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UNITED' STATES NUCLEAR REGULATORY COMMISSION W ASHINGTON. D. C. 2058 5 Docket No. STil 50-447 JUL 1 1975 General Electric Company
,9ECE!VED ATTH: Mr. Ivan F. Stuart REI0$$2o.,-, t it.C.
Manager, Safety and Licensing 175 Curtner Avenue JUL e e m San Jose, California 95125 i
g Gentlemen:
Ud II12'l 2 b 4'5,6 J
l il We have reviewed your application for Preliminary Design Approval (PDA) of the GESSAR-230 standard nuclear steam supply system (NSSS). The scope of our review has included only the interfaces between the liSSS and the balance-of-plant (DOP), site, and options. All other aspects
.of the NSSS were evaluated by the liRC staff as part of the review of the GESSAR-238 nuclear island.
To maintain our proposed licensing review schedule which is currently i
under review by !!RC management, we require complete responses to the requests contained in Enclosure 1 by July 25, 1975.
If you cannot meet tne required date, please notify us, since it is likely that the overall schedule for this project will have to be extenced.
t Please contact us if you require additional discussion or clarification regarding the information requested.
Sincerely,
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7 gg II ydL4 f
I go J hn F. Stolz, Chief
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ight Water Reactbrs Branch 2-1 l
Division of Reactor Licensing l
Enclosure:
Request for Additional Information--Round 1 cc: Mr. W. Gilbert, ilanager Mr. L. Gifford, flanager Safety and Standards Regulatory Operations Unit General Electric Company General Electric Company 175 Curtner Avenue 4720 tlontgomery Lane San Jose, California 95114 Bethesda, itaryland 20014 k.hk bhke m.
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t ENCLOSURE 1 REQUESTS FOR ADDITIONAL INFORMATION ROUND'l.
GENERAL ELECTRIC. COMPANY GESSAR-238 NSSS DOCKET NO. STN 50 447 s
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000-1 000.0 REACTOR PROJECTS 000.1 The methods used in the SSAR to identify and describe the safety related interface-information are not completely acceptable i
because the information is not presented in a systematic manner that permits us to determine that all of this information has been identified and defined.
Interface information must be adequately identified to serve as guidance for designers of the interfacing systems and NRC staff reviewers concerned with the review of interfacing systems.
Therefore, we wi11 require the following additional information:
(1) Identification of the physical location of all safety l
related interfaces between the GESSAR-238 NSSS and the balance of plant and site. Where optional items are identified'within the scope of design, interface information must be identified for both cases in which the optional item may or may not be included in the utility application that references GESSAR-238.
(2) Cross-reference of interface safety design criteria to item (1).
a (3) Cross-reference of interface safety design information to item (1).
For item (1), we require that all of the safety relai:ed interfaces for each system within your scope of design,' be numerically identified on a drawing, schematic, diagram, etc.
In most cases this could be accomplished using figures which already exist in the SSAR.
For item (2), we require a cross-reference of the interface design criteria that must be specified to assure that the NSSS (and options) can be designed to meet safety requirements when combined with a BOP design and site parameters. ' These criteria should include, where appropriate, the following:
Regulations General Design Criteria (Appendix A to 10 CFR 50)
WRC Staff Branch Technical Problems Industry Codes and Standards 9
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000-2 These interface safety design criteria should be appropriately tabulated with adequate cross-referencing to the figure which identifies the interface in the SSAR.
For item (3), we require a cross-reference of the interface safety design information on which you based the safety design 1
of the system. The following types of safety design information for steady state, transient,-and accident conditions should be included where appropriate:.
Electrical i
Environmental (External) 1 Environmental (Internal) e.g. water quality, conductivity Hydraulic Metallurgical
. Operating Requirements j
Physical QualityGroup. Classification Radiological Seismic Classification i
Structural Thermal The tabulation of all such safety design information should be
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adequately cross-referenced to the figure which identifies the interface.
If the design of the system has not progressed to the stage where this safety design information can be quantified, we require that you identify all of the safety related design parameters and,specify when these parameters will be quantified.
Additional questions relating to specific interfaces are contained in subsequent sections.
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010-1 010.0 Effluent' Treatment Systems 010.1 For plants thet do not employ the GESSAR optional liquid radwasto (11.1) system, provide a tabic giving the flow rates and concentrations of radioactive materials in streams from the NSSS that must be processed in the balance of plant liquid radwaste system to meet the requirements of General Design Criterion,60 of, Appendix A to 10 CFR Part 50.
010.2 Provide a table giving the flow rates and concentrations of radio-III*I) active materials in streams from the NSSS that must be handled in the balance of plant solid waste system, for normal operation, antici-to meet the pated op'erational occurrences, and the design condition:
a requirements -of General Design Criterion 63 to 10 CFR Part 50.
l 010.3 Provide a' tabic giving Icakage rates'of radioactive fluids from NSSS (1 I) components into areas serviced by balance of plant ventilation systems foi normal operations, anticipated operational occurrences and the design conditions so that we may evaluate the capability of the effluent monitoring systems to meet the requirements of General Design Criterion 64 of Appendix A to 10 CFR Part 50.
Anticipated
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operational occurrences should include refueling, relief valve vent-ing, and isolation scrams.
i 010.4 The scope of the NSSS as defined in,GESSAR is inconsistent with O*) Amendment' 1 to WASil-1341, since GESS AR includes the Main Condenser Of f-Gas System as part of the NSSS.
To be consistent with WASn-1341,-
the Main Condenser Off-Gas System should be designated an optional system and the interface information with the NSSS should be provided for a balance-of-plant of f-gas system, i
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.,i 020-1 020.0-AUXILIARY AND POWER CONVERSION SYSTEMS 020.1 To assist in the review of the future utility-applicant dockets which reference GESSAR-NSSS, describe and discuss the guidance or other assistance that will be given the utility-applicant in assigning descriptive titles to BOP portions of systems inter-facing with GESSAR-NSSS systems to enhance the clarity of the application.
020.2 To assist in the review of future utility-applicant dockets which (1.11) reference GESSAR-NSSS, describe and discuss the guidance or other assistance th't will be given the utility-applicants in the a
preparation of the BOP application to assure a common usage of symbols and abbreviations, throughout its P&ID's and text, to be consistent thereby avoiding confusion. Provide revised table of contents for GESSAR which updates the present application.
020.3 Many GESSAR-238-NSSS system P&ID's contain notes on them that impose requirements on others in the design and/or construction of the facility.
Identify.those applicable notes which involve functional performance, and/or interface requirements.
Include a reference to l
these requirements in Section 1.11.
020.4 Some sections of the SAR have the statement " applicability to be (1.11) confirmed by the applicant." Proyide in tabular form the major interface requirements associated with each such statement and provide a means of locating this information by reference in Section 1.11.
020.5 In regards to the NRC staff review conducted on the GESSAR-Nuclear (1.11)
Island (GESSAR-NI) concept, (a) identify those questions and responses that fully apply to the GESSAR-NSSS concept (b) identify those questions and responses that do not fully apply to the GESSAR-NSSS concept and (c) for those portions of questions and responses that do not fully fall in either of.the above categories, identify those portions of the questions and responses that do apply to the GESSAR-NSSS concept.
020.6 In regards to GESSAR-NSSS and BOP containment boundary interfaces (1.11) provide the following in Section 1.11 to assist in the review process:
(a) provide exterior drawings of the outer surface of the contain-ment which locates all interfaces between GESSAR-NSSS and BOP, including all fluid (gas and liquid) lines.
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t 020-2 (b) identify each interface and assign a unique identification number to it for use by the utility-applicant in his B0P application and the reviewer when a BOP review is being j
carried out.
(c) in those cases where a number of different items have been grouped under one of the above interfaces, identify the particular items at the interface.
(d) for each of the above interface items reference the relevant portions of the SAR that describes the conditions and postulated adverse circumstances which establish the GESSAR-NSSS critical combination of interface parameters.
020.7 Expand the NSSS-B0P Safety Interface Matrix Table 1.11-0 so as to (1.11) enable the NRC staff (by appropriately referenced sections of GESSAR) to readily go directly to the section of GESSAR that (a) describes and discusses the interfaces and the allowable adverse combinations of significant parameters (without having to resort to the intervenin.g Tables 1.11.1 through 1.11.22), (b) provide and/or identify the s
appropriate GESSAR-NSSS P&ID diagrams showing the HSSS interfaces (c) assign a unique number to each 1nterface in order to enable the i
NRC staff of the BOP (future utility-applicants are.also expected to use the identical interf ace numbers) to verify that the BOP interfa'ce information is compatible with that presented in GESSAR-NSSS, (d) reference the appropriate section numbers of GESSAR-NSSS j
which describe and/or discuss the classification, codes, standards, j
regulatory guides, and other requirements that have been complied
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with in the GESSAR-NSSS portion of the application (e) reference the appropriate s ection of GESSAR-NSSS which presents the system j
design criteria and bases, the classification, codes, standards and regulatory guides that the utility-applicant BOP portion of the system must comply with and (f) for each system reference the appropriate section numbers of GESSAR-NSSS which presents the system criteria, bases and interf ace requirements that must be met by the utility-applicant BOP portion of the systems.
1 The above requested information should include the Main Steam Isolation Valve Main Condenser and Evacuation System and the four
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operational modes of the RRR System.
020.8 Table 1.11-0, the NSSS-BOP Safety. Interface Matrix, does not identify j
(1.11) any specific long term cooldown interface requirements with the BOP assuming a LOCA and/or a loss of offsite power.
Identify, describe and discuss the combinations of minimum interface requirements that must be met by the BOP in order to provide assurance that the GESSAR-NSSS portion of the plant will function properly during the above assumed long term cooldown conditions.
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k 020.9 Table 1.11-0, GESSAR/NSSS-BOP Safety Interface Matrix, indicates that (1.11) a number of options are available with the GE supplied USSS systems.
Tabulate all options that are being offered the utility-applicant.
In regards to these options provide and/or indicate the appropriate sections of GESSAR-NSSS that describes and discusses the following:
(a) the interfaces, with and without options (b) identify and quantify all parameters at each interface that must be controlled or limited if proper operation is to be attained, with and without each individual option.
(c) identify the GESSAR-NSSS P&ID's showing each interface and the unique identification number associated with each interface (that 1
the utility applicant is also expected to use to clarify and j
expedite the BOP review) with and without each individual option.
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i (d) the classification of the GESSAR-NSSS portions of the systems with 4
and without each available option.
i (e) compliance of the GESSAR-NSSS supplied portions of the systems with
' the General Design Criteria and Regulatory Guides, with and without each individual option.
(f) the system criteria and BOP requirements that will be imposed on the applicant, with and without each individual option.
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1 020.10 Table 1.11.22 indicates that Sections 5.5 through and including the (1.11) first two paragraphs of 5.9.3 fall within the scope of GESSAR-238-NSSS.
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Since the table of contents and text stops at Section 5.5.14.4, j
provide additienal information to clarify the scope of GESSAR-238-NSSS.
j 020.11 Table 1.11.22 indicates that for a GESSAR-NSSS facility that the (1.11) systems and equipment described in Section 9.4 and 9.5 will be the utility-applicant's responsibility under the BOP.
Since these systems contribute to the overall safety of the facility and in some cases are directly related to the functional capability I
of the GESSAR-NSSS portions of the plant, identify, describe and discuss
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the design criteria, and' bases that will be imposed on the utility-applicant portions of the plant for each of the systems described under Section 9.4 aid 9.5.
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020-4 1
i 020.12 Revise Table 1.11.22 so that it..will:
1 (1.11)
(a)' only consist of the GESSAR-NSSS without options (b) identify every system or subsystem, as appropriate with a title as well as the corresponding section number (c) indicate where in the SAR and/or provide the information regarding the interface data 020.13 The tabulation in Table 3.2.1, Equipment Classification, does not (1.11) reflect the latest amendments. To readily permit the reviewer to establish what portions are within the scope of GESSAR-NSSS or BOP,*
jirovide a revised and updated Table 3.2.1 which provides this information.
020.14 The footnote on page 8.3-2, Sec' tion 8.3.1.1.6 Standby Power Supply (1.11) indicates that the description of the three Class IE onsite diesel generators will be described in the utility-applicant's SAR.
The footnote on page 9.2-14a Table 9.2.2, Equipment Requiring Essential Service Water To Ensure Safe Plant Shutdown, indicat'es that procuretient of. Division I and II diesel generators is the responsibility of the s
applicant.
