Regulatory Guide 1.96
| ML003740263 | |
| Person / Time | |
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| Issue date: | 06/30/1976 |
| From: | Office of Nuclear Regulatory Research |
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| References | |
| RG-1.96, Rev 1 | |
| Download: ML003740263 (5) | |
Revision 1 June 1976 U.S. NUCLEAR REGULATORY COMMISSION
REGULATORY GUIDE
OiFFICE OF STANDARDS DEVELOPMENT
REGULATORY GUIDE 1.96 DESIGN OF MAIN STEAM ISOLATION VALVE LEAKAGE CONTROL SYSTEMS
FOR BOILING WATER REACTOR NUCLEAR POWER PLANTS
A. INTRODUCTION
isolate the reactor system in the event of a break in a steam-line outside the primary containment, a design General. Design Criterion 54, "Piping Systems Pene basis LOCA, or other events requiring containment trating Containment," of Appendix A, "General Design isolation. In the case of a steam line break, the isolation Criteria," to 10 CFR Part 50, "Licensing of Production valves would terminate the blowdown of reactor coolant and Utilization Facilities," requires, in part, that piping in sufficient time to prevent an uncontrolled release of systems penetrating primary containment be provided radioactivity from the reactor vessel to the environment.
with leak detection, isolation, and containment capabili In the case of a LOCA, the valves would isolate the ties having redundancy, reliability, and performance reactor from the environment and prevent the direct capabilities that reflect the impoitance to safety of iso release of fission. products from the containment.
lating these piping systems. This guide describes a basis The valves are part of the reactor coolant pressure acceptable to the NRC staff for implementing General boundary. As such, they are Quality Group A compo Design Criterion 54 with regard to the design of a nents and their integrity must be maintained by strict leakage control system for the main steam isolation inservice inspection and testing requirements. However, valves of boiling water reactor (BWR) nuclear power operating experience has indicated that degradation has plants to ensure that total site radiological effects do not occasionally occurred in the leak-tightness of main steam exceed guidelines of 10 CFR Part 100, "Reactor Site isolation valves, and the specified low leakage require Criteria," in the event of a postulated design-basis ments have not always been maintained.
loss-of-coolant accident (LOCA). If an applicant pro poses to use a method different from that described in The staff has considered the need to provide addi this guide for implementing General Design Criterion 54 tional features to ensure the low-leakage characteristics with regard to the control or limitation of leakage past of the main steam isolation valves in the event of a the main steam isolation valves of a boiling water postulated design-basis loss-of-coolant accident. The
1 reactor, the acceptability of the alternative method will use of a leakage control system would reduce direct
.be determined by the staff on a case-by-case basis. The untreated leakage from the isolation valves when isola Advisory Committee on Reactor Safeguards has been tion of the primary system and the containment is consulted concerning this guide and has concurred in the required.
regulatory position.
The results of staff analyses have indicated that
B. DISCUSSION
calculated doses resulting from the maximum leakage Direct cycle boiling water nuclear power plants supply steam directly from the reactor vessel to the turbine via main steam lines. The main steam lines lln its letters on the construction permit reviews of the Duane Arnold and Shoreham plants (December 18, 1969) and the installed on current BWR plants are provided with dual James A. FitzPatrick plant (January 27, 1970), the Advisory quick-closing isolation valves. These valves function to Committee on Reactor Safeguards noted that additional features to control main steam Isolation valve leakage should be
- Lines indicate substantive changes from previous Issue. considered.
USNRC REGULATORY GUIDES Comments should be sent to the Secretary of the Commission. U.S. Nuclear Regulatory Commission. Washiogton. D.C 20555. Attention: Docketing and available to the public Sergice Section.
