Regulatory Guide 1.7: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
Line 1: Line 1:
{{Adams
{{Adams
| number = ML13350A213
| number = ML070290080
| issue date = 09/30/1976
| issue date = 03/23/2007
| title = Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident
| title = Control of Combustible Gas Concentrations in Containment
| author name =  
| author name = Pulsipher J
| author affiliation = NRC/OSD
| author affiliation = NRC/NRO/DSRA
| addressee name =  
| addressee name =  
| addressee affiliation =  
| addressee affiliation =  
| docket =  
| docket =  
| license number =  
| license number =  
| contact person =  
| contact person = Pulsipher J C, NRO/DSRA, 415-2811
| document report number = RG-1.007, Rev. 1
| case reference number = DG-1117
| document report number = RG-1.007, Rev. 3
| package number = ML070290076
| document type = Regulatory Guide
| document type = Regulatory Guide
| page count = 10
| page count = 13
}}
}}
{{#Wiki_filter:7 U.S. NUCLEAR. REGULATORY  
{{#Wiki_filter:The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staffconsiders acceptable for use in implementing specific parts of the agency's regulations, techniques that the staff uses in evaluating specific problemsor postulated accidents, and data that the staff need in reviewing applications for permits and licenses.  Regulatory guides are not substitutesfor regulations, and compliance with them is not required.  Methods and solutions that differ from those set forth in regulatory guides will be deemedacceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Comm ission.This guide was issued after consideration of comments received from the public.  The NRC staff encourages and welcomes comments and suggestionsin connection with improvements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed. The NRC staff will revise existing guides, as appropriate, to accommodate comments and to reflect new information or experience.  Written commentsmay be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.Regulatory guides are issued in 10 broad divisions:  1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities;4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review;and 10, General.Requests for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission,Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov. Electronic copies of this guide and other recently issued guides are available thr ough the NRC's public Web site under the Regulatory Guides documentcollection of the NRC's El ectronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/
COMMISSION
and through the NRC's Agencywide DocumentsAccess and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070290080.U.S. NUCLEAR REGULATORY COMMISSIONMarch 2007Revision 3 REGULATORY GUIDEOFFICE OF NUCLEAR REGULATORY RESEARCHREGULATORY GUIDE 1.7(Draft was issued as DG-1117, dated August 2002)CONTROL OF COMBUSTIBLE GAS CONCENTRATIONSIN CONTAINMENT
Revision 1 REGULATORY  
GUIDE September
1976 OFFICE OF STANDARDS
DEVELOPMENT
REGULATORY  
GUIDE 1.7 CONTROL or COMIBUSTIBLE
GAS CONCENTRATIO6NS
IN CONTAMN14ENT
FOLLOWING
A LOSS-OF-COOLAINT
ACCIDENT*


==A. INTRODUCTION==
==A. INTRODUCTION==
Criterion
In September 2003, the U.S. Nuclear Regulatory Commission (NRC) issued a revision ofSection 50.44, "Combustible Gas Control for Nuclear Power Reactors" (Ref. 1), which amended Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), "Domestic Licensing of Production and Utilization Facilities" (Ref. 2).  This regulation is applicable to all reactor construction permits or operating licenses under 10 CFR Part 50, except for those facilities for which the certifications requiredunder Section 50.82(a)(1) have been submitted, and to all reactor design approvals, design certifications,combined licenses or manufacturing licenses under 10 CFR Part 52, "Early Site Permits; Standard DesignCertifications; and Combined Licenses for Nuclear Power Plants" (Ref. 3).  This regulatory guide describes methods that are acceptable to the NRC staff for implementing the revised Section 50.44 for reactors, subject to the provisions of Sections 50.44(b) or 50.44(c).
35, "Emergency Code Cooling," of Appendix A, "General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50, "Licensing of Pro-duction and Utilization Facilities," presently requires that a system be provided to provide abundant emergency core coolin
This regulatory guide relates to information collections that are covered by the requirements of 10 CFR Parts 50 and 52, which the Office of Management and Budget (OMB) approved under OMB


====g. Criterion ====
control numbers 3150-0011 and 3150-0151, respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement
50, "Contain-ment Design Basis," presently requires that the reactor containment structure be designed to accommodate, without exceeding the design leakage rate, condi-tions that may result from degraded emergency core cooling functionlng...
Criterion
41, "Containment Atmospl-2re Cleanup," presently.requires that sys-tems to control hydrogen, oxygen, and other substances that may be released into the reactor containment be provided as necessary control the concen-trations of such substances following postulated accidents and ensure that containment intuegrity is maintained.


In addition, the Commission has published progposed amendments to Part 50 containing standards for combustible gas control systems. This guide describes methods that would be acceptable to th e.NRC staff for implementing t.he pro-posed regulation, assuming it is promul gated.'as an effective rule by the Commission after consideration of.'ipublic comments, for l..ight-water reactor plants with cylindrical, zircaloy-clad', uxide fuel. Light-water reactor plants with stainless steel cladding:and those with noncylindrical cladding will continue to be consideied on an individual basi
unless the requesting document displays a currently valid OMB control number.


====s. Bi DISCUSSION====
Rev. 3 of RG 1.7, Page 2
Following'U
loss-of-coolant accident (LOCA), hydrogen gas may accumulate within the containment as a result of 1. Metal-water reaction involving the zirconium fuel cladding and the-.reactr r.*..coolant, This guide replaces Safety Guide 7, dated 3/10/71, and Supplement to Safety Guide 7, dated 10/27/7


===1. USNRC REGULATORY ===
==B. DISCUSSION==
GUIDES Co.... is Mot be sent to 'he SetiJavy 0 ite Conission.
Section 50.44 provides requirements for the mitigation of combustible gas generated bya beyond-design-basis accident. In existing light-water reactors, the principal combustible gas is hydrogen.In an accident more severe than the design-basis loss-of-coolant accident (LOCA), combustible gas is predominately generated within the containment as a result of the following factors:(1)fuel clad-coolant reaction between the fuel cladding and the reactor coolant
(2)molten core-concrete interaction in a severe core melt sequence with a failed reactor vessel If a sufficient amount of combustible gas is generated, it may react with oxygen present in the containment at a rate rapid enough to lead to a containment breach or a leakage rate in excess of


U.S. Nuclei.Nlguvistory Gulde ate istst d to describe eind makhe availabte to the Public Rlegulaloiro Wai0"ingcn.
technical specification limits. Additionally, damage to systems and components essential to continued


D C Attention Oc.clitiong end met hodaI accopti ble to the NFIC stalf of Implementing ssociftic part of the Seritice Secti,
control of the post-accident conditions could occur.
*igulttlois.


to delineite tethniQuoi used by the itaff in eovie ihe guides ase issus d In thefollowwg tinbtow d divisions iting specific problems at postulated accidents.
In SECY-00-0198, "Status Report on Study of Risk-Informed Changes to the Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to
10 CFR 50.44 (Combustible Gas Control)" (Ref. 4), the NRC staff recommended changes to 10 CFR 50.44 that reflect the position that only combustible gas generated by a beyond-design-basis accident is a risk-significant threat to containment integrity.  Based on those recommendations, the September 2003 revision of 10 CFR 50.44 eliminates requirements that pertain only to design-basis LOCAs.


or providi guidanc, to ippli cents. Regulatory Guides em. not substitiute&
Attachment 2 to SECY-00-0198 (Ref. 4) used the framework described in Attachment 1 to the paper with risk insights from NUREG-1150 (Ref. 5) and the integrated plant evaluation programs
lo, regulations.


and compliance
to evaluate the requirements in 10 CFR 50.44. In so doing, Attachment 2 noted that containment types
$. Power Reactors 4 P-odccts with ihlem isi olt quttied. MerhOds ind solutions different from those set out in 2. Reiteech end teal Rtactort, 7 the g uIdes wi-l be ifecepteble sIt atd bi o the indings etquils to 3 fuel, and Materitsals, acdtiev Occupational  thitissualceao l c ofmlicont:
bythie Commision
4A f riOn ninentotendSiling
9 Antitrust Cohmirncnis and fo improvements in these guides ite encowteged S Materatls and Pleant Protection tO Genetal*I alt times. end guides will he revised. mi appropriati.


to ¢ccommodalt r.om mints and to rltIct new inlotmnetin or e.ptorince.
that rely on pressure suppression concepts (i.e., ice baskets or water pools) to condense the steam from a design-basis LOCA have smaller containment volumes, and in some cases lower design pressures, than pressurized-water reactor (PWR) large-volume or subatmospheric containments. Consequently, the smaller volumes and lower design pressures associ ated with pressure suppression containment designs make them more vulnerable to combustible gas deflagrations during degraded core accidents because


However. ccrnmniett on Cope* of pubtished guides mey be obtained bn -.Iltlen idcatimg the tht tuide. t receivied within about two momthsl ofter its 4suance witt h p. divisions to the U S Nuclear tegualtoty Comn -sson Washington
the pressure loads could cause structural failure of the containment. Also, because of the smaller volume of these containments, detonable mixtures could be formed. A detonation would impose a dynamic pressure load on the containment structure that could be more severe than the static load from an equivalent deflagration.  However, the staff noted in SECY-00-0198 that the risk of early containment failure
0 C.liculaill useful'tn evaluating it o I.tt. eIy--on 2MM. Attention Ottecto'.
Otlice ot Standards Oeiaoopmnit
'i"A , ,.,,.,. 4,.
..............
2. Radiolytic decomposition of the postaccident emergency cooling solutions (oxygen will also evolve in this process), 3. Corrosion of metals by solutions used for emergency cooling or con-taenment spray.If a sufficient amount of hydrogen is generated, it may react with the oxygen present in the containment atmosphere or, in the case of inerted containments, with the oxygen generated following the accident.