From Table 1.11.22 it would appear that the HPCS Diesel Generator Set is within the GESSAR-SSSS scope of supply. Further, 4
Table 1.11.22 indicates that all of equipment described in Section 9.5 falls within the BOP scope of supply including the DG Fuel Oil Storage and Transfer System, DG Cooling Water System, DG Starting Air System and DG Lubrication System. Provide clarification as to what portions the three DG units and associated subsystems are within the responsibility and scope of supply by GESSAR-NSSS by appropriate j
modifications to the above referenced sections of the SAR, and provide the required interfaces and associated requirements for the BOP design.
020.15 Section 9.4.7 Diesel Generator Building Ventilation System is (1.11) independent of the combustion air.
The tables in Section 1.11 indicate the HPCS diesel generator is within the GESSAR-238-SSSS scope of supply.
For the standard design, describe and discuss the capability of the HPCS diesel engine combustion intake air cleaning system in relation to the maximum potential dust or other' particulate burden that may exist at any given utility-applicant site during emergency conditions and extreme natural phenomena such as high winds, tornadoes, and haboobs.
020.16 Table 1.11.9, RHR-NSSS/15vP Interfaces (Item 9-R-1) indicates that the (1.11)
RHR provides backup makeup water for the fuel pool cooling system.
Modify the P&ID and verify that a seismic Category I spent fuel pool makeup water system will be provided that does not share any component or portion of the RHR system.
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020-5 020.17 Table 1.11.22 page 1.11-44 indicates that Section 9.3.4 is applicable (1.11) to the BOP (i.e. not within the scope of CESSAR-NSSS).
Section 9.3.4 indicates the Standby Liquid Control System falls under this heading and is described in Subsection 4.2.3.
Table 1.11.2 page 1.11-34 shows
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all of Section 4 except Table 4.4.4 falls under the scope of GESSAR-NSSS. Describe, discuss and clarify what. portions of the Standby Liquid Control System fall within (a) the responsibility of the utility-applicant supplied equipment (b) the responsibility of GESSAR-MSSS supplied equipment.
In addition describe and discuss design criteria, and bases imposed on the utility-epplicant portion of the system as well as acceptable range of significant variables that could occur at the interface.
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020.18 In regards to the 'overall stability of the integrated GESSAR-NSSS (1.11) system and the BOP power conversion system, using the overall stability model block diagram of GESSAR-NI review (Question 18.21.3 second example), identify, describe and discuss all interfaces between GESSAR-NSSS and the BOP power conversion system.
In addition, the response should elaborate on the interface information presented.
in Section 1.11 for those power conversion syste= interfaces associated with engineered safeguards s.sfety functions and/or those associated with attaining.and maintaining a safe and orderly plant shutdown by referencing the appropriate sections of GESSAR that describes and each interface.
discusses the limiting set of acceptable parameters at Table 1.11.17 and Table 1.11-0 appear to indicate that portions of 020.19 (1.11) the GESSAR-238-NSSS supplied equip =ent will impose dif fering interfacing requirements in the BOP systems depending on the temperature of BOP cooling water.
Identify, tabulate, describe and discuss each situation where the equipment within the scope of supply i
of GESSAR-238-NSSS will be altered as a result of site parameters in order to maintain overall compatibility between it and the BOP.
020.20 Provide a tabulation of all GESSAR-238-NSSS high and moderate energy interface with the BOP.
For each such system piping systems that (3.6) provide an analysis which demonstrates the means and adequacy by which this protection (within the GESSAR-238-NSSS system plus options scope of supply) will be afforded in accordance with that set forth in the enclosed Branch Technical Position (BTP APCSB 3-1),
For the portions of systems outside the scope of CESSAR-238-NSSS, identi'" the interfaces and limiting parameters at the interface in order that the BOP portion will be compatible, without modifications,-
to GESSAR-238-NSSS.
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.1 020-6 020.21 Amendment 27 page 9.2-4a, Station Service Water System states'that i
(9.2.1) the entire essential service water system must conform to the Safety
. Design Bases (Section 9.2.1.1.1.1).
Further, the revised page deleted a sentence stating "The applicant is also responsible for certain of the control and instrumentation components within the Reactor Island which are more appropriately associated with his part of,the system."
Explain and clarify the following:
(a) the reason for the deletion I
(b) is it correct to assume, because of'the deletion, that General Electric assumes the responsibility for all portions of the control and instrumentation for the entire essential service water system?
If the above assumption is not correct, clearly set.forth what j
control and instrumentation will General Electric be responsible j
for. Describe and discuss the extent of GE responsibilities j
regarding the measurements of process parace'ers located outside their areas of responsibility which must operate for the GESSAR.__
'NSSS supplied systems to operate properly.
(c) In reference to the statement that the entire essential service water system must conform to the Safety Design Bases Section
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9.2.1.1.1.1, provide an explanation so that the reviewer will be able to readily establish the party responsible for areas judged to be in noncompliance during a review which utilizes GESSAR-NSSS as a source of information to support the application.
.020.22 Section 5.5.9.3, Amendment 18, states the NSSS portion of the main (9.3.6) steam piping extends to the second isolation' valve.
Further, that a remote manual motor operated gate valve, having a closing time of approximately 120 seconds, is located downstream of the second isolation valve.
j Since this third valve is instrumental in serving as the outermost valve of~the outboard Main Steam Isolation Valve Leakage Control System (MSIVLCS) described in Se.ction 9.3.6, describe and discuss the design criteria, bases and interface requirement imposed on the GESSAR-NSSS utility-applicant for this third valve in order to attain the Icakage control capabilities described in Section 9.3.6.
Further, describe and discuss with the aid of P&ID's, the scope and compliance of GESSAR-NSSS supplied MSIVLCS with Attachment 020-0.
At each interface, including the Standby Gas ' Treatment System (SGTS),
identify the significant variables and the acceptable range of each variable.
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020.23 In regards to the fire protection system to be provided for the (9.5.1) facility by the utility-applicant, identify, describe and discuss those portions of.the system described in CESSAR-238-NI Section 9.5.1 that will be imposed as requirements on the utility-applicant's BOP in order to assure the safety related design objectives and functional capabilities of GESSAR-238-NSSS supplied equipment.
020.24 In regards to the overhead handling systems, Table 1.11 ' indicates (9.1.4) the spent fuel cask crane will be the utility-applicant's responsi-t bility.
Section'9.1.4.2.2.3 states that "All drawings and design calculations by the cask crane contractor will be submitted to
'GE for review and approval." Describe, discuss and demonstrate that guidance'to be given to the cpplicant will be censistent with the guidelines listed in the. attached APCS Branch Technical Position 9-1, " Overhead Handling Systems for Nuclear Power Plants."
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's January 1975 BRANCH TECHNICAL POSITION APCSB 3-1 l
PROTECTION ACAINST POSTULATED PIPING FATLURCS IN FLUID SYSTEMS OUTSIDC CONTAINNEh7 A.
BACKCROUhT)
General Design Criterion 4, " Environmental and Missile Desihn Baucc,"
of Appendix A* to 10 CFR Part 50, "Genern1 Design Criterin for Nuclear Power Plahts," requires that systems and conponents important to safety" shall be appropriately protected against dynamic eficcts, including the effects of missiles, pipe. chipping, and discharging fluids, that may,
result from equipment failures and from events and conditions outside the nucient power unit."
Cuidance on acceptable design approaches to noct l
General Design Critorien 4 for exisitng plants and for',plents for td ich applications for construction pert:its were then under review ucs provide::
in letters to appliennts and licensees from A.-Cisabusso, Deputy Director of Licensing for Reactor Projects, cost of which were dated in December The guidance document from these letters is attached as Appendix B l
1972.
to this position.
Similar interim guidance for new p1sucs was provided in a letter to applicants, prospective applicants, reactor vendors, and i
architect 2cngineers frca J. F. O' Leary, Direct,or of Licensing, dated 1
July 12, 1973. This document is attached as Appendix C to this position.
Cu'idance is available.for protection against pipe whipping and other effects of postulated fluid system piping failures (e.g., a hren!c or rupture resulting in a loss-of-coolant accident) of systems and cc. pencats impertsnt
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to safety located within primary reactor containment. As an exampJc, this problem is addressed by Regulatory Guide 1.46, " Protection Against Pipe Whip Inside Containment."
Reviews of nuclear power plant designs have indicated that the functional or structural integrity of systems and componente required for safe shetouwn of the reactor and maintenance of the cold shutdown conditions could be endangered by fluid system piping failurcs at locations outreide containment.
The staff has evolved an acceptable approach for the design including the arrangement, of fluid systems located outside of containment. This approach i
is set forth in this position and in the companion branch technical position (BTP) attached to Standard Review plan 3.6.2, BTP MEE 3-1.
It ;is the intent of this design approach that postulated piping failures in fluid systems should not cause a loss of function of c.wantial safety-related systems and that nucicar plants should be able to withstand postulated failurcs of any fluid systc= piping outside containment, tahir.g into account the direct results of such a failure and the further failure of any singic ac ive compopent, with acceptable offsite consequences.
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. 2-The detailed provisions of the position below and of BTP l!EB 3-1 are intended to impicment this intent with due consideration of the special nature of certain dual purpose systems and the need to define and to limit to a finite number the types and locations of piping failures to be analyzed. :Although various measures for the protection of safety-related systems and components are outlined in this position, the preferred method of protection is baced upon separation and 1. solation by plant arrangement.
B.
BRAllCH TECliNICAL POSITION 1.
Plant Arrangement Protection of' essential systems and components /
1 against,ppstuinted piping f ailgos, in 11pjl or moderate erprey fluid systens thr.t 1
operate during normal plant conditicus,and that are located outside of containment should be provided by one of the following plant l
arrangement considerations:
Plant arrangements should separcte fluid system pipint; from a.
essential sys tems and corponents.
Separation should be achieved by plant physical layouts that provide sufficient distances betucen esscntial systens and ec me: ents and iluid system piping such that the eff ects of any costuh:ted *,.inz failure therein (e.g., pipe whip, jet fapingement, and the environmental conditions resulting frem the escape of contained fluids as appropriate to hich or nederate-encrev f]uid systcm piping) cannot impair the integrity or operability of essentini systems and connenents, b.
Fluid system piping or portions thereof not satisfying the provisions of B.1.a should be enclosed within structures or compartments designed to protect nearby essential systems and
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Alternatively, essential systems a,nd cecoonents, components.
may be enclosed within structures or components deaigned to withstand.the effects of postulated riping failures in necrby fluid systen.
Plant arrangements or r,ystem features that do not satisfy c.
the provisions of either B.1.a or B.1.b should be 1,fmfted to those for which the above provisions are impractical because of the stage of design or construction of tha plant';
because tae plant design is based upcn
- hat of an earlier plant accepted by staff as a base plans under the Coricsion's
,l standardization and replication policy or for other substantive rease I
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1 See, Appendix A for definitions of underlined phrases.
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such as particular design features of the fluid systeus.
Such cauen may arise, for exEmple, (1) at interconnections between fluid systems and essential systems and compo_ngn,tyi, or (2) in fluid systems havina, dual functions (i.e., ruquired to operate durit:g normal plant conditions as uc11 as to shut down the ' reactor).
In these cases, redundant design features that are separated or otherwisc protected from postutgted I
piping failures, or additional protection, should be provi. led so that the effects of Epstulated pining,failurcs are shown by the analyses and guidelines of L.3 to be ac.ceptabic.
Additional protection may be provided by restraints and barriers or by design -
a ing or testing essential systens and corpenergs,to withetand the effects ascociated with postulated nininc failures.
2.
Desinn Features a.
Essential svstems and compon'ents should be designed to nect the scismic design requirements of negulatory Guide 1.29.
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Protective structures or compartments, fluid systen piping re-straints, and other protective t.easures should be designqd in accordance with the following:
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(1)
Protectit'e structures or compartments needed to implement B.1 shoulo ce designed to seismic Cater,ory'I requirements.
The protective structures should be designed to withotand i
the effects of.a postulated nipinc failer.3 (i.e., pipe whip, jet impingement pressurization of ccmpartuents, unter spray, and flooding, as appropriate) in combination eith loadings associated uith the operating basis carthquche and safe shutdown earthquake within the respective design load limits for structures.
Piping restrainte, if used,. cay be taken into account to linit effects of the postulated y,iair J
failure.