Regulatory Guides we issued to describe and make specific parts of the methods acceptable to the NRC staff of implementing Commission's regulations. to delineate techniques used by the staff in evalu- The guides are Issued in the following ten broad divisions:
sting specific problems or postulated accidents, or to provide guidance to appli cants. Regulatory Guides are not substitutes for rgulations. and compliance I Power Reactors 1 Products with them is not required Methods anJ solutions different from those set Out in 2. Research and Test Reactors 7. Transportation the guides will be acceptable if they provide a basis for the findings iequisata to I. Fuels and Materials Facilities . Occupational Health the issuance or continuance of a permit or license by the Commission 4. Environmental and Siting 9. Antitrust Review Comments and suggestions for .mprovements in these guides are encouraged S. Materials and Plant Protection 10. General at all limes, and guides writ be revised. as appropriate. to accommodate com ments and to reflect new information or experience This guide was revised as a Copies of published guides may be obtained by written request indicating the result of substantive comments received from the public and additional stalff divisions desired to the U S Nuclear Regulatory Commission. Washington. 0 C
review 20655. Attention. Director. Office of Standards Development
allowed under the technical specification for the main "realistic" or "best-estimate" dose calculations and steam isolation valves in postulated design-basis LOCA hence in showing margins that might exist above the situations would be a small fraction of the 10 CFR Part limit-type calculations of the staff, a more positive
100 guidelines, 2 provided the main steam system from method of reducing the radiological effects of potential the isolation valves up to and including the turbine leakage of the main steam system isolation valves should condenser remains intact. However, results of staff be provided. The staff also has concluded that some analyses on some typical plants using the standard limited credit for transport delay effects is appropriate conservative assumptions for considering the offsite in determining the design basis for such leakage control consequences of a postulated design-basis LOCA (e.g., systems.
loss of leak-tightness beyond the turbine stop valve, uncontrolled leakage of the main steam isolation valves Staff analyses of the contribution of main steam at or above current typical technical specification limits isolation valve leakage to total calculated offsite doses in of 11.5 standard cubic feet per hour per valve at typical postulated design-basis loss-of-coolant accidents made calculated containment pressures combined. with site with conservative allowances for transport delay effects meteorological data typical of that being presently show that the 2-hour site boundary- dose is not submitted with license applications) have indicated that affected by the subject leakage. The long-term dose in the calculated doses would be in excess of Part 100 the low population zone, however, is affected for guidelines. uncontrolled Isolation valve leakage rates typical of current technical specification values. Thus the staff has The position of the staff with respect to the seismic concluded that a fully automatic quick-acting leakage design classification of the steam system does not control system Is not required to meet the system require Seismic Category I design requirements for the objectives. A manually initiated leakage control system turbine stop and control valves, steam line piping capable of being actuated within about 20 minutes of an beyond the stop valve, the turbine, the turbine con accident requiring .use of the system would be accept denser, or connecting piping of less than 2% inches in able.
diameter. However, there is a need for design improve ments to provide appropriate safety margins for the large It should be noted that any leakage from the stem numbers of plants now planned. The staff believes that, packing of the outboard isolation valve would contribute unless systems can be relied on to remain intact and to the 2-hour dose, since in most designs such leakage capable of providing significant dose reduction factors in would escape to the turbine building and the environ postulated accident conditions, a leakage control system ment via the steam tunnel. Reduction and control of steam packing leakage or other direct leakage to the for main steam isolation valves should be provided for new boiling water reactor plants3 to supplement the steam tunnel from the outboard isolation valve should be a design objective of the leakage control system or of isolation function of the main steam isolation valves and other systems provided for this purpose.
reduce uncontrolled or untreated leakage from the steam line valves.
C. REGULATORY POSITION
It has been proposed that dose reduction factors due The isolation function of the main steam isolation to the transport delay time of the containment atmos valves in boiling water reactor plants should be supple phere in passing through the main steam lines within mented by a leakage control system (LCS). An accept containment or through the main steam lines from the able approach for such a leakage control system is isolation valves to the turbine stop valves should be provided by the following design basis:
included in staff calculations of postulated accident effects. Analyses by some applicants, based on assump 1. The leakage control system and any necessary tions different from those used by the staff, have subsystems, including the source of any sealing fluid if a indicated that long transport delays might occur. On fluid seal type of system is used, should be designed in that basis, it has been argued that a leakage control accordance with Seismic Category I and Quality Group system is not necessary to reduce potential leakage from B requirements, with the exception of any portion of the steam systems of boiling water reactor plants. The LCS piping that connects to main steam system piping staff has considered these analyses and has concluded between inner and outer containment isolation valves of that, although they are useful in making so-called the main steam system for either single- or dual-barrier containment structures. Such piping, up to and including
2 the first isolation valve in the LCS piping, should be Part 100 guidelines, as used in this guide, refer to the radiation designed in accordance with Seismic Category I and dose limits used in determining the boundaries of the exclusion Quality Group A requirements supplemented by Appen area and the low population zone pursuant to 10 CFR Part
100. dix A of this guide.