The reac-tion would take place at rates rapid enough to lead to high temperatures and significant overpressurization of the containment, which could result in a leakage rate above that specified as a limiting-condition for operation in the Technical Specifications of the license. Damage to systems and components essential to the continued control of the post-LOCA
from combustible gas combustion in these types of containments can be limited by the use of mitigative
conditions could also occur.The extent of metal-water reaction and associated hydrogen production depends strongly on the course of events assumed for the accident and on the effectiveness of emergency cooling systems. Evaluations of the per-formance of emergency core cooling systems (ECCS) included as engineered safety features on current light-water-cooled reactor plants have been made by reactor designers using analytical models described in the "Interim Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Power Reactors" published in the Federal Register on June 29, 1971, and as amended on December 18, 1971.* These calculations are further discussed in the staff's concluding statement in the rule making hearing on the Acceptance Criteria, Docket RM-50-1.**
The result of such evaluations is that, for plants of current design operated in confdrmance with the Interim Acceptance Criteria, the calculated metal-water reaction amounts to only a fraction of one percent of the fuel cladding mass. As a result of the rule making hearing (Docket RM-50-1), the Commission adopted regulations dealing with the effectiveness of ECCS (10 CFR Part 50, § 50.46).The staff believes it is appropriate to consider the experience obtained from the various ECCS--related analytical studies and test programs, such as code developmental efforts, fuel densification, blowdown and core heatup studies, and the PWR and BWR FLECHT tests, and to take account of the increased conservatism for plants with ECCS evaluated under § 50.46 in setting the amount of initial metal-water reaction to be assumed for the purpose of establishing design requirements for combustible gas control systems. The staff has always separated the design bases for ECCS and for containment systems and has required such containment systems as the combustible gas*4 36 FR 12248 and 36 FR 24082.A copy of the docket file may be examined in the NRC public dociment room.1.7-2 4
4 control system to be designed to withstand a more degraded condition of the reactor than the ECCS design basis permits. The approach is consistent with the provisions of General Design Criterion
50 in which the need :o provide safety margins to account for the effects of degraded ECCS ['fiction is noted. Althoutgh the level of degradation considered might lead to an assumed extent of metal-water reaction in excess of that calculated for acceptable ECCS performance, it does not lead to a situation involving a total failure of the ECCS.The staff fuels that this "overlap" In protection requirements provides ain appropriate ind prudent safety margin against unpredicted events during the course of accidents.


Accordingly, the amount of hydrogen assumed to be generated by metal-water reaction in establishing combustible gas control system performunce requirements should be based on the amount calculated in demonstrating compliance with 5 50.46, but the amount of hydrogen required to be assumed should include a margin above that calculated.
features:  (1) inerting in Mark I and II containments and (2) using igniter systems in Mark III


To obtain this margin, the assumed amount of hydrogen should be no less than five times that calculated in accordance with 5 50.46.Since the amounts of hydrogen thus determined may be quite small for many plants (as a result of the other more stringent .equirements for ECCS performance in the criteria of 9 50.46), it is consistent with tihe considera- tion of the potential for degraded ECCS performance discussed above to establish also a lower limit on the assumed ameunt of hydrogen generated by metal-twater reactions in establishing combustible gas control system require-menrts. In establishing this lower limit, the staff has considered the fact that tile maximum metal-water reaction permitted by the ECCS performance criteria is one percent of the cladding mass. Use of this "one percent of tile mass" value as a lower limit for assumed hydrogen production, however, would unnecessarily penalize reactors with thicker cladding, since for the same thermal conditions in the core in a postulated LOCA, the thicker cladding would not, in fact, lead to increased hydrogen generation.
and ice condenser containments.  As a result, the revised Section 50.44 has the following requirements:(1)All boiling-water reactor (BWR) Mark I and II type containments must be inerted.  By maintaining an oxygen-deficient atmosphere, combustible gas combustion that could threaten containment


This is because the hydrogen generation from metal-water reaction is a surface phenomenon.
integrity is prevented.(2)All BWRs with Mark III type containments and all PWRs with ice condenser type containments must have the capability to control combustible gas generated from a metal-water reaction


A more appropriate basis for setting the lower limit would be an amount of hydrogen assumed to be generated per unit cladding area. It is convenient to specify for this purpose a hypothetical uniform depth of cladding surfacce reaction.
involving 75% of the fuel cladding surrounding the active fuel region (excluding the cladding


The lower limit of metal-water reaction hydrogen to be assumed is then the hypothetical amount that would be generated if all the metal to a specified depth in the outside surfaces of the cladding cylinders surrounding the fuel (excluding the cladding surrounding the plenum volume) were to react.In selecting a specified depth to be assumed as a lower limit for all reactor designs, the staff has calculated the depth that could correspond to the "one percent of the mass" value for the current core design with the 1.7-3... ....:...
surrounding the plenum volume) so that there is no loss of containment structural integrity.
..... .... ... ... .thinnest cladding.


This depth (0.01 times; the thickness of the thinnest fuel cladding is used) is 0.00023 inch.In summary, the amount of hydrogen to be generated by metal-water reaction in determining the performance requirements for combustible gas control systems should be five time:; the maximum amount calculated in accordance with 1 50.46, but no less than the amount that would result from reaction of all the metal in the outside surfaces of the cladding cylinders surrounding the fuel (excluding the cladding surrounding tie plenum volume)to a depth of 0.00023 inch.It should be noted that the extent of initi.al metal-water reaction calculated for the first core of a plant and used as a design basis for the hydrogen control system becomes a limiting condition for all reload cores in that plant unless the hydrogen control system is subsequently modified and reevaluated.
The deliberate ignition systems provided to meet this existing combustible gas source term


The staff believes that hydrogen control systems in plants receiving operating licenses on the basis of ECCS evaluationn under the "Interim Acceptance Criteria" should continue to be designed for the 5 percent initial metal-water reaction specified In the original issuance of this guide (Safety Guide 7). As operating plants are reevaluated as to ECCS performance under S 50.46, a change to the new hydrogen control basis enumerated in Table 1 may be made by appropriate amendments to the Technical Specifications of the license. For plants receiving construction permits on the basis of ECCS evaluations under tile Interim Acceptance Criteria, the applicant would have the option of using either a 5 percent initial metal-water reaction or five times the maximum amount calculated in accordance with S 50.46, but no less than the amount that could result from reaction of all the metal in the outside surfaces of cite cladding cylinders surround-ing the fuel (excluding the cladding surrounding tite plenum volume) to a depth of 0.00023 inch.No assumption as to rate of evolution was as:;ociated with the magnitude of the assumed metal-water reaction originally given in Safety Cuide 7.The metal-water reaction is of significance when establishing system per-formance requirements for containment designs that employ time-dependent hydrogen control features.
are capable of safely accommodating even greater amounts of combustible gas associated with


The staff recognizes that it would be unrealistic to assume a' instantaneous release of hydrogen from an assumed metal-water reaction.
even more severe core melt sequences that fail the reactor vessel and involve molten


For the design of a hydrogen control system, therefore, it should be assumed that the initial metal-water reaction would occur over a short period of time early in the LOCA transient, i.e., near the end of the blowdown and core refill phases of the LOCA transient.
core-concrete interaction.  Deliberate ignition systems, if available, generally consume the combustible gas before it reaches concentrations that can be detrimental to containment integrity.


Any hydrogen thus evolved would mix with steanu and would be rapidly distributed throughout the containmoent compartments enclosing the reactor primary coolant system by steam flowing from the postulated pipe break. These compartments include the "drywell" in typical boiling water reactor containments, the 1. 7-4  
Rev. 3 of RG 1.7, Page 3(3)For all applicants for and holders of a water-cooled reactor construction permit or operating license under 10 CFR Part 50, and all applicants for a light-water reactor design approval, or design certification, or combined license under 10 CFR Part 52 that are docketed after October 16, 2003, the effective date of the rule, the following requirements apply.  All containments must have an inerted atmosphere or limit combustible gas concentrations in containment during and following an accident that releases an equivalent amount of combustible gas as would be generated from a 100% fuel-clad coolant reaction, uniformly distributed, to less than 10%
7. ."lower volume" of ice condenser containments, and the full volume of "dry" containments.
(by volume) and must maintain containment structural integrity.  The requirements of this paragraphapply only to water-cooled reactor designs with characteristics (e.g., type and quantity of claddingmaterials) such that the potential for production of combustible gases is comparable to light-water reactor designs licensed as of October 16, 2003.(4)For all construction permits and operating licenses under 10 CFR Part 50, and all design approvals,design certifications, combined licenses, or manufacturing licenses under Part 52, for non-water-cooled reactors and water-cooled reactors that do not fall within the description in paragraph 3 (above), any of which are issued after October 16, 2003, applications subject to this paragraph


The duration of the blowdown and refill phase is generally several minutes, and the assumptions oi a two-minute evolution time, which represents the period of time during which the maximum full heatup occurs, with a constant reaction rate and with the resulting hydrogen uniformly distributed in the containment compartments enclosing the primary coolant*ystem, are appropriately conservative for the design of hydrogen control systems. The effects of steam within the subcompartments and containment should be considered in the evaluation of the mixture c.omposition.
must include the following:(a)information addressing whether accidents involving combustible gases are technically relevant for their design(b)if accidents involving combustible gases are found to be technically relevant, information demonstrating that the safety impacts of combustible gases during


The rate of production of gasrs from radiolysis of coolant solutions depends on (1) the amount and quality of radiation energy absorbed in the specific coolant solutions used and (2) the net yield of gases generated from the solutions due to the absorbed radiation energy. Factors such as coolant flow rate.. and turbulence, chemical additives in the coolant, impurities, and coolant temperature can all exert an influence on the gas yields from radiolysis.
design-basis and significant beyond-design-basis accidents have been addressed to ensure adequate protection of public health and safety and common defense and security.


The hydrogen production rate from corrosion of materials within the containment, such as aluminum, depends on the corro-sioii rate, which in turn depends on such factors as the coolant chemistry, thie coolant. pul, the metal and coolant temperatures, and the surfbce area exposed to attack by the coolant. Accurate values of these parameters are difficult to establish with certainty for the conditions expected to pre-vail following a LOCA.Table I defines conser, tive values and assumptions that may be used to Cv.Iluate the production of cu:.-,)ustible gases following a LOCA.If thes;e ansumptions are used to calculate the concentration of hydrn-g-cn (and oxygen) within the containment structures of reactor plants following a LOCA, the hydrogen concentration is calculated to reach the limit within periods of less than a day after the accident for the smalle.st containments and up to more than a month for the largest ones.The hydrogen concentration could be maintained below its lower flammable limit by purging the containment atmosphere to the environs at a controlled rate after the I.OCA, however, radioactive materials in the containment would also be released.
The combustible gas control systems, the atmosphere mixing systems, and the provisionsfor measuring and sampling that are required by Section 50.44 are risk-significant, as they have the abilityto mitigate the risk associated with combustible gas generation caused by significant beyond-design-basis accidents. The recommended treatments for those systems are delineated in the regulatory position


Therefore, purging should not be the primary means for controlling combustible gases following a LOCA. It Is advisable, however, that the capability for controlled purging be provided to aid in containment atmosphere cleanup.The Bureau of Mines has conducted experiments at its facilities with initial hydrogen volume concentrations on the order of 4 to 12 volume percent.On the basis of these experiments and a review of other reports, the NRC staff concludes that a lower flammability limit of 4 volume percent hydrogen in air or steam-air atmospheres is well established and is adequately conservative.
in Section C of this regulatory guide.The hydrogen monitors should be able to assess the degree of core damage during a beyond-design-basis accident and confirm that random or deliberate ignition has taken place. Hydrogen monitors, in conjunction with oxygen monitors, are further relied on to implement severe accident management


For initial concentrations of hydrogen greater than about 6 volume percent, it is possible in the presence of sufficient ignition 1.7-5
strategies to address a potential breach of containment integrity or to consider containment purging
* -.. .*~..i'*sources that the total accumulated hydrogen could burn in the containment.