(2) Hinh-energy fluid system piping restraints and protective measures should be designed such that a postulated brer.k in one pipe cannot, in turn, lead to rupture of other nearby pipes or components if the secondary rupture could result in consequences that would be considered unacceptable for the initial postulated break. An unrestrained whipping
, pipe should be considered capable of (a? rupturin.g inpacted pipes of sma'.ler nominal. pipe sizes and (b) developing through-vall :.eakage crack's in larger nominal pipe sizes with thinner wall thicknesses, e:: cept where experimental or analytical data for the expected range of impact energics
' demonstrate the capability to withstand the impact without failure.
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Fluid sygt_e.,m piping between containment icolation valves shonid c.
meet the following design provisions:
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(1) Portions of finfd system piping between isolatten valves or singic barrier containment structures (includ 'a any rigid conntetion to the containment penetration) that i
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connect, on a cuntinuous or int ermittent bases, to the l
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reactor coolant pressure boundary, or the steam and
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feedwater systens of pWR plants, should be designed to i
the stress limits specified in E.1.b er D.2.b of tranch
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Technical Position..(BTP) MEB 3-1, attached to Standard Review Plan 3.6.2.
These portions of hi:;h-ener_pv fluid systym piping should be provided with pipe whip restraints. that are capabic of resisting bonding and torsional moments produced by a postulated piping failure cither.upttream or de.enttream of' the containment isolation valves. The restraints should be located reasonably close to the containr.cnt isolation valves and should be designed to withstand the loadings
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resulting from a postulated ninir; failure bayend these-portions of piping so that ncither isolation' valve opercbflity 4
a nor the leaktight integrJty of the containment 1/111 be in-l paired.
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(2) Tortions of fluid,_sveten piping betroen iso'Istion valves.of i
. dual barrier centicinment structures should also meet the design provisionsief D.2.c. (1).
In addition', those porticas of piping that pass through the containment annulus, and 4
whose postulated failure could affact the lenktight intc3rity l
of the containment structure or result in pressurization of I
the containment annulus'beyond design limits should be provided with an enclosing protective structure.
For the purpose of estab11ching the design parameters l
(i.e., pressure, temperature) of the enclosing proccetive structure, a full flow area opening should be assumed in that portion of piping within the enclosing structure and taking into account vent arcas, if provided, in th.e enclosing structure. Where guard pipes for individual proccas pipes are used as an enclosing protective structurc, such guard pipes should be desig:'ed to meet the requirements specified in B.1.b. (6) of LTP. " :, 3-1.
(3) _ Terminal ends of the piping runs extending beyond these portions of hiph-energy fluid system piping should be con-n.sidered to originate at a point feraediately outside or beyond these required pipe whip fbstraints Jocated inside and outsido containment.
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d.
Inservice examination and related design provisions should be in accordance with the following:
(1) The protective measures, structurcs, and guard pipes should not prevent the access required to conduct the
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inservice cy.aminations specified in the ASME Doller and Pressure Vessel Code,Section XI, Division 1, " Rules for Inspection and Testing of Components in Light-Water.
Cooled Plants."
(2)
For those portions of high energy fluid ji ggtjerj piping identified in D.2.c, the extent'of inservice exaufnations completed during cach inspet: tion intervn1 (1RA-2400 AS:E Code,Section XI) should provide 100 percent volumetric examination of circumferential 'and longitudinsi pipe welds within the boundary of these portions of pipin;.
(3)
For those portions of fluid systems piping enclosed in guard pipes, inopection ports should be provided in guard pipes to permit the required ennnination of circum-ferential pipe welds.
Inspection ports should not be ic.
cated in that portion of the guard pipe passing through the s
annulus of dual barrier containment structurcs.
(4) The areas subjea:to examination shoul? be defined.in accordance with Examination Categorics C-F and C-G for Class 2 pipin; ucids L1 Tabic IWC-2520.
3.
Analysas and Effcets of rectuinted pipinn Failures To show that th'c plant arrangement and design features provide a.
the necercary protection of esseutfal syster.s and ecm7enenii.
piping fcilures should be postuinted in accordance uith DTP FEB 3-1, attached to Stcndard Review Plan 3.6.2.
In applying the provisions of ETP >ED 3-1, ench longitudinal or circumferen-tial break in hip.h-emernv fluid system piping er leaknee crach in moderate energy _ fluid s'. stem piping should be considcred separately as a single postulated initial event occurring during normal plant conditions. An analysis should be made of the effects of cach such event.taking into account the provisions of BTP MEB 3-1 and of the system and component operability con-siderations of B.3.b belew.
The. effects of each po.<t" Int..d ninine failurn thould be shoun to result in ofisite consequences within the guidelines of 10 CFR Part 100 and to meet the provisions of B.3.c and d below.
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_p b.
In analyzing the effcets of nostulated of_p, Inn failures, the following assumptions should be uade with regard to the operability of systems and components:
(1) Offsito po:<cr should be assumed to be unavailabic if a trip of the turbino-generator system or reactor protection systen is a direct consequence of the postulated piping failuro.
(2) A sinnie act.1ve compone,nt failure should be assumed in
. systc=s used to mitigate consequences of the ppstuinted piping failure,and to shut down tl.e reactor, e>:wpt au noted in IJ. ') b. (3) below.
The sin :le active comonent f_n_ilure is assumed to occur in addition to thc hur.elnted pipinn f_s,1,1ure and any other direct consequences of the piping' failure, such as unit trip and loss of offsite power.
(S) Where the postulated piping failure is assumed to occur in enc of two er more redundant trains of a dual-purpose moderate-energy essential system, i.e., one required to operate during normal plant conditions as uell as to shut
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down the reactor and nitigate the consequences of the piping failure, singic failures of corponents in the o h.r train or trains of that system only n2cd not be cssu=cd provided the systen is designed to seismic Category I standards, is powered from both offsita and cnsite sources, and is constructed, operated, and inspected to quality assurance, testing, and inservice inspection stt.ndards appropriate for nuclear safety syctems. D ampics of rystc s
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that mny, in some pinat detiens, qualify as dual-purpesa
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essentini systems include service water systems, component cooling systems, and residual heat renoval syntets.
(4) All available systems, includinr, those actuated by operator actions, may be ecployed i.o mitigate the consequences of a po,stulated pining f ailure.
In judging the availability cf systems, account should be taken of the postulated failure and its direct consequences such as unit trip and loss of offsite power, and of the assumed sint:1e active.corpon:m-l f ailure and its direct consequences.
The fencibility of carrying out operator actions should be judged on the'besis i
of ample time and adt.quate access to equipment bo.'.c'; available j
for the proposed actions.
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-7 The effects'of a postul_a,tji,_p_fpine failure, including d
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environmental conditionn resu'ltinr, from the escarc of contained fluids, should not preclude habitability of the l
control room or access to surrounding areas important to the safe control of reactor operations needed to cope with I
the. consequences of the piping' failure.
l d.
A postulated failure of piping not dcsip.ned to scismic Category I standards should not result in any loss of capability of er.sential avstems and components to uithotand the further effects of any single act tvo co non.:nt foiliire and still perform all functions required to shut do'.:n the reactor and mitigate the conccquences of the portu1ated ofping failurg. -
4.
Impicmentation Designs for plants for which' construction permit applications are a.
tendered af ter July 1,1975 should conform to the provisions of this position.
I b.
Design of plants for which construction permit applications tre tendered after July 1, 1973 and before July 1, 1975 should ceaform to the provisions of cither (a) the letter of July 12, 1973 frem J. F. O' Leary, Appendix C to this positioni or (b) this position,'
at the option of the applicants.
Design of plants for which construction permit applications were c.
tendered before July '1,1973 and for which opercting Jicenses c.re i.
issued after July 1, 1975 should follow the guidance provided in the December 1972 letter frem A. Cicabusso Appendix B to this position. Analyses, made in' conformance with B.3 of this pesicion, should be presented as part of the operating license applier.:i.:n for these plants to demonstrate that acceptable protection casiest the effects of piping failurcs cutside centainncnt has beo", pro-vided. Alternately, this position ray be used in its entirety as an acceptable basis for this finding.
For plants in this category for which construction permits cre not issued as of February 1,1975, a ccamitment by the' applicant to either (a) follow the guidance of Appendix B and submit B.3 analyses with the plant final safety analysis report (FSAiU, or (b) confotm the. plant des 1;n to the. provisions of this position, should previde an acceptable basis for issuance of the const uction permit with l
regard to effects of piping failurec outside containment.
1 d.
Designs of plants for which operating licenses are issued before July 1, 1975 are considered acceptable with regard to effects of piping failures outside containment on the basis of the analyses
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I made and measures taken by appifcants and licensees in response to the December 1972 1ctter from A. Ciambuc80, Deputy Director for Reactor Projects, Directorate of
. Licensing, and the staff revicu and acceptance of these 7
analyscs and peasures.
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For plants in this category for which the staff review and acceptance of protection against the effects of piping failurcs outside containment is not substantially complete as of February 1,1975 a comnit:aent by the npplicant to carry out analyses according to E.3 of this position, to submit them for staff revice, and to carry out any system modifications, found necessary before extended operation of the plant at pouer icvels above cac-half the licen'sc power icvel, should provide an acceptabic basis for issuance of the opercting license.
C.
REFERENCES 1.
10 CFR Part 50, Appendix A, General Design Criterion 4, " Environmental and Missile Design Bases."
a 2.
Regulatory Guide 1.46, " Protection Agains't Pipe tihip Inside Containment."
3.
Letter from A. Giambusso, Deputy Director for Reactor Projects, Direc-torate of Liccasing, to applicants and licensect., December 1972, and attachment entitled "Cencral Information Required for Consideration of the Effects of a Piping System Lreak Outsida Containment."
The corrected attachment is Appendix B to this positien.
5.
ASME Boiler and Pressure Vessel Code, Sections III and XI, American Society of Mechanical Engineers.
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1 APPENDIX A Branch Technical Position APCSB 3-1 DEFINITIONS
~ Essential Systems and Components.
Systems and components required to shut down the reactor and mitigate the consequences of a postu]sted
_ piping failure, without offsite power.
Fluid Systems. Uit:h. and noderate eneriv f 3nid cvstens that are subject to the postulation of piping failures outside containment accinct which.
protection of essential systems and components is needed.
High-F.nprev F3uid Sys tets.
Fluid systees that, during notral. slant conditions, are either in operation'or maintained pressurized under conditions where either or both of the following are met:
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a.
Maximum operating temperature exceeds 200'F, or
,b. Maximum operating pressure excceds 275 'psic.
a Moderate-Energv Fluid Systems.
Fluid systems that, during norg,n1 nirmt cond_itions, are either in operation or caintained pressurized (abova atmospheric pressure) under conditions 1;harc both of the. following arc met:
Maximum operating temperature is 200*F or less, and a.
b.
Maxir.um operating pressure is 275 psig or less.
Normal Plant Coaditions.
Plant operating cenditiens during reactor startup, operation at power hot standby, or recctor cooldoun to cold shutdoun condition.
_ Upset Plant Conditions. Plant operating conditions during system treincients that may occur with noderate frequency during plant service life and arc anticipated operatLnal occurrences, but not during system testing.
Postulated Pioine Failures. Longitudinal and circumferential breaks in high-energv fluid fysten piping and through-vall leakage crac1s in' modern e-encrev tluid cystem piping postulated according to the provisiens of BTP - FID 3-1.
S and S. Allowabic stresses at maximum (hot) temperature and allousbic h
3 stress range for thermal expansion, respectively, as defined in Article NC-3600 of the A$ME Code, Section Ill.
S.
Design stress intensity.as defined in Article NB-3600 of the ASME Code,Section III.
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_Sinnlo Activo Component Failulq. { malfunction or loss of function of a component of clectrical or fluid Yiyst(ns. The failure of an active l
^ function as a result of mechanicalcomponent of a fluid ' system is cons
, hydraulic, pneumatic or electrien1 l
malfunction, but not'the loss of component structural integrity.
direct consequ' ences of a single active compnnont failure are considered The to be part of the single failure.>
Terminal Eng.
Extremities of piping runs that connect to structures,
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componcuts - (c.c., vessel pumps,.. valves), or-pipe anchors. tha t act as rigid constraints to piping thet;Taal expansion..
A branch connection to-a main piping run is a_termfuni cnd of the branch run.
i In piping runs schich are maintained pressurized durJng normal conditions for_only a portion of the:run (1.'e.,'up to the firstf) o n,t, closed valve) a tort..Inal end of such runs is the piping cennection tonor:. ally this closed valve.
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APPE! MIX U BRANCH TECHNICAL POSITION APCSB 3-1
, This appondix consists of the attachment to the letters sent by A. Ciambusco, Deputy Director for ucactor Projects, Directorate of Licensing, in D ccmber 1972 to appliennts and licencecs on the subject of postulated piping failurcs outside containacnt. The I
attachnent provided guidance en tensures to be enhen and on information to be subnitted. An errata shee.t for the attachment i ns sent in Jar. nary 1973 to recipients of the original letters.