3 The staff defimes "new" boiling water plants to be those plants utilizing the General Electric Company's BWR 6/Mark IlI 2. The LCS (and any necessary subsystems) should design or subsequent BWR designs. be capable of performing its safety function, when
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necessary, considering effects resulting from a LOC*k, 10.'The plant should be designed to permit testing of including (a) missiles that may result from equipme nt the operability of controls and actuating devices of the failures, (b) dynamic effects associated with pipe wh ip LCS during power operation to the extent practical and and jet forces, and (c) normal operating and accider it- testing of the complete functioning of the system during caused local environmental conditions consistent with plant shutdowns.
Is the design-basis event. Further, any portion of the L(
that is Quality Group A and is located outside he tC 11. The LCS should be designed so that (a) any primary containment structure should be protected fro,m effects resulting from use of a fluid sealing medium, e.g.,
missiles, pipe whip, and jet force effects originati ng thermal stresses, pressures associated with flashing, and outside containment so that containment integrity is thermal deformations under the loading conditions maintained. associated with the activated system, will not affect the structural integrity or operability of the main steam lines
3. The LCS should be capable of performing iIts or main steam isolation valves and (b) any deformation safety function following a LOCA and assumed single of Isolation valve internals will not induce leakage of the active failure (including failure of any one of the ma in main steam line isolation valve beyond the capacity or stearn isolation valves to close). capability of the LCS.
4. The LCS should be designed so that effec:ts 12. Equipment should be provided, as part of the resulting from failure of a single active component of tiie LCS or other systems, to prevent or control valve stem leakage control system will not affect the integrity or packing leakage or other direct leakage from main steam operability of the main steam lines or main stea m line isolation valves outside containment. If such equip isolation valves. ment is not part of the LCS, it should meet the same design standards as the LCS.
5. The LCS should be capable of performing iits safety function following a loss of all offsite pow er
D. IMPLEMENTATION
coincident with a postulated design-basis LOCA.
The purpose of this section is to provide information
6. The LCS should be designed with sufflcie:nt to applicants and licensees regarding the NRC staff's capacity and capability to control leakage from the ma in plans for using this regulatory guide.
steam lines for as long as postulated accident conditic.as require containment integrity to be maintained. This guide reflects current regulatory practice. There fore, except in those cases in which the applicant
7. The LCS may be manually or automatical ly proposes an acceptable alternative method for comply actuated and should be designed to permit actuati(on ing with specified portions of the Commission's regula within about 20 minutes after a postulated design-ba! sis tions, the method described herein is being and will LOCA. This time period is considered to be consiste.nt continue to be used in evaluating submittals for con with loading requirements of the emergency electric:al struction permit and operating license applications.
Although this guide may recommend backfitting in buses and with reasonable times for operator action.
certain cases that have already been docketed, as
8. Instrumentation and circuits necessary for tihe described below, it does not require it. Such require functioning of the LCS should be designed in accordan ce ments will be formulated on an individual basis pursuant with standards applicable to an engineered safety to 10CFR §50.109.
feature.
1. In the case of boiling water reactor plants for
9. The LCS controls should include interlocks to which construction permits were issued prior to March prevent inadvertent operation of the LCS. In particul,1r, 1, 1970 (see Table 1), applicants and licensees should interlocks should be provided to prevent damage to tihe continue the established inservice inspection programs to LCS or possibly to the main steam system due to ensure that isolation valves are maintained in such a inadvertent opening of any LCS isolation valves whe n- manner that leakage is within Technical Specification ever the pressure in the connecting main steam pipi rig limits. If the valve inspections show recurring problems exceeds LCS design pressure and to prevent significant with excessive leakage, the staff recommends that release of radioactivity to the environment. Whe!re consideration be given to installation of a supplementary appropriate to the LCS design, interlocks should Ibe leakage control system.