For hydrogen concentrations in the range of 4 to 6 volume percent, partial burning of the excess hydrogen above 4 volume percent may occur. The staff believes that a limit of 6 volume percent would not result in effects that would be adverse to containment systems. Applicants or licensees proposing a design limit in the range of 4 to 6 volume percent hydrogen should demonstrate through supporting analyses and experimental data that contain-ment features and safety equipment required to operate after a LOCA would not be made inoperative by the burning of the excess hydrogen.In small containments, the rduount of metal-water reaction postulated in Table 1 may result in hydrogen concentrations above acceptable limits.The evolution rate of hydrogen from the mrtal-water reaction would be greater than that from either radiolysis or corrosion, and since it is difficult for a hydrogen control system to process large volumes of hydrogen very rapidly, an alternative approach is to operate some of the smaller containments with inert (oxygen-deficient)
or venting.
atmospheres.


This measure, the"inerting" of a containment, provides sufficient time for ccmbustible gas control systems to become effective following a LOCA before a flammable mixture is reached in the containment.
1Section 50.44 does not require the deliberate ignition systems used by BWRs with Mark III type containmentsand PWRs with ice condenser type containments to be available during station blackout events.  The deliberate ignition systems should be available upon restoration of power.  Additional guidance concerning the availability of deliberate ignition systems during station blackout sequences is being developed as part of the staff's review of Generic Safety Issue 189, "Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen CombustionDuring a Severe Accident."Rev. 3 of RG 1.7, Page 4


Hydrogen recombiners can process the containment atmosphere at a rate of only 100 scfm per recombiner.
==C. REGULATORY POSITION==
1.Combustible Gas Control Systems The following design guidance is applicable to combustible gas control systems installedto mitigate the risk associated with combustible gas generation attributed to beyond-design-basis accidents.


Therefore, for a 300,000 cubic foot containment with a 13 volume percent hydrogen concentration that was generated in the first two minutes of the LOCA, an inordinately larger number of recombiners would be required.There are presently no other methods of combustible gas control except for purge systems, which release radioactive materials.
Structures, systems, and components (SSCs) installed to mitigate the hazard from the generation of combustible gas in containment should be designed to provide reasonable assurance that they will operate in the severe accident environment for which they are intended and over the time span for which they


For all containments, it is advisable to provide means by which combustible gases resulting from the postulated .metal-water reaction, radiolysis, and corrosion following a LOCA can be mixed, sampled, and controlled without releasing radioactive materials to the environment.
are needed. Equipment survivability expectations under severe accident conditions should consider


Since any system for combustible gas control is designated for the protection of the public in the event of an accident, the system should meet the design and construction standards of engineered safety features.Care should be taken in its design to ensure that the system itself does not introduce safety problems that may affect containment integrity;
the circumstances of applicable initiating events (such as station blackout
for example, if a flame recombiner is used, propagation of flame into the containment should be prevented.
1 or earthquakes) and the environment (including pressure, temperature, and radiation) in which the equipment is relied upon


In most reactor plants, the hydrogen control system would not be required to be operated for seven days or more following a postulated design basis LOCA. Thus, it is reasonable that hydrogen control systems need not necessarily be installed at each reactor. Provision for either onsite or offsite storage or a shared arrangement between licensees of plants in close proximity to each other may be developed.
to function. This guidance was presented in SECY-93-087, "Policy, Technical, and Licensing Issues


An example of an acceptable arrangement would be to provide at least one hydrogen control 1.7-6 system per site with the provision that a redundant unit would be available from a nearby site.C. REGULATORY
Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs" (Ref. 6).
POSITION 1. Each boiling or pressurized light-water nuclear power xeactor fueled with uranium oxide pellets within cylindrical zircaloy cladding should have the capability to measure the hydrogen concentration in the containment, mix the atmosphere in the containment, and control combusti-ble gas concentrations without relying on purging of the containment atmosphere following a LOCA.2. The continuous presence of redundant combustible gas control equipment at the site may not be necessary provided it is available on an appropriate time scale. However, appropriate design and procedural provisions should be made for its use. In addition, centralized storage facilities that would serve multiple sites may be used, provided these facilities include provisions such as maintenance, protective features, testing, and transportation for redundant units to a particular site.3. Combustible gas control systems and the provisions for mixing, measuring, and sampling should meet the design, quality assurance, redundancy, energy source, and instrumentation requirements for an engineered safety feature. In addition, the system icself should not introduce safety problem! that may affect containment integrity.
The required system performance criteria will be based on the results of design-specific reviews that include probabilistic risk assessment as required by 10 CFR 52.47(a). Because these requirements address beyond-design-basis combustible gas control, SSCs provided to meet these requirements need notbe subject to the environmental qualification requirements of 10 CFR 50.49, quality assurance requirements of Appendix B to 10 CFR Part 50, and redundancy/diversity requirements of Appendix A to 10 CFR Part 50.


The combustible gas control system should be designated Seismic Category I (see Regulatory Guide 1.29, "Seismic Design Classification"), and the Group B quality standards of Regulatory Guide 1.26, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive-Waste- Contanaing Components of Nuclear Power Plants," should be applied.4. All water-cooled power reactors should also have the installed capability for a controlled purge of the containment atmosphere to aid In cleanup. The purge or ventilation system may be a separate system or part of an existing system. It need not be redundant or be designated Seismic Category I (see Regulatory Guide 1.29), except insofar as portions of the system constitute part of the primary containment boundary or contain filters.5. The parameter values listed in Table I should be used in calculating hydrogen and oxygen gas concentrations in containments and evaluating designs provided to control and to purge combustible gases evolved in the course of loss-of-coolant accidents.
Guidance such as that found in Appendices A and B to Regulatory Guide 1.155 (Ref. 7) is appropriate


These values may be changed on the basis of additional experimental evidence and analyses.6. Materials within the containment that would yield hydrogen gas due to corrosion from the emergency cooling or containment spray solutions should be identified, and their use should be limited as much as practical.
for equipment used to mitigate the consequences of severe accidents. This guidance was used to review


1.7-7... .... ..........................  
the design of evolutionary and passive plant designs, as documented in NUREG-1462 (Ref. 8),
..............................
NUREG-1503 (Ref. 9), and NUREG-1512 (Ref. 10).The combustible gas control systems in all BWRs with Mark III-type containments and all PWRs with ice condenser type containments must meet the requirements in Section 50.4
-pV~T'~~Nfl.


.. .. -.,.D. IMPLEXENTATION
===4. The staff considers===
This guide will be used by the staff in the evaluation of l.icensut.s'
and applicants'
compliance with the specified portions of the Commisston's, regulations, including the proposed § 50.44 when it is adopted as an effec-tive regulation.


If the proposed regulations are not ndopted or are adopted in substantially different form, this guide will be withdrawn or appro-priately revised.1.7-8 Table I. Acceptable Assumptions for Evaluating the Production of Combustible Gases Following a Loss-of-Coolant Accident (LOCA)Parameter Fraction of fission product radiation energy absorbed by the coolant*Hydrogen yield rate G(H 2)Oxygen yield rate G(O 2)Acceptable Value (a) Beta (1) Betas from fission products in the fuel rods: 0 (2) Betas from fission products intimately mixed with coolant: 1.0 (b) Gamma (1) Gammas from fission products in the fuel rods, coolant in core region: 0..i**(2) Gammas from fission products intimately mixed with coolant, all coolant: 1.0 0.5 molecule/l0OEv
that the combustible gas control systems installed and approved by the NRC as of October 16, 2003, are acceptable without modification.2.Hydrogen and Oxygen Monitors2.1Hydrogen Monitors Section 50.44 requires that equipment be provided for monitoring hydrogen in the containment. The equipment for monitoring hydrogen must be functional, reliable, and capable of continuously measuring the concentration of hydrogen in the containment atmosphere following a beyond-design-basis accident
0.25 molecule/100Ev Hydrogen production is 5 times the extent of the maximum calculated reaction under 10 CFR Part 50,§50.46, or that amount that would be evolved from a core-wide average depth of reaction into the original cladding of 0.00023 inch, whichever is greater, in 2 minutes.200 mils/yr (This value should be adjusted upward for higher temperatures early in the accident sequence.)
50% of the halogens and 1% of the solids present in the core are intimately mixed with the coolant water.1.7-9 Extent and evolution time of initial core metal-water reaction hydrogen production from the cladding surrounding the fuel Aluminum corrosion rate for aluminum exposed to alkaline solutions Fission product distribution model., *. "-
"M ": ..- -I "POM Wý-. IWPTMWý .Table 1 (Continued)
Parameter Acceptable Volue All noble gases are released to the containment.


All other fission products remain lit fucl rods.Hydrogen concentration limit 4 v/o***P.AO-d~id It IMAIiUj Limit (I V/0 ttt-geii Is present. )For water, borated water, and borated alkaline solutions;
for accident management, including emergency planning. Safety-related hydrogen monitoring systems
for other solutions, data should be presented.


This fraction is thought to be conservative;
installed and approved by the NRC prior to October 16, 2003, are sufficient to meet these criteria.
further analysis may show that it should be revised.The 4 v/o hydrogen concentration limit should not be exceeded if ourning is to be avoided and if more than 5 v/o oxygen is present in the containment.