The attachuent as given l
here has been corrected to include the errata.
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_ General Information Required for Consic!eratien of the Effects of a Pipine. System Lreak Outside Cont:ir-ant i
Tha following is a gene:1 list of infor:gation required for AEC rehic t of the effects of a piping cystem breck outsida contein.ent, includin a the doubic-ended rupture of the largest pipe in the main stena cnd feedwater systcas, and for AEC revicu of any proposed design c!.nnges that nay be fouted nece.tsary.
Since piping layouts are substantially different from plant to plant, applicants and 11canseco shonid deter:.;ipe on an individeal plant baric the applicability of each of the in11c : int, items for inclusion in their submittals.
i 1.
The systems (or portions of systems) for schich protectiva against pipe whip is required should be identified.
Protection from pipe whip need not be provided if any of the follening conditiens will exist:
(a)
Both of the following piping. system ccaditions are met:
(1) the service temperature is Icss than 200'F; and (2) the design pressure is 275 psig or less; or j
(b) The piping is physically ceparated (or isolated) frca structures, i
systems, or compenents important to safety by protcetive barriers, or restrained from whipping by plar.t design features, such as concrete encacement; or
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(c)
Following;n single break, the unrestrained pipe moves.ent of either end of the ruptured pipo'in any poscibiu direction about a plastic hinge formed at the nearest pipe whip re-straint cannot impact any structure, system or component
- 1mportant to.ssfety; or (d) The internal encrcy levell! associated with the whipping pipe can be demonstrated to be insufficient to impair the safety function of any structure, system or component to an unacceptable Icvol.
-l 2.
Design basic break locations should be selected in accordance with
~
the follouing. pipe ubip protection criteria: however, where pipes -
carrying hich enery fluids are routed in the. vicinity of structures and sys tems necessary for safe shutdown of the nuclear pint.t, supplemental prbtection of those structures and systens M:all be provided to cope with the environmental effects-(including the effects of jet inpingenent) of a single postulated open crack at the most adverse locatien(s) with regard to there essential-structurc:
and systees, the length of the crach being chosen not.to exceed the critical crach size. The critical. crack site is taken to be 1/2.the pipe diameter in length and 1/2 the wall thickne:s in vidth.
+
a The criteria used to determine the design basis piping brech locaciens in the piping systems should be equivalent to the fo11cvir.g:
(a) ASME Section III Code Class *I piping"j breaks chould be n
postulated t cur at the follouing locations in each piping run3# or branch run:
(1) The' terminal ends; (2) Any interme'diate locations between terr.inal ends where the primary plus secondary stress intensities S (circun-ferential or longitudinal,1). derived on an clastica.11y I
calculated basis under the loadings associated with one-halfsafegputdo'..ncarthquakyandoperationciplcnt conditione-e:cceeds 2.0 S d for ferritic steel, and 2.4 S,for austenitic stceS; (3) Anyinternediatelocationsbetweynterminalendswhec the cumulative usage f actor (U)E derived from the piping
[fatiguc analysis and based on all normal, upset, and l
testing plant conditions exceeds 0.1; and Footnotes are c911ected at the end of this appendix.
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~3-i (4) At intermediate Iqcations in additien to those determined by (1) and (2) above, selected on a reasonabic basis as neccusary to provide protection.
At a minimum, there should be two intermediate loca-l tions for cach piping run or branch run.
(b) ASME Section III Code Class 2 and 3 piping breaks should.be postulated to occur at'the following locations in ecch piping run or branch run:
(1) The terminal ends; (2) Any intermediate locations between terninal cuds uhere either circumferential or longitudinal stresres dertved on an classically calculated basis under the leadings associated with scisnic events ard operational plant conditions execed 0.8 (Sh+ A) frt expansien stresses
. exceed 0.8 S ; and (3)
Intermediate locations in addition to these determined by (2) above, selected on a reasonabic basis as nee. ass,ary to provide protection. As a minimu:a, there should be two a
intermediate locations for each piping run or branch run.
3.
The criterla used to determine the pipe break orientation at th bresk.locatiens as specified under.(2) ubove should ha c"uivalent to the follcwing:
(a) Longitudinal 8/ breaks in pipir.g runs and branch runo, 4 inches nominal pipe size and larger, and/or (b)
Circumferential9/
- breaks in pipin3 runs and branch runs execeding 1 inch nominal pipe size.
l 4.
A summary should be provided of the dyn:mic cnalyses applicable to the design of Category I piping and associated supports which deternine i
the resulting loadings as a result of a postulated pipe break incle igg:
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'(a) The locations and number of design basis breahs on which tite dynamic analyses are based.
(b) The postulated rupture orientation, such as a circumferentini and/or longitudinal breah(s),.for each postulated design basis break location.
I (c) A description of the forcing functions used for the pipe whip dynamic analyses including the direction, rise tico, nagnitude, duration,,and initial conditions that adequately represent the jet stream dynamics and the system pressure difference.
e
4-(d) Diagrams of mathematical models used for the d namic analysia.
3 (e) A summary of.the analyscs which derrorstrates that unrestrained motion of ruptured lines will not dauano to.an unacceptr.ble degrec, structure, syst:ms, or components i:nportant to safety, such as the centrol roor.i.
5.
A description should be provided of the measures, an applicabic, to protect against pipe whip, blowdown jet, and reactive forces including:
(a) Pipe restraint design to provent chip icipact; (b)
Protective provisions for structures, syste:ns, and ccmponents l
required for safety again.st pipe whip and blowde.cn jet cnd reactive forces:
(c) Separation of redundant features; (d)
Provisions to ceparate physically piping and other componer.ts l
of redundant features; and (e) ' A description ~ of the typical' pipe rhip restrei:.ts ::nd a st= mary of number and location of all restraints in each system.
6.
The procedures that will be used to evalunte the structurcl adequx:y of Category I structurcs and to design nc.i scismic Catc,ery I structures should be provided including:
(a) The method of evaluating stresses, e.g., the workine stress methed end/or the ultimate strength t ethod that vi>.1 be uscd; (b) The allowable design stresses and/or strains; and (c) The load factors and the load combinations.
7.
The struttural design loads, including the pressure and temperature transients, the dead, live, and equipment loads, and the pipe end equipment static, thermal, and dyncmic reactions should ba provided.
8.
Seismic Category I structural elements such as floore, intericr eclls, exterior valls, building penetrations, and the buildings e.s a whole should be analyced for eventual reversalof loads due to tr.c postulated
'ccident.
a 9.
If new openings are to be provided in existing structures, the
. capabilities of the audified structures to carry the design loads should be demonstrated.
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- 10. Verification that failure of any structure,11ncluding non-ceismic
. Category I structurcs. - caused by,' thu accident, vill not caew
' failure of any other structure in a manner to advursely affect:
(a) Mitigation of the consequences of the accidents; and J
i (b); Capability' to bring the unit (s) to a cold shutdown condition.
I
- 11. l Verification that rupture of a p.f pe carrying hikh ci[crgy fluid will not directly or. indirectly recult in:
(a). Loss of required redundancy-in any portion of the pro:ection
- system (as defined in IEEE-Std 279), Class IE cic-etric syst.n'.
I (as defined in IEEE-Std. 30S), engineered snfety fenterc equipment,. cabic penetrations, cr: their intercennectir.g cables ~
required t'o mitigate the c, consequences of that accident.and 4
place the reactor (s) in a cold shutdewn conditior.; or (b)'-Environ =cutally. induced failurcs causc'd by a leak or rupture-1 of the, pipe '.thich'vould not.of itself result in protective action but does disabic protection functions..In this.reanr3.
1
, a loss of redundancy is permitted'; but a icss of, function is a-not permitted. Tor such situations, plant shutdcrn is required.
- 12. Assurance should be* provided that the control roc: will be hr. bite 31e and its equipment functicnal after a staan ifnc or'feedeater liac
. break or that the cepabili:y fer shu:devn and coci;orn on'the' unit (s) will be availabic in another habitable crce.
- 13. Environmental qualification should be demonstrated by test for thct i
electrical equipment required to.functien in th: stec=-air en.irennant resulting from a high crcrsy line brcak.
The inforection required for our review should include the'fc11ewins:
(a)
Identification of all electrical equipment necessary to teet-requirements of (11) above. 'The time after the accident in which they are' required to opertte should be given.
(b) The test conditions and the results of test d:ta showins that 1
the systens will pe.rform their intended function in the environment resulting fren the postulated accident and time interval of the accident.
Environmental conditions used'for the tests should be selected from a conservative evaluctica of' accident conditions.
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4 (c). Tlue results of a study of steam systems identIfyfug locations where barriufs vill. be requirud to pr event steam jet impingement from disabling a protection systet.
=The design criteria for the barriors should be neated and the capability of the equipment to survive within the 6
protected. environment should be described.
-(d) An evaluation of the capability for safety-re]ated electrien1 equipment in the control room to function in the environ-L ment that may exis't following a pipe break eccident shcold be provided.
Environmental conditions used for the ovuluntion should be sniccted'from conservative calculations of accident conditions.
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(e) An evaluation to assure that the onsite pouer distribution system and onsite s'oprecs (dicscis and bottorica) uill remain operable throushout the event.
- 14. Design disgrams and drawings of the steam and feeduatcr Jines includins branch lines showing the routing from ccatrine. cat to thy turbine building should be provided.
The drawings'should sliou ele,vations and include the location relatite to the piping runn cf s
safety reisted equipment Jncluding ventilation-equipe.ent, intakes, and ducts.
i.
- 15. A discussion should be provided of the potentici for flonijng of safety-related equipment (in the event of fcilure of a feedvater line or any other line carrying high energy fluid.
- 16. A description should be provided of the quality control cnd innptction prograns that will be requiredior have been utili:cd for pipinn systccs outside containnsnt.
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If leak detcetion equip ent is to be used in the propeced modifientions.
8 a discussion of its capabilities should be provided.
18.'
A setmary should be provided of the energency proceduros that would j
be folloued after a pipe break accident, including'the automat!c ::nd' j
manual operations required to place the reactor unft(s) a a cold i
shutdown ennditich. The estincted times follouina the accident for j
all equipment and personnel operational actions should be included in the' procedure summary.
-19.
A description shonid be provided of the seismic and quality cincrif f ea-tion of the high energy fluid piping systems including the steam and feeduater piping that run near structurcs, systems,or components impor tant to safety.
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- 20. A description should be provided of the assumptions, methods, and results of analyses, including secam generator blowdown, used to calculatc~ the piessure and temperature transients in compart.ments, pipe tunnels, intermediate buildings, and the turbine building following a pipe rupture in these arcas.
The equipment ar.sumed to function in the analyses should be identified and the capability of systems required to function to meet a j
single active component failure should be described.
I
- 21. A description should be provided of the octheds or analyses per-formed to demonstrate that there will be no adverac cifcets on the primary and/or secondary containment structurcs due to a pipe rupture outside these structures.
1 1/ The internal fluid energy 1cvel associated with the pipe break reaction may take into account any line restrictions (e.g., flow liniter) be:::cen the pressure source and break location, and the effects of either single-ended or double-ended flow conditions, as applicable.
The i
energy level in a whipping pipe may be considered as insuf ficient to rupture an impacted pipe of equal or greater nominal pipe size and equal or heavier vall thickness.
2/ Piping is a pressure-rctaining component censisting of straight or curved pipe and pipe fittings (e.g., elbows, tecs, and reducers).
I 3/ A piping run interconnects components such as pressure vcssels, pe as, and rigidly fi:.:cd valves that may act to restrain pipe novcment beyend that required for design thermal displacceent.
A branch run differs from a piping run only in that it originates at a pipind interrectic.n, as a branch of the main pipe run.
4/ Operational plant conditions include normal reactor operation, upset conditions (e.g., anticipated operational occurrences) and testing conditions.
5/ S is the design stress intensity'as specified in Section III of the
" ASME Boiler and Pressure Vessel Code, " Nuclear Plant Couponents."
6/. U is the cumulative usage factor as specified in Section III of the AMSE Boiler and Pressure Vessel Code, "Nuc1 car Power Plant Cotponents "
2/ S is the stress calculated by the rules of NC-3600 and ND-3600 for h Class 2 and 3 components, respectively, of the ASME Code section III Winter 1972 Addenda.