provided to preclude direct communication with t he post-LOCA containment atmosphere in the event th.at 2. In the case of boiling water reactor plants of the inboard main steam isolation valve does not fully designs preceding the BWR 6/Mark III design and for close. AU such controls and interlocks should be 1 which construction permits have been issued after March activated from appropriately designed safety systems or 1, 1970 (see Table 1), the staff recommends that circuits. applicants and licensees install a supplemental leakage
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control system. Leakage control systems for these plants including and subsequent to the first BWR 6/Mark III
should be designed in accordance with the recommend (i.e., the Grand Gulf project), the staff recommends that ations in the regulatory position of this guide to the applicants and licensees install a supplemental leakage extent practical considering the stage of plant design and control system to ensure the isolation function of the construction. The system provided for each plant and main steam isolation valves. Leakage control systems for the schedule for installation will be reviewed on a the plants should be designed in accordance with the case-by-case basis. recommendations in the regulatory position of this guide.
3. In the case of boiling water reactor plants of the BWR 6/Mark III design (or subsequent BWR design),
TABLE 1 LIST OF BWR PLANTS
SECTION D.A PLANTS CP REPORT ISSUED
Dresden I 7/55 5/56 Big Rock Point 3/60 5/60
Humboldt Bay 6/60 11/60
Lacrosse 12/62 3/63 Oyster Creek 8/64 12/64 Nine Mile I 10/64 4/65 Dresden 2 11/65 1/66 Millstone 1 3/66 5/66 Dresden 3 8/66 10/66 Quad Cities 1,2 12/66 2/67 Browns Ferry 1,2 3/67 5/67 Monticello 4/67 6/67 Vermont Yankee 6/67 12/67 Peach Bottom 2,3 10/67 1/68 Cooper 3/68 6/68 Browns Ferry 3 5/68 7/68 Pilgrim 1 4/68 8/68 Hatch 1 5/69 9/69 Brunswick 1,2 10/69 2/70
SECTION D.2 PLANTS
FitzPatrick 1/70 5/70
Duane Arnold 12/69 & 2/11/70 6/70
Fermi 2 2/71 9/72 Zimmer 9/71 10/72 Hatch 2 11/71 12/72 Hanford 2 10/72 3/73 Shoreham 12/69 & 2/70 4/73 LaSalle 1,2 12/71 9/73 Susquehanna 1,2 4/72 11/73 Hope Creek 1,2 8/71 & 2/74 Pending (ex. Newbold Island)
Limerick 1,2 8/71 6/74 Bailly 10/71 5/74 Nine Mile 2 7/73 6/74
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APPENDIX A
SUPPLEMENTAL DESIGN FEATURES FOR QUALITY GROUP A PORTION OF
LEAKAGE CONTROL SYSTEM PIPING
This appendix provides supplemental design features If the calculated maximum stress range of Equation for any portion of piping for a leakage control system (10) exceeds 3 Sm, the stress ranges calculated by both (LCS) that connects to steam system piping between Equation (12) and Equation (13) in Paragraph NB-3653 inner and outer containment isolation valves of the main should not exceed 2 A4 Sm and the cumulative usage steam system for either single- or dual-barrier contain factor should be less than 0.1.
ment structures. Such piping, up to and including the first isolation valve in the LCS piping, should be 2. Welded attachments to this portion of piping for constructed to meet the requirements of the ASME pipe supports or other purposes should be avoided.
Code in Subarticle NE-1 120 of Section III, supple mented by the following: 3. The number of circumferential and longitudinal welds in the piping should be minimized.
1. The following design stress and fatigue limits should not be exceeded:
4. The portion of piping extending to the first shutoff valve should be as short as practical.
a. The maximum stress range should not exceed
2.Sm.I
5. The design of piping restraints should not require b. The maximum stress range between any two welding directly to the outer surface of the piping.
load sets (including the zero load set) should be calculated by Equation (10) in Paragraph NB-3653, 6. The design of this portion of the leakage control ASME Code,Section II, for upset plant conditions and system should permit the conduct of inservice examina an operating basis earthquake transient. tions required by the rules of Section XI of the ASME
Boiler and Pressure Vessel Code, and the extent of If the calculated maximum stress range of Equation examinations during each inspection interval should
(10) exceeds 2 . 4Sm but is not greater than 3 Sm, the provide 100 percent volumetric examination of the cumulative usage factor should be less than 0.1. piping welds within this portion of piping.
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