This amount may be increased to 6 v/o, with the assumption that the 2 v/o excess hydrogen would burn in the containment (if more than 5 v/o oxygen is present).  
Non-safety-related commercial-grade hydrogen monitors can also be used to meet these criteria if they
The effects of the resultant energy and burning should not create conditions exceeding the design conditions of either the containment or the safety equipment necessary to mitigate the consequences of a LOCA.Applicants and licensees should demonstrate such capability by suitable analyses and qualification test results.1.7-10 J.. ........S- ~ ~.., ~}}
 
comply with the following criteria:
Rev. 3 of RG 1.7, Page 5(1)Equipment Survivability: 
The hydrogen monitoring equipment need not be qualifiedin accordance with 10 CFR 50.49.  However, these systems are required to be functional, reliable, and capable of continuously measuring the appropriate parameter in the beyond-design-basis
 
accident environment.
 
The evaluation of survivability should consider the effects of the post-accident environmentfor the specific type of facility and monitoring system design.  The procurement for such equipment should address equipment reliability and operability in the beyond-design-basis accident
 
environmental conditions for the specific facility and monitoring system design.
 
Acceptable approaches for demonstrating equipment survivability are described in Chapter 19
 
of the ABWR FSER (Ref. 9) and the AP1000 FSER (Ref. 11).(2)Power Source:
  The instrumentation should be energized from a high-reliability power source,not necessarily standby power, and should be backed up by batteries where momentary interruption is not tolerable.(3)Quality Assurance: 
The instrumentation should be of high-quality commercial grade and should be selected to withstand the specified service environment.(4)Display and Recording: 
The instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand.
 
If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders.
 
Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.
 
Intermittent displays such as data loggers and scanning recorders may be used if no significant
 
transient response information is likely to be lost by such devices.(5)Range: 
If two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided.  If the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments should be used.(6)Servicing, Testing, and Calibration: 
Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation.  If the required interval
 
between testing is less than the normal time interval between plant shutdowns, a capability
 
for testing during power operation should be provided.
 
Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the access to such removal means.
 
The design should facilitate administrative control of the access to all setpoint adjustments, module calibration adjustments, and test points.
 
Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118, "Periodic Testing of Electric Power and
 
Protection Systems" (Ref. 12), pertaining to testing of instrument channels.  (Note:  Response
 
time testing not usually needed.)(7)Human Factors: 
The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.
 
The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications
 
potentially confusing to the operator.  Human factors analysis should be used in determining
 
the type and location of displays.
 
Rev. 3 of RG 1.7, Page 6 To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident
 
situations, instruments with which they are most familiar.(8)Direct Measurement: 
To the extent practicable, monitoring instrumentation inputs should befrom sensors that directly measure the desired variables.  An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.
 
The above provisions can be met with a program based on compliance with a pre-specified,structured program of testing and calibration; alternatively, these items can be met with a less-prescriptive, performance-based approach to assurance of the hydrogen monitoring functio
 
====n. Such an approach====
 
is consistent with SECY-00-0191, "High-Level Guidelines for Performance-Based Activities" (Ref. 13).
Specifically, assurance of the reliability, availability, and capability of the hydrogen monitoring function can be derived through tracking actual reliability performance (including calibration) against targets
 
established by the licensee based on the significance of this function, which is determined on a plant-
 
specific basis.  Thus, for hydrogen monitoring, it is acceptable to accomplish the functions of servicing, testing, and calibration within the maintenance rule program provided that applicable targets are established based on the functions of the hydrogen monitors delineated above.
 
Section 50.44 also requires that hydrogen monitors be functional.  Functional requirementscan be found in Three Mile Island (TMI) Action Plan Item II.F.1, Attachment 6, in NUREG-0737 (Ref. 14), which states that hydrogen monitors are to be functioning within 30 minutes of the initiation of safety
 
injection.  This requirement was imposed by confirmatory orders following the accident at TMI Unit 2.
 
Since that requirement was issued, the staff has determined that the 30-minute requirement can be overly burdensome.  Through the "Confirmatory Order Modifying Post-TMI Requirements Pertaining toContainment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2" (Ref. 15), the staff developeda method for licensees to adopt a risk-informed functional requirement in lieu of the 30-minute requirement.
 
As described in the confirmatory order, an acceptable functional requirement would meet the following
 
requirements:(1)Procedures shall be established for ensuring that indication of hydrogen concentration in the containment atmosphere is available in a sufficiently timely manner to support the role
 
of information in the emergency plan (and related procedures) and related activities such as
 
guidance for the severe accident management plan.(2)Hydrogen monitoring will be initiated on the basis of the following considerations:(a)The appropriate priority for establishing indication of hydrogen concentration within containment in relation to other activities in the control room.(b)The use of the indication of hydrogen concentration by decision-makers for severe accident management and emergency response.(c)Insights from experience or evaluation pertaining to possible scenarios that result in significant generation of hydrogen that would be indicative of core damage
 
or a potential threat to the integrity of the containment building.
 
The NRC staff has found that adoption of this functional requirement by licensees results in the hydrogen monitors being functional within 90 minutes after the initiation of safety injection.
 
This period of time includes equipment warm-up but not equipment calibration.
 
Rev. 3 of RG 1.7, Page 72.2Oxygen Monitors Section 50.44 requires that equipment be provided for monitoring oxygen in containments that use an inerted atmosphere for combustible gas control.  The revised rule requires the equipment for monitoring oxygen to be functional, reliable, and capable of continuously measuring the concentration of oxygen in the containment atmosphere following a beyond-design-basis accident for combustible gas
 
control and accident management, including emergency planning.  Existing oxygen monitoring systems approved by the NRC prior to October 16, 2003, are sufficient to meet these criteria.  Non-safety-related oxygen monitors would also meet these criteria if they comply with the following provisions:(1)Equipment Survivability:
The oxygen monitoring equipment need not be qualifiedin accordance with 10 CFR 50.49.  However, these systems are required to be functional, reliable, and capable of continuously measuring the appropriate parameter in the beyond-design-basis
 
accident environment.
 
The evaluation of survivability should consider the effects of the post-accident environmentfor the specific type of facility and monitoring system design.  The procurement for such equipment should address equipment reliability and operability in the beyond-design-basis accident
 
environmental conditions for the specific facility and monitoring system design.
 
Acceptable approaches for demonstrating equipment survivability are described in Chapter 19
 
of the ABWR FSER (Ref. 9) and the AP1000 FSER (Ref. 11).(2)Power Source:
The instrumentation should be energized from a high-reliability power source, not necessarily standby power, and should be backed up by batteries where momentary
 
interruption is not tolerable.(3)Channel Availability:
The out-of-service interval should be based on normal technical specification requirements on out of service for the system it serves where applicable or where
 
specified by other requirements.(4)Quality Assurance:
The recommendations of the following regulatory guides pertaining to quality assurance should be followed:*Regulatory Guide 1.28, "Quality Assurance Program Requirements (Design and Construction)" (Ref. 16)*Regulatory Guide 1.30, "Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment" (Ref. 17)*Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)"(Ref. 18)*Regulatory Guide 1.176, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Graded Quality Assurance" (Ref. 19)(5)Display and Recording:
The instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand.
 
If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders.
 
Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.
 
Intermittent displays such as data loggers and scanning recorders may be used if no significant
 
transient response information is likely to be lost by such devices.
 
Rev. 3 of RG 1.7, Page 8(6)Range:
If two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided.  If the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments should be used.(7)Interfaces:
The transmission of signals for other use should be through isolation devices that are designated as part of the monitoring instrumentation and that meet the provisions
 
of the criteria presented here.(8)Servicing, Testing, and Calibration:
Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation.  If the required interval
 
between testing is less than the normal time interval between plant shutdowns, a capability
 
for testing during power operation should be provided.
 
Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the access to such removal means.
 
The design should facilitate administrative control of the access to all setpoint adjustments, module calibration adjustments, and test points.Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118, "Periodic Testing of Electric Power
 
and Protection Systems," (Ref. 12) pertaining to testing of instrument channels.
 
(Note:  Response time testing not usually needed.)
The location of the isolation device should be such that it would be accessible for maintenance during accident conditions.(9)Human Factors:
The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.
 
The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications
 
potentially confusing to the operator.  Human factors analysis should be used in determining
 
the type and location of displays.
 
To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident
 
situations, instruments with which they are most familiar.(10)Direct Measurement:
To the extent practicable, monitoring instrumentation inputs should befrom sensors that directly measure the desired variables.  An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.3.Atmosphere Mixing Systems Section 50.44 requires that all containments have a capability for ensuring a mixed atmosphere.
 
This capability may be provided by an active, passive, or combination syste
 
====m. Active systems may====
 
consist of a fan, a fan cooler, or containment spray.  For passive or combination systems that use
 
convective mixing to mix the combustible gases, the containment internal structures should have
 
design features that promote the free circulation of the atmosphere.
 
2The NRC staff believes that current lumped parameter analytical codes may overestimate mixing processes(in particular, natural convection).  Applicants should substantiate the applicability of these codes to their analysesthrough sensitivity studies, validation with data, or other means.Rev. 3 of RG 1.7, Page 9 All containment types should have an analysis of the effectiveness of the method used for providing a mixed atmosphere.  This analysis should demonstrate that combustible gases will not
 
accumulate within a compartment or cubicle to form a combustible or detonable mixture that could
 
cause loss of containment integrity.
 
2 Atmosphere mixing systems prevent local accumulation of combustible or detonable gasesthat could threaten containment integrity or equipment operating in a local compartment.  Active systems installed to mitigate this threat should be reliable, redundant, single-failure-proof, able to be tested
 
and inspected, and remain operable with a loss of onsite or offsite power.  The NRC staff considers
 
atmosphere mixing systems installed and approved by the NRC as of October 16, 2003, to be acceptable
 
without modification.
 
References 20 through 23 provide important insights into the potential for detonation of hydrogen-air mixtures.4.Hydrogen Gas ProductionMaterials within the containment that would yield hydrogen gas by corrosion from the emergency cooling or containment spray solutions should be identified, and their use should be limited as much as
 
practicable.5.Containment Structural Integrity Section 50.44 requires that containment structural integrity be demonstrated by use of an analytical technique that is accepted by the NRC staff.  This demonstration must include sufficient
 
supporting justification to show that the technique describes the containment response to the structural
 
loads involved.  The following criteria of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 24) provide an acceptable method for demonstrating that the requirements are met:(1)Steel containments meet the requirements of the ASME Boiler and Pressure Vessel Code (edition and addenda as incorporated by reference in 10 CFR 50.55a(b)(1)), Section III,
Division 1, Subsubarticle NE - 3220, Service Level C Limits, considering pressure
 
and dead load alone (evaluation of instability is not required).(2)Concrete containments meet the requirements of the ASME Boiler and Pressure Vessel Code, Section III, Division 2, Subsubarticle CC - 3720, Factored Load Category, considering pressure
 
and dead load alone.
 