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S is the allouabic stress range for expansion etresc cniculated A by the rules of NC-3600 of the /S:!E Lode,Section III, ur the USA Standard Code for Pressure Piping ANSI B31.1.0-1967.
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Longitudinal breaks are parallel to the pipe axis and oriented at any point nrnund the pipe ' circumference. The break area is-equal to the effective cross-sectional flow area upst.rcam of the j
break location.
Dynamic forces resulting from such breaks arc i
assuced to cause lateral pipe movements in the' direction normal to the pipe axis.
9/
Circumferential breaks are perpendicular to the pipe axis, and'the break area is equivalent to the internal' cross-scetional area of, i
.the ruptured pipe.
Dynamic forces rcsulting from such brenhs are assteed to sep,rnto the piping axially, and cause whipping in' nny direction normal to the pipe axis.
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~' APPENDIX C r'I BRANCH TECHNICAL' POSITION APCSB 3-1
- This appendi':: consists of the letter and attachment sont by
}.
J. F. O' Leary,' Director of Licensing, to applicants, reactor vend, ors, 7
and architect-engineers on the subject. of postulated piping failurcs l
outside containment..The letter was dated July 12, 1973.
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Late last year, the Atomic Energy l Commission's Regulatory staff requested.
those utilities that operato nucicar -pover plants, have applied fer
. operating licenses, or have plants' chose construction p.ernit review was essentially complete, to assess the -effects end consequence.s of a I[
3 postulated rupture of piping containing high-energy fluids and located
' outsido of the containment struct ure.
These'.recuests vera issued by i
Mr; A.: Gia:busso, L'oputy Director of Reactor Proj ects, Directorate of l '~
Licensing, in letters, most of which were dated in Decembcr 1972.
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Because these plants were either in operation or in advanced stages of engineering design and construction, the request included guidance for corrective codifications that could be impic=ented by in-site measures.
Such modifications included relocation or rerouting of piping, installatica of i=pingerent barriers and encapsuistion cleeves around high stressed piping regicas, provisions for venting of cc=partments. subject to pressurization, addition of piping restraints, and strengthening of structural cenponents of buildings.
From our review of responses submitted to the Regulatory stcff. and
' from discussions with architect-engineering firms., we have learned that some 'of' these-organizations have inferred that the criteria contained in Mr. A. Giacbusso's letter pertaining. to corrective codificcticas for plants in advanced stages of construction and operation are applicable for the. design of high-energy fluid systems outside the containment in new designs of nuclear power plants.
It was not our intent that the criteria for corrective piant modifications be applied to new power plants that are in the initial design stages. We believe that a more direct approach, involving a rearrangccent of the physical plant layout with a view to relocation of essential safety systems and components is appropriate for the new plants.
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For the precent, pending iccunnce of a pinnned AEC Roculatory cuide -
" Protection Againut Postulatud Events and Accidents Out: side ContMn-ment," on acceptabic impicnenration of Criterion 4 of the Connission's Cencral Design Criteria listed in Appendix A of 10 Crn Part 50, as applied to new plants with respect to the design of ceructures, systems, and components important to safety and located outside of containment is as follows:
I.
PIPINC SYSTr::s CONTAINI;:0111Cli-EtiETCY FLUIDS
- Dt'nl::C 1:On"AL REACTOR OP N rION (a)
The piping systems are isolated by adequnto physical separdtion and repotely located from safety systems and co..ponents that -
are required to shut down the reactor sarely and maintain the plant in a cold shutdown condition.
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(b) Where isolatica by rerote location is impracticabic, systens j
containing high-energy fluids, or portions of the systems, are enclosed uithin the structures suitably designed to protect adjoining safety systems and components required to shut do.rn l
the reactor safely and maintain the plant in a cold shutdo'.m d
conditien frc= postulated ~ pipe fai. lures within the enclostre.
(c)
Where both isolation by remote location (as specified'in I.a)
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and enclosure in protective structures (as specified in I.bT
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areicpracticable, systems,containinghich-energyfluids,ch portions of the systems, are provided with restrainta and protective ceasures such that the operability and inte.3rit~y of structures, safety syste=s and cc ponents that are requ1 red to shut down the reactor safely and =aintain the plant in a cold shutdown condition are not inpaired.
l (d)
Protective enclosures for the piping systens containing high-i energy fluids are designed as Seistic Category I structures to withstand the combined effects'of a postulated pipe break, the dynamic effects of pipe shipping, the jet impingement forces, n
and the compartment pressurization as a consequence of discharging fluids in combbation with the specified seismic event of the Safe Shutdown Earthquake and normal operating Icads.
-(e)
Piping systems containing high-energy fluids are designed so that the effects of a siny'.e postulated pipe break cannc.,
in turn, cause fai]ures of other pipes or components with unacceptable coascquences.
- Refer to Appeadix A for identificationn of high-energy fluid systems.
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- - - - - - - - - - - - - - - - - - - - - - - - ' - - ~ - ^ - - ' - " ' - - - ~ ~ - - ~ ~ - - - - - -
In addition, any systems, or portions of systetan that are designed to mitigate the consequence of a postulated pipe failure, and to place the reactor in the cold chutdown condition, are provided with design featurce that vill casure the performance of their safety function, assuming a singic active component failure.
(f)
Fe.r a postulated pipe failure, the escape of steam, water, and heat frem structures enclosing the high-energy fluid containing piping does not preclude:
- 1) the accessibility I to surrounding areas important to the safe control of reactor operations, 2) the habitability of the centrol room, 3) the ability of inctrtrr.entatien, electric pouer supplies, and
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components. and controls to Initictc, actuate and cc-picte a safety actien.
In this regard, a loss of redundancy is permissibic but not the loss of function.
(g).The criteria for determination of postulated brcck locations are contained in the attached Appendix A, " Criteria for l
Determination of Postulated Pipe Parcak or Leakc:;c Locations in Fluid Piping Systcr.s Outside Containments."
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PIPISC SYSTEMS CONTAINI:'G MODERATE'-E"ERGY FLUIDS
- DU;INC P.EACTOR OPE!L'. TIM (a) Piping systems containing noderate-ener:;y fluids are designr.J '
to~ comply uith the criterin applied to high-energy fluid piping syste:.s cs listed under I., above, e::cepe that ti 2 piping is pestulated to develop e limited-siac through-un11 leakage crach ins: cad of a pipe break.
(b)
For ecch postulated leakage, design nessures are included that provide protection frc the effects of the resulting unter spray and ficoding to the scmc extent required to satisfy criterion 1(c).
(c) The criteria for determination of postulated Icahage locations are contained in Appendi:: A.
I The measures taken for the protcetion of structures, syste as and lcomponentsinportant to safety should not preclude the conduct of inservice examinations of AS:E Class 2 and 3 pressure-retaining components as required by the rules.of ASME Leller and Pressure Vessel Code - Section~ XI, " Inservice Inspection of Nuclear Power Plant Co:iponents."
- Refer to Appendix.A for identification of modcrate-energy fluid systems.
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_. _ _ _ _ _ _ _ _ _ - - - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ - - - - ' - - ' -
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J Although co,pliance with the dec.lgn criteria listed above should j
be accoraplished by plant arrangement and layouts utilizing the separation concept to the extent practicchic, special consideration will be necensary to provide adequ:.te protection where interconnection is unavoidable between high-energy fluid cor.tcining piping and piping of syste=s important to safety.
We' are prcpered to discuss with you these guidelines,for the desi;;n l
i of new nuclear powcr plants with regard to protection required against l
postulated breshs of hir,h an,d moderate ennrgy piping outuide of j
containment, particularly fo'r thone plants with constrt:ction permit applications currently under consideration.
Sincerely, l
John F. O' Leary, Director
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l Directorate of Licensing l
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Enclosure:
Appendix A
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1 APPENDIX A (To.J. F. O' Leary Letter Dated July 12, 1973)
CRITERIA FOR D TERMIt* TION OF p0STULATED EREAR AND LCM' AGE LOCA'lI_0MS 1:;
'HIGik A!1D MON: RATE ENERGY Fl.UID PlPIMC SYS1DW OUTSIDE OF CD:TrAI:0g STRUCTURE & 0/
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High-Dercy Fluid Systems 4
1.
For piping systens that by picn't arrapacment and lnyout are isolatedbyrerotelocationfro7structurcs,g; stems,and components inportant to safety 2, pipe breako-need not be postulated provided the requirements of A.4 cre catisfied.
2.
For' piping systc=s that are enclosed in suitably designed concrete structures or compartments to protect structures, systems, and cc ponents.1=portant to safety, pipe brcaks-should be postulated at the following locations in each piping
'or branch run within the protective structure:
a.
tL2 terminal ends /
9 of the piping or branch run (except
. as exempted by the provi'sions of A.I), if located within the protective structure pr compartment, and b.
each fitting (i.e., elbou, tee, cross, nen-standard fitting), and a minimum of one break selected in cach piping or branch run c.
within the protective structure or compartment at a locction that results in the maximum loading from the i= pact of the postulated ruptured pipe and jet discharge force on wc11, floor, and roof of the structure or compartncnt, including internal pressurization, and taking into account any piping restraints provided to limit pipe motions.
3.
For portions of piping.cysteEs that can neither be isolated as specified in A.1, nor enclosed in protective structur'es as specified in A.2, pipe breaks should be postulated at the fo11 cuing locations.n each piping or branch run within the confines of the structures or compartments that enclose or adjoin areas containing systems and components important to safety:
- l 1Footnotes are co11ceted at the end of this Appendix.
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e 2-the tctninal end.e/ of piping. or branch run (except as o
a.
exempted by A.4), if located within the boundary of the confining structure or each compartment within the' structure; and l
b.
any intermediate location within the boundary of cl.e confining structure or each compartment within the structure where the stresser under the leadings associated with eventsb a7 operational plant conditions.O c::specified sei d
cced 0.0
)
(Sh+S) r, in lieu f these calculated stress-related A
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locations, at each fitting (i.e., elbow, tee, crors,' non-s tandard j
fitting); and I
c.
a minicun of two separated locations within the boundary of the O
confining structure or bach cenpartment within the etructure 1n j
piping or branch runs exceeding twenty pipe diancters in length; I
a mininum of one location in' pipir; or branch runs twenty pl.pc-1 diameters or less in icngth except that no intermediate locatiens U
need to be postulated in branch runs that are three pipe-dlatecars f
or less in length.
Intermediate break locations should be selected such thr.t the maxice t pipe whip and jet inpingecont vill
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. result, assuming fur this purpose an unrestrained ruptured pipc.
I 4.
Fo c those portions of the piping passing through prir.ary contain cat j
pc actrations and extending to the first outside isolation valve, p!.po breaks n2ed not be postulated provided such pipitig is conservatively reinforced and restrained beyond the valvo such that, in the event cf j
a postulated pipe breah outside containment, the trans:.itted pipe j
loads vill neither impair the operability of the valve nor the integrity of the piping or tbc containment per.atrap.cn.
(A termin11 cnd of such piping is considered to originacc at thic restraint loc;tien.
- g. -
B.
Iloderate-Energv Fluid Svste s g
f 1.
For piping systems that by plant arrangement and laycut are isolated and physically separated and remotely 1ccated from systems and components 1.portant to safety, through-wall Icakage cracks need not be postulated.
I 2.
For piping systems'that are located in the sane area as high-energy fluid systems which, by the criteria of A.1 to A.3 have postulated pipe break locations, through-wall leakage cracks:need not be g
postuinted.
3.. For piping systems that are located in areas containing systems and components inportant to safety, but there non hi":h-energy fluid syr.tc=s are present, through-wall leakage crccks should be postulated at the most adverse location to determine the protection needed to withstand the effects of the resulting water spray and flooding.
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- i C.. Size and Types of pipe Breaks _and Crackn 1..'The following types of' breaks should be postulated at the locations specified by the criteria listed under A.- High-Ene rgy.
' Fluid Systems:
i Longitudinal breaks in piping runs and branch runs.with a.
- nominal pipe sizes of 4 inches and. larger.
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circumferentici breaks in piping runs and branch runs exceeding a nominal pipe size of 1 inch.
2.
The following leakage cracks are postulated at the locations specified"by the criteria listed undcr B.-
Systems:
Moderate-Energy Fluid' through-vall Icakage cracks in piping and branch runs enceeding-a.
a nominal pipe size of 1 inch, where the crach opening !.s assuned as 1/2 the pipe diameter in length and 1/2 the pipa vall thickness in width.
a FOOTNOTES 1/
H1gh-energy systems inclede those systems where cither of the following conditions are mat:
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a) the maximum opera' ting temperature exceeds 200*T, and-b) the maximum operatir:g pressure exceeds 275 psig.
