE-1120, and are designed such that the maxi-mum stress range does not exceed 0.4 (1.2 Sh + S ) for ASME Code, A
Section III, Class 2 piping.
c.
Fluid Systems In Areas Other Than Containment Penetration (1) Through-wall leakage cracks should be postulated in fluid system piping located adjacent to structures, systems or components important to safety, except (1) where exempted by B.2.b and B.2.d, or (2) where the maximum stress range in these portions of Class 1 piping (ASME Code,Section III) is less than 1.2 S,, and Class 2 or 3 or non-safety class piping is less than 0.4 (1.2 Sh + S )-
A The cracks should be postulated to occur individually at locations that result in the maximum effects from fluid spraying and flooding, with the consequent hazards or environmental conditions developed.
(2) Through-wall leakage cracks should be postulated in fluid system piping designed to nonseismic standards as necessary to satisfy B.3.d of BTP ASB 3-1.
3.6.2-15 Rev. 1 - July 1981
d.
Moderate-Energy Fluid Systems in Proximity to High-Energy Fluid Systems
~
Cracks need not be postulated in moderate-energy fluid system pipi g located in an area in which a break in high-energy fluid system pi ing i
is postulated, provided such cracks would not result in more limit ng i
environmental conditions than the high-energy piping break. Where a i
postulated leakage crack in the moderate-energy fluid system piping results in more limiting environmental conditions than the break in i
proximate high-energy fluid system piping, the provisions of B.2.c should be applied.
e.
Fluid Systems Qualifying as High-Energy or Moderate-Energy Systems Through-wall leakage cracks instead of breaks may be postulated in the piping of those fluid systems that qualify as high-energy fluid systems for only short operational periods 8 but qualify as moderate-energy fluid systems for the major operational period.
3.
Type of Breaks and Leakage Cracks in Fluid System Piping a.
Circumferential Pipe Breaks The following circumferential breaks should be postulated individually in high energy fluid system piping at the locations specified in B.1
{
of this position:
(1) Circumferential breaks should be postulated in fluid system piping and branch runs exceeding a nominal pipe size of 1 inch, except where the maximum stress range '4 exceeds the limits 3
specified in B.I.c(1) and B.1.c(2) but the circumferential stress range is at least 1.5 times the axial stress range.
Instrument lines, one inch and less nominal pipe or tubing size should meet the provisions of Regulatory Guide 1.11.
(2) Where break locations are selected without the benefit of stress calculations, breaks should be postulated at the piping welds to each fitting, valve, or. welded attachment.
Alternatively,a i
single break location at the section of maximum stress range may be selected as determined by detailed stress analyses (e.g.,
finite element analyses) or tests on a pipe fitting.
(3) Circumferential breaks should be assumed to result in pipe severance and separation amounting to at least a one-diameter lateral displacement of the ruptured piping sections unless physically limited by piping restraints, structural members, or 6An operational period is considered "short" if the fraction of time that the system operates within the pressure-temperature conditions specified for high-energy fluid systems is about 2 percent of the time that the system operates as a mocerate energy fluid system (e.g., systems such as the reactor decay heat removal system qualify as moderate-energy fluid systems; however, systems such as auxiliary feedwater systems operated during PWR reactor startup, hot standby, or shutdown qualify as high-energy fluid systems).
3.6.2-16 Rev. 1 - July 1981 l
)
piping stiffness as may be demonstrated by inelastic limit analysis (e.g., a plastic hinge in the piping is not developed underloading).
(4) The dynamic force of the jet discharge at the break location j
should be based on the effective cross-sectional flow area of A oipe and on a calculated fluid pressure as modified by an anal)+.ically or experimentally determined thrust coefficient.
i limitei pipe displacement at the break location, line restric-tior,s, flow limiters, positive pump-controlled flow, and the n arm of energy reservoirs may be taken into account, as W. cable, in the reduction of jet discharge.
(5) Pipe whipping should be assumed to occur in the plane defined by the piping geometry and configuration, and to initiate pipe movement in the direction of the jet reaction.
1 b.
Longitudinal Pipe Breaks The following longitudinal breaks should be postulated in high-energy fluid system piping at the locations of the circumferential breaks l
i specified in B.3.a-(1) Longitudinal breaks in fluid system p 3fng and branch runs should be postulated in nominal pipe sizes 4-inch and larger, except t
where the maximum stress range '4 exceeds the limits specified 3
in 8.1.c(1) and B.1.c(2) but the axial stress range is at least 1.5 times the circumferential stress range.
(2) Longitudinal breaks need not be postulated at:
(a) Terminal ends.
(b) At intermediate locations where the criterion for a minimum number of break locations must be satisfied.
(3) longitudinal breaks should be assumed to result'in an axial split without pipe severance.
Splits should be oriented (but not concurrently) at two diametrically opposed points on the piping circumference such that the jet reactions causes out-of-plane bending of the piping configuration.
Alternatively, a single split may be assumed at the section of highest tensile i
stress as determined by detailed stress analysis (e.g., finite 1
element analysis).
(4) The dynamic force of the fluid jet discharge should be based on a circular or elliptical (2D x 1/20) break area equal to the effective cross-sectional flow area of the pipe at the break location and on a calculated fluid pressure modified by an analytically or experimentally determined thrust coefficient as determined for a circumferential break at the same location.
Line restrictions, flow limiters, positive pump-controlled flow, j
and the absence of energy reservoirs may be taken into account, as applicable, in the reduction of jet discharge.
3.6.2-17 Rev. 1 - July 1981 I
(5) Piping movement should be assumed to occur in the direction of the jet reaction unless limited by structural members, piping restraints, or piping stiffness as demonstrated by inelastic limit analysis, c.
Through-Wall Leakage Cracks l
The following through-wall leakage cracks should be postulated in moderate-energy fluid system piping at the locations specified in B.2 of this position:
(1) Cracks should be postulated in moderate-energy fluid system piping and branch runs exceeding a nominal pipe size of 1 inch.
These cracks should be postulated individually at locations that result in the most severe environmental consequences.
(2) Fluid flow from a crack should be based on a circular opening of area ecual to that of a rectangle one-half pipe-diameter in length anc one half pipe wall thickness in width.
(3) The flow from the crack should be assumed te result in an environ-ment that wets all unprotected components within the compartment, with consequent flooding in the compartment and communicating compartments.
Flooding effects should be determined on the basis of a conservatively estimated time period required to effect corrective actions.
C.
REFERENCES 1.
10 CFR Part 50, Appendix A and Missile Design Basis.", General Design Criterion 4, " Environmental 2.
" Boiler and Pressure Vessel Code,"Section III and XI, American Society-of Mechanical Engineers.
3.
Regulatory Guide 1.11, " Instrument Lines Penetrating Primary Reactor Containment. "
l l
l l
l 3.6.2-18 Rev. 1 - July 1981 r