As a minimum, the specific code requirements set forth for each type of containment should be met for a combination of dead load and an internal pressure of 45 psig. The staff will consider modest
 
deviations from these criteria, if the applicant shows good cause.
 
These criteria, which no longer are contained in Section 50.44, remain acceptable to the NRC
staff for meeting the current regulations.  The acceptability of licensee analyses using the ASME Code
 
criteria remains unaffected by this rulemaking.
 
Rev. 3 of RG 1.7, Page 10
 
==D. IMPLEMENTATION==
The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide.  No backfitting is intended or approved
 
in connection with the issuance of this guide.
 
Except in those cases in which an applicant or licensee proposes or has previously established an acceptable alternative method for complying with the specified portions of the NRC's regulations, the NRC staff will use the methods described in this guide to evaluate (1) submittals in connection with
 
applications for construction permits, standard plant design certifications, operating licenses, early site
 
permits, and combined licenses; and (2) submittals from operating reactor licensees who voluntarily
 
propose to initiate system modifications that have a clear nexus with the subject for which guidance
 
is provided herein.
 
REGULATORY ANALYSIS / BACKFIT ANALYSISA separate regulatory analysis was not prepared for this guide.  The regulatory analysis preparedfor the revision of 10 CFR 50.44, "Standards for Combustible Gas Control System in Light-Water-CooledPower Reactors" (Ref. 25), provides the regulatory basis for this guide and examines the costs and benefits for the rule as implemented by this guide.
 
The backfit analysis for this regulatory guide is available in Draft Regulatory Guide DG-1117,"Control of Combustible Gas Concentrations in Containment" (Ref. 26).  The NRC issued DG-1117
 
in August 2002 to solicit public comment on the draft of this Revision 3 of Regulatory Guide 1.7.
 
3 All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are availableelectronically through the Federal Register Main Page of the public GPOAccess Web site, which the U.S. Government Printing Office maintains at http://www.gpoaccess.gov/fr/index.html.  In addition, 68 FR 54123 is available electronically through the Electronic Reading Room on the NRC's public Web site at http://www.nrc.gov/
reading-rm/doc-collections/cfr/fr/2003/20030916.pdf.  Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR's mailing address
 
is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548;
 
email PDR@nrc.gov
.4 All NRC regulations listed herein are available electronically through the Public Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/cfr/.  Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD;
 
the PDR's mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;
 
fax (301) 415-3548; email PDR@nrc.gov
.5 All Commission papers (SECYs) listed herein were published by the U.S. Nuclear Regulatory Commission, and are available electronically through the Public Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/commission/secys/.  Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR's
 
mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;
 
fax (301) 415-3548; email PDR@nrc.gov
.6 Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC
20402-9328 (telephone 202-512-1800); or from the National Technical Information Service (NTIS) by writing NTIS
 
at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov , by telephone at (800) 553-NTIS(6847) or (703)605-6000, or by fax to (703) 605-6900.  Copies are also available for inspection or copying for a fee
 
from the NRC's Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the
 
PDR's mailing address is USNRC PDR, Washington, DC 20555-0001.  The PDR can also be reached by telephone
 
at (301) 415-4737 or (800)397-4209, by fax at (301)415-3548, and by email to PDR@nrc.gov.  In addition, NUREG-0737 and NUREG-1793 are available electronically through the Electronic Reading Room on the NRC's
 
Public Web site http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/
.Rev. 3 of RG 1.7, Page 11 REFERENCES
1.Federal Register , "Combustible Gas Control in Containment," Volume 68, No. 179, pp. 54123-
54142, U.S. Nuclear Regulatory Commission, Washington, DC, September 16, 2003.
 
32.10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities,"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
43.10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants,"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
44.SECY-00-0198, "Status Report on Study of Risk-Informed Changes to the Technical Requirementsof 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to 10 CFR 50.44 (Combustible Gas Control)," U.S. Nuclear Regulatory Commission, Washington, DC,
September 14, 2000.
 
55.NUREG-1150, "Severe Accident Risks:  An Assessment for Five U.S. Nuclear Power Plants,"
U.S. Nuclear Regulatory Commission, Washington, DC, December 1990.
 
66.SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs," U.S. Nuclear Regulatory Commission, Washington, DC, April 2, 1993.
 
5
7 All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission or its predecessor, the U.S. Atomic Energy Commission.  Most are available electronically through the Electronic Reading Room
 
on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/.  Single copies of regulatory guides may also be obtained free of charge by writing the Reproduction and Distribution Services Section, ADM, USNRC, Washington, DC 20555-0001, by fax to (301) 415-2289, or by email to DISTRIBUTION@nrc.gov. Active guides may also be purchased from the National Technical Information Service (NTIS).  Details may be
 
obtained by contacting NTIS at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov
,by telephone at (800) 553-NTIS (6847) or (703) 605-6000, or by fax to (703) 605-6900.  Copies are also available for
 
inspection or copying for a fee from the NRC's Public Document Room (PDR), which is located at 11555 Rockville
 
Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-000
 
===1. The PDR===
 
can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email
 
to PDR@nrc.gov
.8 Copies are available electronically through the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html , under Accession #ML021270103.Rev. 3 of RG 1.7, Page 127.Regulatory Guide 1.155, "Station Blackout," U.S. Nuclear Regulatory Commission, Washington, DC.
 
78.NUREG-1462, "Final Safety Evaluation Report Related to the Certification of the System 80+
Design, Docket No. 52-002" U.S. Nuclear Regulatory Commission, Washington, DC,
August 1994.
 
69.NUREG-1503, "Final Safety Evaluation Report Related to the Certification of the Advanced Boiling-Water Reactor Design, Docket No. 52-001," U.S. Nuclear Regulatory Commission, Washington, DC, July 1994.
 
610.NUREG-1512, "Final Safety Evaluation Report Related to the Certification of the AP600Standard Design, Docket No. 52-003," U.S. Nuclear Regulatory Commission, Washington, DC, September 1998.
 
611.NUREG-1793, "Final Safety Evaluation Report Related to the Certification of the AP1000Standard Design, Docket No. 52-006," U.S. Nuclear Regulatory Commission, Washington, DC, September 2004.
 
612.Regulatory Guide 1.118, "Periodic Testing of Electric Power and Protection Systems,"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
713.SECY-00-0191, "High-Level Guidelines for Performance-Based Activities,"
U.S. Nuclear Regulatory Commission, Washington, DC, September 1, 2000.
 
514.NUREG-0737, "Clarification of TMI Action Plan Requirements," U.S. Nuclear Regulatory Commission, Washington, DC, November 1980.
 
615."Confirmatory Order Modifying Post-TMI Requirements Pertaining to Containment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2," U.S. Nuclear Regulatory Commission, Washington, DC, September 28, 1998.
 
816.Regulatory Guide 1.28, "Quality Assurance Program Requirements (Design and Construction),"
U.S. Nuclear Regulatory Commission, Washington DC.
 
717.Regulatory Guide 1.30, "Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment," U.S. Nuclear Regulatory Commission, Washington, DC.
 
7
9 Copies may be purchased from the American Society of Mechanical Engineers, Three Park Avenue, NewYork, NY
10016-5990; phone (212) 591-8500; fax (212) 591-8501;
www.asme.org
.10 This regulatory analysis is available electronically under Accession #ML031640482 in the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html , and through the Public Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/
doc-collections/commission/secys/2003/secy2003-0127/2003-0127scy.pdf#pagemode=bookmarks.  Copies are also available for inspection or copying for a fee from the NRC's Public Document Room (PDR), which is located at
11555 Rockville Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001.
 
The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email to PDR@nrc.gov
.11 Draft Regulatory Guide DG-1117 is available electronically under Accession #ML022210067 in the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room (PDR), which is
 
located at 11555 Rockville Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC
 
20555-0001.  The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax
 
at (301) 415-3548, and by email to PDR@nrc.gov
.Rev. 3 of RG 1.7, Page 1318.Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation),"
U.S. Nuclear Regulatory Commission, Washington, DC.
 
719.Regulatory Guide 1.176, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:
Graded Quality Assurance," U.S. Nuclear Regulatory Commission, Washington, DC.
 
720.NUREG/CR-4905, "Detonability of H-Air-Diluent Mixtures," prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC, June 1987.
 
621.NUREG/CR-4961, "A Summary of Hydrogen-Air Detonation Experiments," prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC,
June 1987.
 
622.NUREG/CR-5275, "Flame Facility" (The Effect of Obstacles and Transverse Venting on Flame Acceleration and Transition to Detonation of Hydrogen-Air Mixtures at Large Scale), prepared
 
by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC,
April 1989.
 
623.NUREG/CR-5525, "Hydrogen-Air-Diluent Detonation Study of Nuclear Reactor Safety Analyses," prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory
 
Commission, Washington, DC, December 1990.
 
624.ASME Boiler and Pressure Vessel Code, Section III, "Rules for Construction of Nuclear Power Plant Components," American Society of Mechanical Engineers, New York, NY, 1992.
 
925."Final Regulatory Analysis for 50.44," Attachment 4 to SECY-03-0127,U.S. Nuclear Regulatory Commission, Washington, DC,  July 2003.
 
1026.Draft Regulatory Guide DG-1117, "Control of Combustible Gas Concentrations in Containment."
U.S. Nuclear Regulatory Commission, Washington, DC, August 2002.
 