Jbf Moderate energy systchs include thoso systems where'both of the following conditions are met:
a) the maximum operating temperature is 200*F or less, and b). the maximum operating pressure is 275 psig or less.
3/
Structures, systems, and components fuportant to safety, as specified herein refer to those plant features required to shutdoun the reactor safely and maintain the. plant in the cold shutdoun condition.
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-Break in piping rauans (a) a complete circumferential pipe severance and, (b)'a longitudinal split opening an arca equal to the pipo aren, but without pipe severance, Such breaks are assumed to occur at each specified break location, but not concurrently, i
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Either circumferential or longitudinni stresses derived on an elastically-calculated basis.
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Specified scisciic events are earthquakes that produce at icast' I
50 percent of the vibratory motion of the Safe Shutdoun.tarthquake l
(SSE) 1 2/.
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Operational plant conditions include norraal reactor operation, i
upset conditions, (e.g., anticipated operational occurrences) and j.
testing conditions.
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S is the allovcble strens at manicum temperature and S is the 3
allowable st'rcss range for expansion stresses, for Class 2'and 3 piping as permitted by the rules of ASME Code Scction Ill.
9/
Terrainal ends of pipe runs orig'inate et points of t axinum constraint (e.g., Connections to vessels, pumps, valves, fittingc that are f
rigidly anchored to structures) ter.inal ends of branch ruas originate at pipe intersections and components that act as rigid i
,constra nts.
i 1,0f These criteria are intended for _ ghe purpose of designing piping restraints cad do not preclude consideration of other nr.pects of the AEC General Design Criteria, such as single failura criteria and other adJitional protective ceasures required to provle;c protection against environmental conditions incident to. postulated accidents.
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ATTACl&fENT 020-2 i
DESIGil 0F MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR DIRECT CYCLE BOILING WATER FIACTOR NUCLEAR POWER PLANTS 1.
The main steam isolation valve leakage control system (MSIVLCS) and any l
necessary subsystems should be designed in accordance with seismic I
Category I and Quality Group B requirements, including the source of any I
sealing fluid, if a fluid seal type of system is used except for the following: Any portion of LCS piping that connects to steam system j
piping between inner and outer ccatainment isolation valves of the j
main steam system for either single or dual barrier containment l
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structures up to and including the first isolation valve in the LCS I
piping should be designed in accordance with seismic Category I and t
Quality Group A and the provisions given in* Appendix A of this guide.
I
'2. The MSIVLCS (and any necessary subsystems) should be capable of performing l'
its safety function, when necessary, cons'dering the effects resulting A
'from a"LOCA, including:
(a) missiles that may result from equipment I
failures, (b) dynamic effects associated with pipe whip and jet forces, i
i and (c) normal operating and accident-caused local environmental conditions
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consistent with the design basis event. Further, any portion of the LCS i
which is Quality Group A and is loca'ted outside the primary containment structure should be protected from missiles, pipe whip, and jet force effects originating outside containment such that containment integrity is maintained.
l 3.
The MSIVLCS should be capable of performing its safety function following a LOCA and an assumed single active failure (including failure of any one of the riain steam isolation valves to close).
4.
The MSIVLCS should be designed so that effects resulting from a single active component failure of the Icakage control systen will not affect l
1 the integrity or operability of the main steam. lines or main steam l
isolation valves.
5.
The MSIVLCS should be capable of performing its cafety function following a loss of all offsite power coincident with a postulated design basis LOCA.
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6.
The HSIVLCS should be designed with sufficient, capacity and capability to control leakage from the' main steam lines consistent with the need' for maintaining containment integrity for as long as postulated accident conditions require.
7.
The MSIVLCS may-be manually or automatically actuated and should be
. designed to permit actuation in a time period of about 20 minutes following a postulated design basis LOCA. This time period should be consistent with loading requirements of the emergency electrical buses i
and with reasonable times for operator action.
t' 8.
Instrumentation and circuits necessary for the-functioning of'the i
HSIVLCS should be designed in accordance with standards applicable to an engineered safety feature.
ihe MSIVLCS controls should include interlocks to prevent inadvertent 9.
operation of the MSIVLCS.
In particular, interlocks should be provided:
'to prevent damage to the LCS or possibly to the main steam system due to. inadvertent opening of any LCS isolation valves whenever the pressure in the connecting main steam piping exceeds LCS design pressure.
All such controls and interlocks should be activated from appropriately designed' safety systens or circuits.
10.
The plant should be designed to permit testing of the operability of controls and actuating devices of the MSIVLCS during power operation to theLextent practical, and testing of the complete functioning of the system'during plant shutdowns.
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- 11. The HSIvLSC should be designed so that any effects resulting from use of'a fluid sealing medium, such as thermal' stresses, pressures associated' with flashing, and thermal deformatio.ns. under the.. loading.
- conditions associated with the activated system, will not affect.the structural integrity or. operability of the main steam lines er main steam isolation valves, and that any deformation of isolation valve internals will not induce 11eakage of the main steam line isolation valve beyond the capacity or. capability of the MSIVLCS.
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12.
Means should be provided, as part of the MSIVLCS or otherwise, to prevent or control valve stem packing leakage or other direct leakage 4
from main steam line isolation valves outside containment.
If such inearis afe not part of the MSIVLdS, then they should meet the same designstandardsasthieMSIVLCS.
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APPENDIX A DESIGN OF MAIN STEAM LINE ISOLATION VALVE LEAKAGE CONTROL SYSTEMS FOR DIRECT CYCLE BOILING UATER REACTOR MUCLEAR Pok*ER PLANTS Any portion of piping for a leakage control system that connects to st.2am system piping between inner and outer containment isolation valves of the main steam system for either single or' dual barrier containment structures up to and including the first isolation valve in the LCS piping should be constructed to meet the requirements of
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the ASME Code,Section III, Subarticle NE'-1120 and the following l
additional requirements.
1.
'l;he following design stress and fatigue limits should not be exceeded:
i
- (a) The maxi =um stress range should not exceed 2.4 Sm.
_(b) The maximum stress range between Any two load sets (including the zero load set) should be calculated by Eq. (10) in Par. NB-3653, ASME Code,Section III, for upset plant conditions and an operating basis earthquake (03E) event transient.
If' the calculated naximum stress range of Eq. (10) exceeds 2.4 Sm but, is not greater than 3S,, the cu=ulative usage factor should be 1ess than 0.1.
f i
If.the. calculated maximum stress range of Eq_(10) exceeds 3S, the m
stress ranges calculated by~ both Eq. (12) and Eq. '(13) in par. NS-3653 should not exceed 2.4 S, and the cumulative usage. f actor should oe less than 0.1.
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2.
Welded attachments, for pipe supports crr other purposes, to these portions of piping should be avoided.
3.
The number of piping circumferential and longitudinal' welds should be minimized.
4.
The length of this portion of piping extending to the first. shutoff valves should be reduced to~the minimum length practical.
i 5.
The design of piping restraints shoiild not require welding directly to the outer surface of the piping.
6.
The design of this portion of the 1e.akage control system should permit the conduct of inservice examinations required by the rules of ASME Bo'11er and Pressure Vessel Code - Section X1, and the extent.of examinations during each inspection interval should provide 100 percent volumetric examination of the piping velds I
within this portion of piping.
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April 1973 Hovision 0 BRANCH TlCliM C.\\L hNITION.\\1 CS!i 9-1 OVERHEAD HANDLING SYSTD!S FOR STCLEAR POWER PLANTS A.
BACKCROUND Overhead handling systems are used for hand 11r.g heavy items at nuclear power plants.
The handling of heavy loads such as c spent fuel'eask raises the possibility of damage to the load and to safety-related equipment or structures under and adjacent to the path on which it is transported should the handling system suffer a breakdown or malfunction.
Two methods are used in nuclear power plants to prevent damage to safety features or release of radioactive material due to dropping of heavy loads, such as a spent fuel cask.
One is protection by physical design of the facility to preclude damage to spent fuel and safety-related systers if a heavy load should be dropped. The other is to provide an overhead handling j
system that is designed so that a connected load would not fall in the event of a failure or malfunction.
An everhead, handling system includes all the structural, mechanical, and electrical components that are needed'to lift and. transfer a load from one location to another. Primary load-bearing components, equipment, and subsystems such as the driving e'quipment, drum, rope reeving, control, and braking j
systems require special attention. Proper support of the rope drums ensures that they would*be retained and prevented fron failing or disengaging from
.the, braking and control system in case of a shaft or bearing failure.
If the I
hoisting system (raising and lowering) includes two mechanical holding 1
brakes, each with better than full-load stopp ng capacity, that are automatically activated when electric power is.off or when mechanically tripped by overspeed or overload devices, a critical load will be safely held or controlled in case of failure in the individual lead-bearing parts of the hofsting machinery.
Failure of the bridge or trolley travel to stop when power is shut off or an overspeed or overload condition due to malfunction or failure in the drive
)
system can be prevented and controlled by appropriate safety and limit devices and brake systems, i
l Since the crane industry has not yet developed codes or standards that adequately
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cover the design, operation, and testing for a " single failure-proof" crane, i
the APCSB has developed a branch position to provide a consistent basis for reviewing equipment and conponents for such overhead' handling systems. The
)
position below delineates acceptabic codes and standards and supplements i
them with specific reco=mendations on features that will prevent, control, or stop inadvertent operation or malfunction of the mechanical supporting and j
moving components of the handling system.
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BRM:Cil TECliNICAT. POSITNN Overhead handling systems intended to provide.singic failure-proof handling of loads should be desianed so that no sing.le failure or malfunction will result in dropping or loosing control of the heaviest (critical) loads to be handled.
Such handling systems should be designed, fabricated, installed, inspected. tested, and operated in accordance with the following:
1.
General perfernance S;ecifications l
l Separate performance specifications should be prepared for a pe,rmanent a.
l crane which is to be used for construction' prior to use for plant l
operation. The allowable design stress li=its should be identical for both cases, anc the sun to:al of simultaneously applied loads i
should not result in stress levels causin'g any permanent defor=ation other than that,due to localized stress concentrations.
b.
The operating environment, including taximu= and minimum pressure, temperature, humidity and rates of change of these parameters, should be specified to determine the venting and drainage required for box girder sections. -The specifications should also state the corrosive and hazardous ccnditions that =ay occur during operation.
a Trac'ture toughness for the steel structural mate' rials r,hould be considered. Plate thickness, with a narl n for the lowest operating i
temperatures, should determine the type of steel that can be used with or without toughness tests. The selection of steel materials will be reviewed on a case by case bases.
The crane should be classified as seismic Category I an'd should be c.
capable of retaining the =aximu= design load during 'a safe shutdown earthquake, although the crane may not be operable after the seismic event. The bridge and trolley should be provided with means for preventing them from leaving their runways with or without the de~iign load during operation or under seismic loadings.
The design rate load plus operational and scistically-induced pendulum and swinging load effects on the crane should be censidered in the design of the trolley, and they should be added to the trolley weight for the design of the bridge, d.
All weld joints for load-bearing structures, including those susceptible to lamellar tearing, shocid be inspected by nondestructive examinations
~
for soundness of the base metal and weld =etal.
A fatigue analysis should be considered for critical load-bearing e.
structures and components of the crane handling system. The cumulative fatigue ussgo f actors shoulc reflect effects of cyclic loadings from both the construct on and operating periocs.
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Preheat and posthent treatment temperatures for all weldments should be specified in the weld procedures. For low-alloy steel, the recommendations of Regulatory Guide 1.50 should be followed.
2.
Safety Features The automatic and nanual controls and devices. required for normal a.
i-crane operation should be designed such that a malfunction of these controls and devices,,and possible subsequent effects during load handling, will not prevent the handling system from being maintained at a safe neutral ho1* ding position.'
b.-
Auxiliary systems, dual components', or ancilliary systems should be l
provided such that in case of subsystem or component failure the load will be retained and held in a stable position.
Means should be provided for devices which can be used in repairing, c.
adjusting, replacing failed components or subsystems when failure of an active componene or subsystem has occurred and the load is supported and retained in the safe (temporary) position with the system immobile. As an alternative to repairing the crane in place, d
means may be provided for ccving the handling system with load to a 1aydown area that has been d'esigned for accepting the load and making
- the repairs.
3.
Equipment Selection Dual load attaching point's should be provided on the load block or
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a.
lif ting device designed so that each attaching point will be able to support a static load of 3W (W is weight of the design rated load),
without permanent deformation other than that due to localized stress concentrations in areas for which additional material has been provided for wear.
Lif ting devices such as lif ting beams, yokes, laddle or trunnion' type
- b.. hooks, slings, toggles, or clevises should be of re6;ndant design with dual or auxiliary devices or combinations thereof. Each device should be designed to support a static load of 3W without permanent deformation.
The vertical hoisting *(raising and lowering) mechanism which uses c.
rope and consists of upper sheaves (head block), lower sheaves (load block), and rope reeving system, should be designed with redundant means for hoisting. Maximum hoisting speed should be no greater than 5 fpm.
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The heat and load blocks shou;d be uesigned to maintain a vertical load balance aboet.the contar of ill: from the load block through the head block, and should have a dual reevin's system. 'The load block should maint. tin alignment 'and a position of stability with either system and be able to support 3'i and maintain load stability and vertical alignment from the center of the head block througa all hoisting components'to the center of gravity of the load.
The design of the rope reeving system should be dual, with each system e.
providi'ng separately the load balance on the head and load blocks through the configuration of ropes and rope equalizers.
Selec' tion of the hoisting rope or running rope should consider the size, construction, lay, and means or type of lubrication to maintain efficient working of the ind1Vidual wire strands as the rope passes '
over the sheaves during the hoisting operation. The effects of impact loadings, acceleration and emergency stops.should be included in selection of the rope and reeving system. The wire rope should be 6 x 37 Iron Wire Rope Core (IWRC) or comparable classification.
The stress in the lead line to the drum during hoisting at the' maximum destgn speed with.the design rated load should not exceed 20%
of the manufacturer's rated strength of the rope. The static stress in ' rope (load is stationary) should not exceed 12-1/2 of the manufacturer's rated strength.
Line speed during hoisting (raisihg or lowering should not exceed 50 fp=.
f.. The maximum fleet angle from drum to lead sheave in the load block' should not exceed 3-1/2 degrees at any point during hoisting and there should be only one 180* reverse band for each rope leaving the drum and reversing on the first or lead sheave on the load block, with no other reverse bends other than at the equalizer if a sheave-type equalizer is used.
The fleet angles for rope between individual sheaves should not exceed 1-1/2 degrees. Equalizers may be beam or sheave type.
For the. recommended 6 x 37 IWRC classification wire rope, pitch diameter of the lead she, ave should be 30 times rope diameter for th'e 150' reverse bend, 25 times rope diameter for running sheaves, and 13 times rope diameter for equalizers. The pitch diameter is measured from the center of the rope in the sheave groove through the sheave center.
The dual reeving system may be a single rope from each end of a drum terminat.ing a: a beam-type load and rope stretch equalizer wit each rope designed for total load, or a 2-rope system may be used from each drum or separate drums with a sheave or beam equalizer, or any other combination which provides two separate and complete reeving systems.
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.Thp vertical hoist;nn system componet.t s, which incluJc the head block,
.l rope reeving systen, load block, and dual load. attaching device, should each be designed to sustain.a "oad of 2W (W is the weight of the,das,ign rated load). A 2W static
- oad test should be ' performed j
for each reeving system and load attaching point at the manufacturer's
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- lant..Each reeving system and each one of the. lcad attaching p
l devices should be assecoled.with'ap' proximately a 6 inch clearance betwcon head and load blocks and should support 200% of the design rated. load without degradation.of the components or permanent
' deformation other than that due'to localized stress concentrations.
l Measurements of the. geometric configuration'n of the attaching points should be made.before and after test followed by nondestructive examination, which should consist of combination of magnetic particle, ultrasonic, radiographic, and dye penetrant. examinations to verify.
the soundness of.' fabrication and assure the integrity.of this portion.
of the' hoisting system.- The results of examinations should be documented and recorded for the hoisting system for each overhead crane.
h.
Means should be provided to' sense such items as electric current, temperature, overspeed, overloading, and overtravel.. Controls s
should be provided to-stop the hoisting movement.within_3 inches maximum.of vertical travel through a combination of electrical power controls. and. mechanical braking and torque control systems should one' rope'of the dual reeving system fail.
1.
The control systems may be design'ed as-combination electrical and mechanical systems and may include such items as contractors.. relays, resistors, and thyristors in combination with mechanical devices'and mechanical braking systems. The electric controls should be s. elected to provide a maximum breakdown torque limit of 175% of the required rating for a.c. motors or d.c. motors (series or. shunt wound) used for the hoisting drive motors.
Compound' wound d.c. motors should not l
be used. The control systems provided should consider hoisting l
(raising and lowering) of all loads,. including the design rated load, l
and the effects of inertia of the rotating hoisting machinery such as motor armatures, shaf ts and couplin,e,s, gea'r reducers, and drums.
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-o-I of the,..o's:in; sys:c..
- j. The cachanical a..4 s:rucct.r 1 compoaen: 3 I
should have the rco.irc; strsn un'to.esist failure shculd ":Vo-The cesicner blocking" 1./ or " load hangup". / occur during hoin:ing.
should provide means to abscrh or con:rol the kinetic energy of rotating cachinery in the event of :v.>-b.ockin or loac The loca:.on anc type cf acchat.. cal orakes and con.rc".s hangup.
should provire pos:.:ive and reliable = cans to stop and hold :he hois:ing dr.:8 for : hose occurrences. The hoist:ng system should be able to wi:hstand :he maxi =u
- orque of :he driving cotor, if a be' shut off
- ~ function occur 6 and ;cwer :c :he driving motor canno:
at the tice of load hangup cr :wc-blocking.
k.
The load hoisting drum on tha trcliey shocid be provided with the drum fro = dropping, structural and sechanic safe:y device s :o prevent disengaging from 1:s holding brake syJ:em, or ro: stir.g, should the drum or any portion of its shaft or bearings fail.
To preclude excessive breakdown torque, the horsepower rating (h'?)
1.
of the electri:a1 notor drive for hois:ing should provide no more than 110% of the calculated F.? requiracent to hoist'the design raced
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load at the maxi =us design, hoist speed.
' raking sys:e= should include one power control m.
The tinimum hois: :
braking syste: (no: nachanical or drag 'crake-type) and two rechanical ho'. ding brakes. The hoicing brakas should be activated when power is
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off and should be automatically trippad by sechanical ceans on q
overspeed to the full holding position if a malfunction occurs in the j
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electrical brake controls. Each holding brake should be designed to 125% - 150% of naximum developed torque a: the poin: of application
)
The minicus J
(location of the brake in the techanical drive).
design requirements for br'aking systems that will be operable for e=ergency lowering af:er a single brake failure should be two'
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Provisions j
holding brakes for s:cpping and controlling drum rotation.
should be cade for manual operation of the holding brakes.
Emergency brakes or holding brakes which are to be used for manual lowering should be capable of operation with full load and at full travel and l
Design for manual brake operation provide adequate heat dissipation.
during emergency lowering should include features to limit the lowering
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speed to less than 3.5 fpm.
1,/ "Two-blocking" is an inadvertan ly continued hoist which brings the load and head block assemblies into physical con:act, thereby preventing further movement of the load block and crea:ing shock loads to rope and reeving sys:em.
hoisting by
,2_/ " Load hangup" occurs when the loac block or load is stopped during entanglement with fixed objects, thereby overloading the hoisting system.
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l Thedynamicandststicaligncento[allhoistingnachinerycomponents n.
including gearing, shaf ting, coupling:,, and bearines should be maintained throughout the range of loads be lifted with all l
components positior.cd and anchored on the trolley machinery platform.
Increment drives fcr hoisting may be provided by stepless centrols o.
or inching motor drives.
Plugging 3/ should not be permitted.
Controls te prevent plugging should be included in the electrical circuits and the ccntrol system.
Flocting point 4/ in the electrical power systen, w..en required for bridge or tolley covement, l
should be provided only for the lowest operating speeds.
l To avoid the possibility of overtorque within the control system, p.
the horsepower rating of the driving notor and gear reducer for trolley and bridge notion of an overhead bridge crane should not exceed 110% of the calculated reeuirecent at maximus speed and i
with the design ratec load.
Incremental or fractional inch movenents, when recuired, should be provided by such items as variable I
speed er inching motor drives.
Control and holding. brakes should each,be rated at 100% of miximum drive torque at the point application.
a If two mechanical brakes are provided, one for control and one'for holding, they should be adjustad with one brake in each system for, both the trolley and bridge leading the.orher and should be activated by release or shutoff of power.
The brakes shculd also be technically of a tripped to'the "on" or " holding" positien in the event Provisions malfunction in the power supply or an overspeed condition.
should be made for manual' operation of the brakes. The' holding brake should be designed so that is cannot be used as a foot-operated slowdown brake. Drag brakes should not be used. Opposite wheels on bridges or trolleys which support the bridge or trolley on the runways should be catched and have identical dia eters. Trolley and bridge speed should be linited. A taxicum speed of 30 fpm for the trolley and 40 fpm for the bridge is reco==er.ded.
3/ Plugging is the mo=entary application of full line power to the drive motor for the purpose of promoting a limited movement.
4/ The point in the lowest range of movement control at which power is on, brakes are off, and motors are not energized.
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The, completc. operdt inr. contral ' system and provisions for emer.wney.
3 controls for the overhead crane..andling, system should be located in the. main cab on the bridge.- Additionti cabs located on the trolley or 1.ifting devices should have-complete control systems similst to the bridge cab. Manual controls for.the bridge may be located on the. bridge. Remote cont rols or, pendar.t. controls for'any of these motions should be the same as those provided in the bridge cab control' panel. Previsions should be cade.in the design for devices f or emergency control or operations. Limiting devices, mechanical and electrical, shculd be provided to indicate,. control, and prevent-overtravelling and overspeed or hoist (raising or lowerinC) and for trolley and bridge travel movement. Buffers for bridge and trolley-
. travel should be ir.cluded.
r.
Safety. devices such as limit type switches provided.fo,r malfunction, inadvertent operation, er failure,shculd be in addition to and separate from the control devices providad for operation.
The operating requirements for all travel movements (vertical and s.
horizontal covements, or rctation,' singly or.in combination) for percanent plant ers.nes should be clearly defined in the operating manual for hoistinr. and for trolley and bridge travel'. The designer The.%C. shou' d not should establish the maxima: working load (XWL).
be less than 85% of the design rated load (DRL) capacity for the new crane at time of operation.- The redundancy provided, design factors, selection of components, and balance of auxiliary-ancilliary and duel items in the design and manufacture should be taken into account in setting the maximun working load for the critical load, handling crane system (s). The MWL should not exceed the DEL for overhead cranc "
handling system.
When.the permanent plant crane is to be used for construction and-the t.
' operating requirements for constructica are not identical to those required. for permanent plant service, the construction operating requirements should be defined separa;ely. The crane should be designed structurally and technically for the construction loads, plant service loads, and the functional performance require =ents-for each. At the end of the construction period, the crane handling sys:em should be adjusted for the perfor ance requirements of permanent plant service.
The conversion er adjusteent may include the replacement of such items as motor drives, blocks, and reeving system.
Af t er construction uso, the crane should be thoroughly inspected using nondestructive examinations and should be performance tested.
If the load and performance requirements are different for construction and plant service periods, then the crane should be tested for both The cranc integrity shculd be verified by the designer and phases.
manufacturer and load tes:ing to 1251 of the design rated load j
required for.the operating plant should be done before the crane is j
i used as permanent plant equipment.
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1 Installation instruction should be prcvided by the manufacturer.
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These should include a full explanation of the crane handling system, its controls, and the limitations for the system, and should cover the requirements for installation, testing, and preparation for operation.
4.
Mechanical Checks, Testinc, and preventative Maintenance A complete mechanical check of all crane systems as installed should a.
be made to verify the method of installation and to prepare the crane for testing. During and after installation the proper assembly of electrical and structural components should be verii'ied. The integrity of all control, operating, and safety systems is to be verified as to satisfaction of ' installation and design requirements.
The crane designer and crane manuf aeturer should provide a canual of and crane information and procedures for use in checking, testing, operatien. The manual should also describe a preventive. maintenance program based on the approved test results and information obtained during the testing; it should include such ite=s as servicing, repair, and replacement requirements, visual examinations, inspecti.ons,.
checking, measurements, problem diagnosis, nondestructive e xamination, crane performance testing, and special instructions.
Information concerning proof testing on components and'subsy stems as required..and perfor=ed at the manufacturer's plant to verify for the component or subsystem ability to perform should be available checking and testing perforced at the place of installation e.f the crane system.