11}}


{{RG-Nav}}
{{RG-Nav}}

Revision as of 01:43, 25 October 2018

Control of Combustible Gas Concentrations in Containment
ML070290080
Person / Time
Issue date: 03/23/2007
From: Pulsipher J
NRC/NRO/DSRA
To:
Pulsipher J C, NRO/DSRA, 415-2811
Shared Package
ML070290076 List:
References
DG-1117 RG-1.007, Rev. 3
Download: ML070290080 (13)
v • d • e


The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staffconsiders acceptable for use in implementing specific parts of the agency's regulations, techniques that the staff uses in evaluating specific problemsor postulated accidents, and data that the staff need in reviewing applications for permits and licenses. Regulatory guides are not substitutesfor regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemedacceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Comm ission.This guide was issued after consideration of comments received from the public. The NRC staff encourages and welcomes comments and suggestionsin connection with improvements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed. The NRC staff will revise existing guides, as appropriate, to accommodate comments and to reflect new information or experience. Written commentsmay be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.Regulatory guides are issued in 10 broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities;4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review;and 10, General.Requests for single copies of draft or active regulatory guides (which may be reproduced) should be made to the U.S. Nuclear Regulatory Commission,Washington, DC 20555, Attention: Reproduction and Distribution Services Section, or by fax to (301) 415-2289; or by email to Distribution@nrc.gov. Electronic copies of this guide and other recently issued guides are available thr ough the NRC's public Web site under the Regulatory Guides documentcollection of the NRC's El ectronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/

and through the NRC's Agencywide DocumentsAccess and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070290080.U.S. NUCLEAR REGULATORY COMMISSIONMarch 2007Revision 3 REGULATORY GUIDEOFFICE OF NUCLEAR REGULATORY RESEARCHREGULATORY GUIDE 1.7(Draft was issued as DG-1117, dated August 2002)CONTROL OF COMBUSTIBLE GAS CONCENTRATIONSIN CONTAINMENT

A. INTRODUCTION

In September 2003, the U.S. Nuclear Regulatory Commission (NRC) issued a revision ofSection 50.44, "Combustible Gas Control for Nuclear Power Reactors" (Ref. 1), which amended Title 10, Part 50, of the Code of Federal Regulations (10 CFR Part 50), "Domestic Licensing of Production and Utilization Facilities" (Ref. 2). This regulation is applicable to all reactor construction permits or operating licenses under 10 CFR Part 50, except for those facilities for which the certifications requiredunder Section 50.82(a)(1) have been submitted, and to all reactor design approvals, design certifications,combined licenses or manufacturing licenses under 10 CFR Part 52, "Early Site Permits; Standard DesignCertifications; and Combined Licenses for Nuclear Power Plants" (Ref. 3). This regulatory guide describes methods that are acceptable to the NRC staff for implementing the revised Section 50.44 for reactors, subject to the provisions of Sections 50.44(b) or 50.44(c).

This regulatory guide relates to information collections that are covered by the requirements of 10 CFR Parts 50 and 52, which the Office of Management and Budget (OMB) approved under OMB

control numbers 3150-0011 and 3150-0151, respectively. The NRC may neither conduct nor sponsor, and a person is not required to respond to, an information collection request or requirement

unless the requesting document displays a currently valid OMB control number.

Rev. 3 of RG 1.7, Page 2

B. DISCUSSION

Section 50.44 provides requirements for the mitigation of combustible gas generated bya beyond-design-basis accident. In existing light-water reactors, the principal combustible gas is hydrogen.In an accident more severe than the design-basis loss-of-coolant accident (LOCA), combustible gas is predominately generated within the containment as a result of the following factors:(1)fuel clad-coolant reaction between the fuel cladding and the reactor coolant

(2)molten core-concrete interaction in a severe core melt sequence with a failed reactor vessel If a sufficient amount of combustible gas is generated, it may react with oxygen present in the containment at a rate rapid enough to lead to a containment breach or a leakage rate in excess of

technical specification limits. Additionally, damage to systems and components essential to continued

control of the post-accident conditions could occur.

In SECY-00-0198, "Status Report on Study of Risk-Informed Changes to the Technical Requirements of 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to

10 CFR 50.44 (Combustible Gas Control)" (Ref. 4), the NRC staff recommended changes to 10 CFR 50.44 that reflect the position that only combustible gas generated by a beyond-design-basis accident is a risk-significant threat to containment integrity. Based on those recommendations, the September 2003 revision of 10 CFR 50.44 eliminates requirements that pertain only to design-basis LOCAs.

Attachment 2 to SECY-00-0198 (Ref. 4) used the framework described in Attachment 1 to the paper with risk insights from NUREG-1150 (Ref. 5) and the integrated plant evaluation programs

to evaluate the requirements in 10 CFR 50.44. In so doing, Attachment 2 noted that containment types

that rely on pressure suppression concepts (i.e., ice baskets or water pools) to condense the steam from a design-basis LOCA have smaller containment volumes, and in some cases lower design pressures, than pressurized-water reactor (PWR) large-volume or subatmospheric containments. Consequently, the smaller volumes and lower design pressures associ ated with pressure suppression containment designs make them more vulnerable to combustible gas deflagrations during degraded core accidents because

the pressure loads could cause structural failure of the containment. Also, because of the smaller volume of these containments, detonable mixtures could be formed. A detonation would impose a dynamic pressure load on the containment structure that could be more severe than the static load from an equivalent deflagration. However, the staff noted in SECY-00-0198 that the risk of early containment failure

from combustible gas combustion in these types of containments can be limited by the use of mitigative

features: (1) inerting in Mark I and II containments and (2) using igniter systems in Mark III

and ice condenser containments. As a result, the revised Section 50.44 has the following requirements:(1)All boiling-water reactor (BWR) Mark I and II type containments must be inerted. By maintaining an oxygen-deficient atmosphere, combustible gas combustion that could threaten containment

integrity is prevented.(2)All BWRs with Mark III type containments and all PWRs with ice condenser type containments must have the capability to control combustible gas generated from a metal-water reaction

involving 75% of the fuel cladding surrounding the active fuel region (excluding the cladding

surrounding the plenum volume) so that there is no loss of containment structural integrity.

The deliberate ignition systems provided to meet this existing combustible gas source term

are capable of safely accommodating even greater amounts of combustible gas associated with

even more severe core melt sequences that fail the reactor vessel and involve molten

core-concrete interaction. Deliberate ignition systems, if available, generally consume the combustible gas before it reaches concentrations that can be detrimental to containment integrity.

Rev. 3 of RG 1.7, Page 3(3)For all applicants for and holders of a water-cooled reactor construction permit or operating license under 10 CFR Part 50, and all applicants for a light-water reactor design approval, or design certification, or combined license under 10 CFR Part 52 that are docketed after October 16, 2003, the effective date of the rule, the following requirements apply. All containments must have an inerted atmosphere or limit combustible gas concentrations in containment during and following an accident that releases an equivalent amount of combustible gas as would be generated from a 100% fuel-clad coolant reaction, uniformly distributed, to less than 10%

(by volume) and must maintain containment structural integrity. The requirements of this paragraphapply only to water-cooled reactor designs with characteristics (e.g., type and quantity of claddingmaterials) such that the potential for production of combustible gases is comparable to light-water reactor designs licensed as of October 16, 2003.(4)For all construction permits and operating licenses under 10 CFR Part 50, and all design approvals,design certifications, combined licenses, or manufacturing licenses under Part 52, for non-water-cooled reactors and water-cooled reactors that do not fall within the description in paragraph 3 (above), any of which are issued after October 16, 2003, applications subject to this paragraph

must include the following:(a)information addressing whether accidents involving combustible gases are technically relevant for their design(b)if accidents involving combustible gases are found to be technically relevant, information demonstrating that the safety impacts of combustible gases during

design-basis and significant beyond-design-basis accidents have been addressed to ensure adequate protection of public health and safety and common defense and security.

The combustible gas control systems, the atmosphere mixing systems, and the provisionsfor measuring and sampling that are required by Section 50.44 are risk-significant, as they have the abilityto mitigate the risk associated with combustible gas generation caused by significant beyond-design-basis accidents. The recommended treatments for those systems are delineated in the regulatory position

in Section C of this regulatory guide.The hydrogen monitors should be able to assess the degree of core damage during a beyond-design-basis accident and confirm that random or deliberate ignition has taken place. Hydrogen monitors, in conjunction with oxygen monitors, are further relied on to implement severe accident management

strategies to address a potential breach of containment integrity or to consider containment purging

or venting.

1Section 50.44 does not require the deliberate ignition systems used by BWRs with Mark III type containmentsand PWRs with ice condenser type containments to be available during station blackout events. The deliberate ignition systems should be available upon restoration of power. Additional guidance concerning the availability of deliberate ignition systems during station blackout sequences is being developed as part of the staff's review of Generic Safety Issue 189, "Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen CombustionDuring a Severe Accident."Rev. 3 of RG 1.7, Page 4

C. REGULATORY POSITION

1.Combustible Gas Control Systems The following design guidance is applicable to combustible gas control systems installedto mitigate the risk associated with combustible gas generation attributed to beyond-design-basis accidents.

Structures, systems, and components (SSCs) installed to mitigate the hazard from the generation of combustible gas in containment should be designed to provide reasonable assurance that they will operate in the severe accident environment for which they are intended and over the time span for which they

are needed. Equipment survivability expectations under severe accident conditions should consider

the circumstances of applicable initiating events (such as station blackout

1 or earthquakes) and the environment (including pressure, temperature, and radiation) in which the equipment is relied upon

to function. This guidance was presented in SECY-93-087, "Policy, Technical, and Licensing Issues

Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs" (Ref. 6).

The required system performance criteria will be based on the results of design-specific reviews that include probabilistic risk assessment as required by 10 CFR 52.47(a). Because these requirements address beyond-design-basis combustible gas control, SSCs provided to meet these requirements need notbe subject to the environmental qualification requirements of 10 CFR 50.49, quality assurance requirements of Appendix B to 10 CFR Part 50, and redundancy/diversity requirements of Appendix A to 10 CFR Part 50.

Guidance such as that found in Appendices A and B to Regulatory Guide 1.155 (Ref. 7) is appropriate

for equipment used to mitigate the consequences of severe accidents. This guidance was used to review

the design of evolutionary and passive plant designs, as documented in NUREG-1462 (Ref. 8),

NUREG-1503 (Ref. 9), and NUREG-1512 (Ref. 10).The combustible gas control systems in all BWRs with Mark III-type containments and all PWRs with ice condenser type containments must meet the requirements in Section 50.4

4. The staff considers

that the combustible gas control systems installed and approved by the NRC as of October 16, 2003, are acceptable without modification.2.Hydrogen and Oxygen Monitors2.1Hydrogen Monitors Section 50.44 requires that equipment be provided for monitoring hydrogen in the containment. The equipment for monitoring hydrogen must be functional, reliable, and capable of continuously measuring the concentration of hydrogen in the containment atmosphere following a beyond-design-basis accident

for accident management, including emergency planning. Safety-related hydrogen monitoring systems

installed and approved by the NRC prior to October 16, 2003, are sufficient to meet these criteria.