The czane system should be prepared for the static test of 125% of the b.
design rated load. The tests should include all positions of h:. isting, and lowering, and trolley and bridge travel with the 125% rated load other positions as reco== ended by the designer and manufacturer..
and adjustments After satisfactory completion of the 125% static test required as a result of the test, the crane handling system shoulc' be given full performance tests with 100% of the design rated' load This i
for all speeds and motions for which the system is designed.-
should include verifying all limiting and safety control devices.
The crane handling system should demonstrate the ability to lower l
l and move the design rate'd load by manual operation and with the use l
of emergency operating controls and devices which have been included 1
in the handling system.
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-10 The conplete i.oi.< ting machinery.60uld be allowed to tvo-block during the hoistina test (lo'as block li=tt and. safety uevices are bypassed). This' test should be conducted without load and at slow speed, to provide assurance of the inter.rity of the design, equipment, controls, and overload protection devices.
The. test
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.should demonstrate'that when the maximum torque that can be j
i developed by the driving system, including the inertia of the rotating parts at the overtorque condition, will be absorbed or contro11'ed prior to two-blocking.
The complete hoisting machinery should be tested for ability to sustain a load hangup concition by a test in which the load block attaching points are secured to a fixed anchor or excessive load. The drum should be capable of one full revolution before starting the hoisting test.
l The preventive maintenance program reco= mended by the designer and c.
manufacturer should also prescribe and establish the MWL for which-the crane will be used. The maxi =um working load should be plainly carked on each side of the crane for each hoisting. unit.
It is-recom= ended that critical' load handling cranes should be continuously maintained at 95% of DRL capacity for the MWL capacity.
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REFERENCES 1.
Regulatory Guide 1.50, " Control of Preheat Temperature for Welding of Low-Alloy Steel."
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" Table of Engineering, Manufacturing, and Operating Standards, Practices.
and References," attached to this position.
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s TABLE OF ENGINEERING, MANUTACIURING, AND 0?EPAION STANDARDS, j
PRACTICES, AND REFERENCES General items for l
Association of Iron and Steel Engineers (Std. No. 6).
overhead cranes and specifically for drems. reeving systems, blocks,
'A$$E controls, and electrical, mechanieni, and structural components.
- Construction, " Manual of Steel Construction."
AISC American Institute of Stee and structural supports.
Runway and bridge design loadings for impact, References for testing, materials, American Society of Mechanical Engineers.
ASME and mechanical components.
American Society for Testing Materials. Testing and selection of materials..
ASTM American National Standards Institute -(410. 33, B6, B15, B29, B30 and N45 ANSI consensus ANSI series N series of ANSI standards for cuality contro.1).
standards for design, manufacturing,'and safety.
Electrical power and
~
Institute of Electrical and Electronics Engineers IEEE -
control systems.
American Welding Society (D1.1.72 - 73/74 revisions). Fabrication AWS requirements and standards for crane structure and weldment.s.
EEI Edison Electrical Institute.
Electrical systems.
' Society of Automotive Engineers, " Standards and Recommended Practice Recommendations and practices for wire rope, shafting, lubrication, SAE fasteners, materials se*_ection,' and load stability.
' Guide for preparing Cranc Manu acturers Association of American (CMAA 70).
functional and performance specifications and component selection.
CMAA Electrical motor, control, National Electrical Manufacturers Association.
NEMA and component selections.
Selection of Wire Rope Technical Board and their manufacturing members; WRIB rope, reeving system, and reeving efficiencies.
Materials'P.andling Institute and their member associations and association members such as American Gear Manufacturing;; Assoc'iation for gears and Fear
}MI reducers, Antifriction Bearing Manufacturers Association for bearings selection, etc.
Welding Research Counci', " Control of Steel Construction to avoid Brittle WRC Fracture," and Bulletin #168, "Lamellar Tearing."
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y 040.0 CONTAINMENT SYSTEMS i
040.1.
From 'our review of Section 1.11 of GESSAR it appears that the following.
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(1.11) items should be included in the specification of interfaces:
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NSSS System 1 (Reactor & Fuel) l (1)- BOP System J (Containment): 'GE should provide blowdown mass and energy release to the drywell head region for a head spray
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line '. ru p ture.
i (2) BOP System L (Hydrogen Mixing System): GE should provide'the total surface area of active fuel cladding in the core and the hydrogen generation due' to pool radiolysis as a function of time.
(3) BOP System M (Hydrogen Recombiner):
Same comments as for the Hydrogen Mixing System. In addition, provide reactor steaming rate (1b/sec versus time) during post-blowdown period.
b.
NSSS System 2 (Nuclear Boiler)
(1) BOP System K (Containment Isolation): The isolation signal should be a drywell pressure of 2 psig not containment pressure.
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4 040-2 (2) BOP System L (Hydrogen Mixing): Specify interfaces, if any.
(3) BOP System J (Containment Design):
Interfaces should be develop-ed to account for loads due to relief valve operation, recognizing that such loads will be a function of the relief valve discharge i
pipe routing and orientation and the suppression pool geometry.
Limits on pool water temperature should also be specified.
c.
NSSS System 9d (RHR - Containment Cooling)
(1) BOP System C '(Water Source): GE should inc1'ude an interface for NPSH.
4 (2) BOP System J (Containment Design):
It appears that this inter-face is one where a BOP supplied system (containment) relies on an NSSS system (RHR) to perform its safety function (Reference Page 1.11-2).
Therefore, the interface should specify that the RHR heat exchangers will provide heat removal capability per BOP designer requirements for containment design.
d.
The containment spray mode of the RHR system should be listed as an NSSS system and interfaces with the BOP identified. Requirements for interlocks between the spray mode and LPCI mode should be l
j considered.
e.
NSSS System 14 (Reactor Water Cleanup System)
)
040-3 l
i (1) BOP System J (Containment Des ~ign): GE should provide blowdown mass and energies for input to subcompartment design.
(2) BOP System K (Containment Isolation): GE should specify closing times for valves as this information is used to determine duration of blowdown and therefore total mass and energy released to con tainmen t.
f.
NSSS System 18 (Hydrocen Recombiner).
Clarify the extent of NSSS responsibility for the recombiner system; 1.e., does it include piping, valves, sizing, environmental qualification?
a 040.2.
Table 1.11.22 identifies information~ in Chapter 6 which is to be supplied (1.11)'
by the NSSS but which does not correspond to its scope of supply; e.g.,
containment analysis, hydrogen mixing.
In addition, some information which should be provided by the NSSS (e.g., blowdown) is not included or listed as an option. Please revise Table 1.11.22 as necessary.
Any information listed as an NSSS option should identify the NSSS system in Table 1.11-0 to which it corresponds.
040.3 Identify the NSSS versus 30P scopes of responsibility with regard to (1.11) defining accident environmental conditions in the drywell and contain-ment and applying these specifications to the design of systems.
040.4. Clarify the NSSS scope of responsibility with regard to containment (1.11) isniation of NSSS supplied systems. Does it include specification of
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040-4 number, type, and arrangement of valves, provisionc for guard pipes, all or some isolation signals, valve closure times, etc.?
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1 130-1 I
i 130.0 STRUCTURAL ENGINEERING l
130.1 Identify all the safety-related components to be supplied by GE as part of the NSSS that need protection from tornado effects.
l 130.2 Identify all the safety-related components to be supplied by GE as part of the NSSS that need protection from flood effects.
1 130.3 Identify all the safety-related components to be supplied by GE as
)
part of the NSSS that need protection from internally and externally i
generated missiles.
130.4 Provide and describe the seismic design requirements at all the appropriate NSSS-B0P interfaces that have to be satisfied by the B0P designer.
Refer to WASH 1341 Amendment 1 for definition of system responsibility.
130.5 Provide and describe the structural design requirements at all the appropriate NSSS-B0P interfaces that have to be satisfied by the BOP designer.
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220-1 220.0 ELECTRICAL, INSTRUMENTATION AND CONTROL SYSTEMS i
220.1 A clear indication of the detailed interfaces between the NSSS and (7.0)
BOP scope of supply in the Instrumentation and Control area is required.
Therefore, (1) list all outputs from NSSS equipment and/or systems which are used to actuate B0P equipment and/or systems.
(2) list all inputs to NSSS equipment and/or systems which are required from 80P equipment and/or systems.
220.2 Specify all power supply requirements of the NSSS scope of supply (8.3.1.1.1) including the neutron monitoring system, the radiation monitoring (7.2) system, and the containment and reactor vessel isolation control system.
220.3 Specify for applicant interface information, the division loads in (8. 3.1.1.1 )
the NSSS scope of supply for the d-c power system.
220.4 The NSSS-BOP interface requirements for Chapter 8 should be clearly defined.
.At a minimum, give the applicable design criteria which i
the B0P supplied power system shall meet in order to satisfy the overall plant design requirements.
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420-1,
420.0 INDUSTRIAL SEC'JRITY AND EMERGENCY PLANNING BRANCH 421.1 In order to evaluate the acceptability of the NSSS design with (13.7) respect to protection against industrial sabotage, addittenal information is required to determine conformance with the design criteria set forth in Regulatory Guide 1.17 and ANSI N18.17-1973.
Identify all vital equipment of the GESSAR-238 NSSS.
Your response shocid be withheld from pijblic disclosure pursuant to 10 CFR 2.790.
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ENCLOSURE 2 i
STANDARD FORMAT FOR LISTS OF REQUESTS FOR ADDITIONAL INFORMATION i
AND REGULATORY STAFF POSITIOIIS
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~;n, 1-BRANCH NUMBERS t"
The following branch numbers shall he used by the contributing branches:
Assistant Branch No.
Directorship Branch Review Aren 000 Reactor Projects Responsible RP Miscellaneous Branch Items 010 Containment' Safety Effluent Treat-All Areas ment Systems 011-Containment Safety Effluent Treat-Applications ment Systems 012 Containment Safety Effluent Treat-
' Systems' Analysis s
ment Systems 020 Containment Safety Aux & Power Con-All Arcas version Systems 040 Containment Safety
. Containment All Areas Systems 041 Containment Safety Containment Section A Systems 042 Containment Safety Containment Section B Systems 110 Engineering Mechanical Engr.
All Areas
'120 Engineering Materials Engr..
All Areas 121 Engineering Materials Engr.
Performance 122 Engineering Materials Engr.
Applications 130 Engineering Structural' Engr.
All Arcas e
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Assistant Branch No.
Directorship _
_ Branch
_ Review Area I
210 Reactor Safety Reactor Systems All Areas j
211 Reactor. Safety Reactor Systems Quali y Group Classifi, cation i
212 Reactor Safety Reactor Systems Section A l
213 Reactor Safety Reactor Systems Section B l
214 Reactor Safety Reactor Systems Section C.,
1 220 Reactor Safety Elect. Inst. &
All Areas t
Control Systems d
221 Reactor Safety Elect. Inst. &
Section A Control Systems 22,2 Reactor Safety Elect. Inst, &
Section B l
Control Systems 223 Reactor Safety Elect. Inst. &
Section C Control Systems 240 Reactor Safety Core Performance All Areas 241 Reactor Safety Core Performance Reactor Fuels 242 Reactor Safety Core. Performance Reactor Physics 243 Reactor Safety Core Performance Thermal and Hydraulics 1
?
310 Site Safety
' Accident Analysis All Areas 320 Site Safety Site Analysis All Areas 321,-
Site Safety Site Analysis Hydrology 322
- Site Safety Site Analysis Meteorology 323 Site Safety Site Analysis Geology / Seismology l
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Assistant Branch No.
Directorship Branch Review Area 330 Site Safety Radiological All Areas Assessment.
331 Site Sa-fety Radiological Radiological Assessment Protection 332 Site Safety Radiological Radiological Assessment Impact 400 Quality Assurance &
Financial 0;perations Matters 410 Quality Assurance &
Quality All Areas Operations Assurance 411 Quality Assurance 6
- Quality QA d
e Operations Assurance 412 Quality Assurance &
Quality Conduct of Operations Assurance Operations 413 Quality Assurance &
Quality Initial Test Operations Assurance and Operations 420
'
- Quality Assurance &
Industrial All Areas Operations Security &
Eme'gency Planning r
421 Quality Assurance &
Industrial Industrial Security Operations Security &
Emergency Planping 422 Quality Assurance &
, Industrial Emergency Planning Operations Security &
Emergency Planning 1
430 Quality Assurance &
Operator All Areas I
Operations Licensing 431 Quality Assurance &
Operator Training Operations Licensing 432 Quality Assurance &
Operator Procedures Operations Licensing
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