Non-safety-related commercial-grade hydrogen monitors can also be used to meet these criteria if they

comply with the following criteria:

Rev. 3 of RG 1.7, Page 5(1)Equipment Survivability:

The hydrogen monitoring equipment need not be qualifiedin accordance with 10 CFR 50.49. However, these systems are required to be functional, reliable, and capable of continuously measuring the appropriate parameter in the beyond-design-basis

accident environment.

The evaluation of survivability should consider the effects of the post-accident environmentfor the specific type of facility and monitoring system design. The procurement for such equipment should address equipment reliability and operability in the beyond-design-basis accident

environmental conditions for the specific facility and monitoring system design.

Acceptable approaches for demonstrating equipment survivability are described in Chapter 19

of the ABWR FSER (Ref. 9) and the AP1000 FSER (Ref. 11).(2)Power Source:

The instrumentation should be energized from a high-reliability power source,not necessarily standby power, and should be backed up by batteries where momentary interruption is not tolerable.(3)Quality Assurance:

The instrumentation should be of high-quality commercial grade and should be selected to withstand the specified service environment.(4)Display and Recording:

The instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand.

If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders.

Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.

Intermittent displays such as data loggers and scanning recorders may be used if no significant

transient response information is likely to be lost by such devices.(5)Range:

If two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided. If the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments should be used.(6)Servicing, Testing, and Calibration:

Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation. If the required interval

between testing is less than the normal time interval between plant shutdowns, a capability

for testing during power operation should be provided.

Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the access to such removal means.

The design should facilitate administrative control of the access to all setpoint adjustments, module calibration adjustments, and test points.

Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118, "Periodic Testing of Electric Power and

Protection Systems" (Ref. 12), pertaining to testing of instrument channels. (Note: Response

time testing not usually needed.)(7)Human Factors:

The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.

The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications

potentially confusing to the operator. Human factors analysis should be used in determining

the type and location of displays.

Rev. 3 of RG 1.7, Page 6 To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident

situations, instruments with which they are most familiar.(8)Direct Measurement:

To the extent practicable, monitoring instrumentation inputs should befrom sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.

The above provisions can be met with a program based on compliance with a pre-specified,structured program of testing and calibration; alternatively, these items can be met with a less-prescriptive, performance-based approach to assurance of the hydrogen monitoring functio

n. Such an approach

is consistent with SECY-00-0191, "High-Level Guidelines for Performance-Based Activities" (Ref. 13).

Specifically, assurance of the reliability, availability, and capability of the hydrogen monitoring function can be derived through tracking actual reliability performance (including calibration) against targets

established by the licensee based on the significance of this function, which is determined on a plant-

specific basis. Thus, for hydrogen monitoring, it is acceptable to accomplish the functions of servicing, testing, and calibration within the maintenance rule program provided that applicable targets are established based on the functions of the hydrogen monitors delineated above.

Section 50.44 also requires that hydrogen monitors be functional. Functional requirementscan be found in Three Mile Island (TMI) Action Plan Item II.F.1, Attachment 6, in NUREG-0737 (Ref. 14), which states that hydrogen monitors are to be functioning within 30 minutes of the initiation of safety

injection. This requirement was imposed by confirmatory orders following the accident at TMI Unit 2.

Since that requirement was issued, the staff has determined that the 30-minute requirement can be overly burdensome. Through the "Confirmatory Order Modifying Post-TMI Requirements Pertaining toContainment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2" (Ref. 15), the staff developeda method for licensees to adopt a risk-informed functional requirement in lieu of the 30-minute requirement.

As described in the confirmatory order, an acceptable functional requirement would meet the following

requirements:(1)Procedures shall be established for ensuring that indication of hydrogen concentration in the containment atmosphere is available in a sufficiently timely manner to support the role

of information in the emergency plan (and related procedures) and related activities such as

guidance for the severe accident management plan.(2)Hydrogen monitoring will be initiated on the basis of the following considerations:(a)The appropriate priority for establishing indication of hydrogen concentration within containment in relation to other activities in the control room.(b)The use of the indication of hydrogen concentration by decision-makers for severe accident management and emergency response.(c)Insights from experience or evaluation pertaining to possible scenarios that result in significant generation of hydrogen that would be indicative of core damage

or a potential threat to the integrity of the containment building.

The NRC staff has found that adoption of this functional requirement by licensees results in the hydrogen monitors being functional within 90 minutes after the initiation of safety injection.

This period of time includes equipment warm-up but not equipment calibration.

Rev. 3 of RG 1.7, Page 72.2Oxygen Monitors Section 50.44 requires that equipment be provided for monitoring oxygen in containments that use an inerted atmosphere for combustible gas control. The revised rule requires the equipment for monitoring oxygen to be functional, reliable, and capable of continuously measuring the concentration of oxygen in the containment atmosphere following a beyond-design-basis accident for combustible gas

control and accident management, including emergency planning. Existing oxygen monitoring systems approved by the NRC prior to October 16, 2003, are sufficient to meet these criteria. Non-safety-related oxygen monitors would also meet these criteria if they comply with the following provisions:(1)Equipment Survivability:

The oxygen monitoring equipment need not be qualifiedin accordance with 10 CFR 50.49. However, these systems are required to be functional, reliable, and capable of continuously measuring the appropriate parameter in the beyond-design-basis

accident environment.

The evaluation of survivability should consider the effects of the post-accident environmentfor the specific type of facility and monitoring system design. The procurement for such equipment should address equipment reliability and operability in the beyond-design-basis accident

environmental conditions for the specific facility and monitoring system design.

Acceptable approaches for demonstrating equipment survivability are described in Chapter 19

of the ABWR FSER (Ref. 9) and the AP1000 FSER (Ref. 11).(2)Power Source:

The instrumentation should be energized from a high-reliability power source, not necessarily standby power, and should be backed up by batteries where momentary

interruption is not tolerable.(3)Channel Availability:

The out-of-service interval should be based on normal technical specification requirements on out of service for the system it serves where applicable or where

specified by other requirements.(4)Quality Assurance:

The recommendations of the following regulatory guides pertaining to quality assurance should be followed:*Regulatory Guide 1.28, "Quality Assurance Program Requirements (Design and Construction)" (Ref. 16)*Regulatory Guide 1.30, "Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment" (Ref. 17)*Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation)"(Ref. 18)*Regulatory Guide 1.176, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Graded Quality Assurance" (Ref. 19)(5)Display and Recording:

The instrumentation signal may be displayed on an individual instrument or it may be processed for display on demand.

If direct and immediate trend or transient information is essential for operator information or action, the recording should be continuously available on redundant dedicated recorders.

Otherwise, it may be continuously updated, stored in computer memory, and displayed on demand.

Intermittent displays such as data loggers and scanning recorders may be used if no significant

transient response information is likely to be lost by such devices.

Rev. 3 of RG 1.7, Page 8(6)Range:

If two or more instruments are needed to cover a particular range, overlapping of instrument span should be provided. If the required range of monitoring instrumentation results in a loss of instrumentation sensitivity in the normal operating range, separate instruments should be used.(7)Interfaces:

The transmission of signals for other use should be through isolation devices that are designated as part of the monitoring instrumentation and that meet the provisions

of the criteria presented here.(8)Servicing, Testing, and Calibration:

Servicing, testing, and calibration programs should be specified to maintain the capability of the monitoring instrumentation. If the required interval

between testing is less than the normal time interval between plant shutdowns, a capability

for testing during power operation should be provided.

Whenever means for removing channels from service are included in the design, the design should facilitate administrative control of the access to such removal means.

The design should facilitate administrative control of the access to all setpoint adjustments, module calibration adjustments, and test points.Periodic checking, testing, calibration, and calibration verification should be in accordance with the applicable portions of Regulatory Guide 1.118, "Periodic Testing of Electric Power

and Protection Systems," (Ref. 12) pertaining to testing of instrument channels.

(Note: Response time testing not usually needed.)

The location of the isolation device should be such that it would be accessible for maintenance during accident conditions.(9)Human Factors:

The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.

The monitoring instrumentation design should minimize the development of conditions that would cause meters, annunciators, recorders, alarms, etc., to give anomalous indications

potentially confusing to the operator. Human factors analysis should be used in determining

the type and location of displays.

To the extent practicable, the same instruments should be used for accident monitoring as are used for the normal operations of the plant to enable the operators to use, during accident

situations, instruments with which they are most familiar.(10)Direct Measurement:

To the extent practicable, monitoring instrumentation inputs should befrom sensors that directly measure the desired variables. An indirect measurement should be made only when it can be shown by analysis to provide unambiguous information.3.Atmosphere Mixing Systems Section 50.44 requires that all containments have a capability for ensuring a mixed atmosphere.

This capability may be provided by an active, passive, or combination syste

m. Active systems may

consist of a fan, a fan cooler, or containment spray. For passive or combination systems that use

convective mixing to mix the combustible gases, the containment internal structures should have

design features that promote the free circulation of the atmosphere.

2The NRC staff believes that current lumped parameter analytical codes may overestimate mixing processes(in particular, natural convection). Applicants should substantiate the applicability of these codes to their analysesthrough sensitivity studies, validation with data, or other means.Rev. 3 of RG 1.7, Page 9 All containment types should have an analysis of the effectiveness of the method used for providing a mixed atmosphere. This analysis should demonstrate that combustible gases will not

accumulate within a compartment or cubicle to form a combustible or detonable mixture that could

cause loss of containment integrity.

2 Atmosphere mixing systems prevent local accumulation of combustible or detonable gasesthat could threaten containment integrity or equipment operating in a local compartment. Active systems installed to mitigate this threat should be reliable, redundant, single-failure-proof, able to be tested

and inspected, and remain operable with a loss of onsite or offsite power. The NRC staff considers

atmosphere mixing systems installed and approved by the NRC as of October 16, 2003, to be acceptable

without modification.

References 20 through 23 provide important insights into the potential for detonation of hydrogen-air mixtures.4.Hydrogen Gas ProductionMaterials within the containment that would yield hydrogen gas by corrosion from the emergency cooling or containment spray solutions should be identified, and their use should be limited as much as

practicable.5.Containment Structural Integrity Section 50.44 requires that containment structural integrity be demonstrated by use of an analytical technique that is accepted by the NRC staff. This demonstration must include sufficient

supporting justification to show that the technique describes the containment response to the structural

loads involved. The following criteria of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref. 24) provide an acceptable method for demonstrating that the requirements are met:(1)Steel containments meet the requirements of the ASME Boiler and Pressure Vessel Code (edition and addenda as incorporated by reference in 10 CFR 50.55a(b)(1)),Section III,

Division 1, Subsubarticle NE - 3220, Service Level C Limits, considering pressure

and dead load alone (evaluation of instability is not required).(2)Concrete containments meet the requirements of the ASME Boiler and Pressure Vessel Code,Section III, Division 2, Subsubarticle CC - 3720, Factored Load Category, considering pressure

and dead load alone.

As a minimum, the specific code requirements set forth for each type of containment should be met for a combination of dead load and an internal pressure of 45 psig. The staff will consider modest

deviations from these criteria, if the applicant shows good cause.

These criteria, which no longer are contained in Section 50.44, remain acceptable to the NRC

staff for meeting the current regulations. The acceptability of licensee analyses using the ASME Code

criteria remains unaffected by this rulemaking.

Rev. 3 of RG 1.7, Page 10

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding the NRC staff's plans for using this regulatory guide. No backfitting is intended or approved

in connection with the issuance of this guide.

Except in those cases in which an applicant or licensee proposes or has previously established an acceptable alternative method for complying with the specified portions of the NRC's regulations, the NRC staff will use the methods described in this guide to evaluate (1) submittals in connection with

applications for construction permits, standard plant design certifications, operating licenses, early site

permits, and combined licenses; and (2) submittals from operating reactor licensees who voluntarily

propose to initiate system modifications that have a clear nexus with the subject for which guidance

is provided herein.

REGULATORY ANALYSIS / BACKFIT ANALYSISA separate regulatory analysis was not prepared for this guide. The regulatory analysis preparedfor the revision of 10 CFR 50.44, "Standards for Combustible Gas Control System in Light-Water-CooledPower Reactors" (Ref. 25), provides the regulatory basis for this guide and examines the costs and benefits for the rule as implemented by this guide.

The backfit analysis for this regulatory guide is available in Draft Regulatory Guide DG-1117,"Control of Combustible Gas Concentrations in Containment" (Ref. 26). The NRC issued DG-1117

in August 2002 to solicit public comment on the draft of this Revision 3 of Regulatory Guide 1.7.

3 All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are availableelectronically through the Federal Register Main Page of the public GPOAccess Web site, which the U.S. Government Printing Office maintains at http://www.gpoaccess.gov/fr/index.html. In addition, 68 FR 54123 is available electronically through the Electronic Reading Room on the NRC's public Web site at http://www.nrc.gov/

reading-rm/doc-collections/cfr/fr/2003/20030916.pdf. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR's mailing address

is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209; fax (301) 415-3548;

email PDR@nrc.gov

.4 All NRC regulations listed herein are available electronically through the Public Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/cfr/. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD;

the PDR's mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;

fax (301) 415-3548; email PDR@nrc.gov

.5 All Commission papers (SECYs) listed herein were published by the U.S. Nuclear Regulatory Commission, and are available electronically through the Public Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/commission/secys/. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDR's

mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;

fax (301) 415-3548; email PDR@nrc.gov

.6 Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC

20402-9328 (telephone 202-512-1800); or from the National Technical Information Service (NTIS) by writing NTIS

at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov , by telephone at (800) 553-NTIS(6847) or (703)605-6000, or by fax to (703) 605-6900. Copies are also available for inspection or copying for a fee

from the NRC's Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the

PDR's mailing address is USNRC PDR, Washington, DC 20555-0001. The PDR can also be reached by telephone

at (301) 415-4737 or (800)397-4209, by fax at (301)415-3548, and by email to PDR@nrc.gov. In addition, NUREG-0737 and NUREG-1793 are available electronically through the Electronic Reading Room on the NRC's

Public Web site http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/

.Rev. 3 of RG 1.7, Page 11 REFERENCES

1.Federal Register , "Combustible Gas Control in Containment," Volume 68, No. 179, pp. 54123-

54142, U.S. Nuclear Regulatory Commission, Washington, DC, September 16, 2003.

32.10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities,"

U.S. Nuclear Regulatory Commission, Washington, DC.

43.10 CFR Part 52, "Licenses, Certifications, and Approvals for Nuclear Power Plants,"

U.S. Nuclear Regulatory Commission, Washington, DC.

44.SECY-00-0198, "Status Report on Study of Risk-Informed Changes to the Technical Requirementsof 10 CFR Part 50 (Option 3) and Recommendations on Risk-Informed Changes to 10 CFR 50.44 (Combustible Gas Control)," U.S. Nuclear Regulatory Commission, Washington, DC,

September 14, 2000.

55.NUREG-1150, "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,"

U.S. Nuclear Regulatory Commission, Washington, DC, December 1990.

66.SECY-93-087, "Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs," U.S. Nuclear Regulatory Commission, Washington, DC, April 2, 1993.

5

7 All regulatory guides listed herein were published by the U.S. Nuclear Regulatory Commission or its predecessor, the U.S. Atomic Energy Commission. Most are available electronically through the Electronic Reading Room

on the NRC's public Web site, at http://www.nrc.gov/reading-rm/doc-collections/reg-guides/. Single copies of regulatory guides may also be obtained free of charge by writing the Reproduction and Distribution Services Section, ADM, USNRC, Washington, DC 20555-0001, by fax to (301) 415-2289, or by email to DISTRIBUTION@nrc.gov. Active guides may also be purchased from the National Technical Information Service (NTIS). Details may be

obtained by contacting NTIS at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov

,by telephone at (800) 553-NTIS (6847) or (703) 605-6000, or by fax to (703) 605-6900. Copies are also available for

inspection or copying for a fee from the NRC's Public Document Room (PDR), which is located at 11555 Rockville

Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-000

1. The PDR

can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email

to PDR@nrc.gov

.8 Copies are available electronically through the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html , under Accession #ML021270103.Rev. 3 of RG 1.7, Page 127.Regulatory Guide 1.155, "Station Blackout," U.S. Nuclear Regulatory Commission, Washington, DC.

78.NUREG-1462, "Final Safety Evaluation Report Related to the Certification of the System 80+

Design, Docket No.52-002" U.S. Nuclear Regulatory Commission, Washington, DC,

August 1994.

69.NUREG-1503, "Final Safety Evaluation Report Related to the Certification of the Advanced Boiling-Water Reactor Design, Docket No.52-001," U.S. Nuclear Regulatory Commission, Washington, DC, July 1994.

610.NUREG-1512, "Final Safety Evaluation Report Related to the Certification of the AP600Standard Design, Docket No.52-003," U.S. Nuclear Regulatory Commission, Washington, DC, September 1998.

611.NUREG-1793, "Final Safety Evaluation Report Related to the Certification of the AP1000Standard Design, Docket No.52-006," U.S. Nuclear Regulatory Commission, Washington, DC, September 2004.

612.Regulatory Guide 1.118, "Periodic Testing of Electric Power and Protection Systems,"

U.S. Nuclear Regulatory Commission, Washington, DC.

713.SECY-00-0191, "High-Level Guidelines for Performance-Based Activities,"

U.S. Nuclear Regulatory Commission, Washington, DC, September 1, 2000.

514.NUREG-0737, "Clarification of TMI Action Plan Requirements," U.S. Nuclear Regulatory Commission, Washington, DC, November 1980.

615."Confirmatory Order Modifying Post-TMI Requirements Pertaining to Containment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2," U.S. Nuclear Regulatory Commission, Washington, DC, September 28, 1998.

816.Regulatory Guide 1.28, "Quality Assurance Program Requirements (Design and Construction),"

U.S. Nuclear Regulatory Commission, Washington DC.

717.Regulatory Guide 1.30, "Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment," U.S. Nuclear Regulatory Commission, Washington, DC.

7

9 Copies may be purchased from the American Society of Mechanical Engineers, Three Park Avenue, NewYork, NY

10016-5990; phone (212) 591-8500; fax (212) 591-8501;

www.asme.org

.10 This regulatory analysis is available electronically under Accession #ML031640482 in the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html , and through the Public Electronic Reading Room on the NRC's public Web site, at http://www.nrc.gov/reading-rm/

doc-collections/commission/secys/2003/secy2003-0127/2003-0127scy.pdf#pagemode=bookmarks. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room (PDR), which is located at

11555 Rockville Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC 20555-0001.

The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax at (301) 415-3548, and by email to PDR@nrc.gov

.11 Draft Regulatory Guide DG-1117 is available electronically under Accession #ML022210067 in the NRC's Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html. Copies are also available for inspection or copying for a fee from the NRC's Public Document Room (PDR), which is

located at 11555 Rockville Pike, Rockville, Maryland; the PDR's mailing address is USNRC PDR, Washington, DC

20555-0001. The PDR can also be reached by telephone at (301) 415-4737 or (800) 397-4209, by fax

at (301) 415-3548, and by email to PDR@nrc.gov

.Rev. 3 of RG 1.7, Page 1318.Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation),"

U.S. Nuclear Regulatory Commission, Washington, DC.

719.Regulatory Guide 1.176, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Graded Quality Assurance," U.S. Nuclear Regulatory Commission, Washington, DC.

720.NUREG/CR-4905, "Detonability of H-Air-Diluent Mixtures," prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC, June 1987.

621.NUREG/CR-4961, "A Summary of Hydrogen-Air Detonation Experiments," prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC,

June 1987.

622.NUREG/CR-5275, "Flame Facility" (The Effect of Obstacles and Transverse Venting on Flame Acceleration and Transition to Detonation of Hydrogen-Air Mixtures at Large Scale), prepared

by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC,

April 1989.

623.NUREG/CR-5525, "Hydrogen-Air-Diluent Detonation Study of Nuclear Reactor Safety Analyses," prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory

Commission, Washington, DC, December 1990.

624.ASME Boiler and Pressure Vessel Code,Section III, "Rules for Construction of Nuclear Power Plant Components," American Society of Mechanical Engineers, New York, NY, 1992.

925."Final Regulatory Analysis for 50.44," Attachment 4 to SECY-03-0127,U.S. Nuclear Regulatory Commission, Washington, DC, July 2003.

1026.Draft Regulatory Guide DG-1117, "Control of Combustible Gas Concentrations in Containment."

U.S. Nuclear Regulatory Commission, Washington, DC, August 2002.

11


Retrieved from "https://kanterella.com/w/index.php?title=Regulatory_Guide_1.7&oldid=1136561"