Regulatory Guide 1.7: Difference between revisions

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{{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, 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 REGULATORY GUIDE
{{#Wiki_filter:U.S. NUCLEAR REGULATORY COMMISSION                                                                   March 2007 Revision 3 REGULATORY GUIDE
U.S. NUCLEAR. REGULATORY COMMISSION                                                                                                         Revision 1 September 1976 OFFICE OF STANDARDS DEVELOPMENT
                                    OFFICE OF NUCLEAR REGULATORY RESEARCH
                                                          REGULATORY GUIDE 1.7 CONTROL or COMIBUSTIBLE GAS CONCENTRATIO6NS IN
                                                REGULATORY GUIDE 1.7 (Draft was issued as DG-1117, dated August 2002)
                            CONTAMN14ENT FOLLOWING A LOSS-OF-COOLAINT ACCIDENT*
        CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS
                                                        IN CONTAINMENT


==A. INTRODUCTION==
==A. INTRODUCTION==
Criterion 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 cooling. Criterion 50, "Contain-                                                                                            'i 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...
In September 2003, the U.S. Nuclear Regulatory Commission (NRC) issued a revision of Section 50.44, Combustible Gas Control for Nuclear Power Reactors (Ref. 1), which amended Title 10,
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 to*.                                                                control the concen- trations of such substances following postulated accidents and ensure that containment intuegrity is maintained.
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 required under 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 Design Certifications; 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.


In addition, the Commission has published progposed amendments to Part 50
The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff need in reviewing applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.
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 basis.


Bi      DISCUSSION
This guide was issued after consideration of comments received from the public. The NRC staff encourages and welcomes comments and suggestions in connection with improvements to published regulatory guides, as well as items for inclusion in regulatory guides that are currently being developed.
            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 10/27/71.


USNRC REGULATORY GUIDES                                          Co.... is 'heMot be sent to           SetiJavy 0 ite Conission. U.S. Nuclei.
The NRC staff will revise existing guides, as appropriate, to accommodate comments and to reflect new information or experience. Written comments may be submitted to the Rules and Directives Branch, Office of Administration, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001.


Nlguvistory Gulde ate istst d to describe eind makhe availabte to the Public          Rlegulaloiro Commissiton*    Wai0"ingcn. D C 2O*bS. Attention Oc.clitiong end met hodaI accopti ble to the NFIC stalf of Implementing ssociftic part of the          Seritice Secti, Commii~o*s'      *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. or *o providi guidanc, to ippli cents. Regulatory Guides em. not substitiute& lo, regulations. and compliance          $. 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 T'Inspom*sron the g uIdes wi-l be ifecepteble                    bi sIt o atdthe indings
Regulatory guides are issued in 10 broad divisions: 1, Power Reactors; 2, Research and Test Reactors; 3, Fuels and Materials Facilities;
                                                                      *    etquils to    3 fuel, and Materitsals, acdtiev          Occupational He*ltt thitissualceao cl      *ofUfricele;0tteit ofmlicont: bythie Commision                  4A f riOnninentotendSiling              9 Antitrust Pat*,#.
4, Environmental and Siting; 5, Materials and Plant Protection; 6, Products; 7, Transportation; 8, Occupational Health; 9, Antitrust and Financial Review;
Cohmirncnis and eugg*ltiOnls fo improvements in these guides ite encowteged            S Materatls and Pleant Protection      tO Genetal
and 10, General.
*I alt times. end guides will he revised. mi appropriati. to ¢ccommodalt r.om mints and to rltIct new inlotmnetin or e.ptorince. However. ccrnmniett on                Cope* of pubtished guides mey be obtained bn -. Iltlen *nquell idcatimg the tht  tuide. t receivied within about two momthsl ofter its 4suance witt h p.            divisions d*sired to the U S Nuclear tegualtoty Comn -ssonWashington 0 C.


liculaill useful'tn evaluating    it        I.tt.
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.


t*o*oin    eIy--on                              2MM. Attention Ottecto'. Otlice ot Standards Oeiaoopmnit
Electronic copies of this guide and other recently issued guides are available through the NRCs public Web site under the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070290080.
                                                                                                                                                                        "A
                                                                                      ,                          ,.,,.,.         4,.


..............
==B. DISCUSSION==
2.     Radiolytic decomposition of the postaccident emergency cooling solutions (oxygen will also evolve in this process),
Section 50.44 provides requirements for the mitigation of combustible gas generated by a beyond-design-basis accident. In existing light-water reactors, the principal combustible gas is hydrogen.
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 conditions could also occur.
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.


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).
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
        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
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.
                            *4
    36 FR 12248 and 36 FR 24082.


A copy of the docket file may be examined in     the NRC public dociment room.
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 associated 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.


1.7-2                                        4
(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.


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 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.


The staff fuels that this "overlap" In protection requirements provides ain appropriate ind prudent safety margin against unpredicted events during the course of accidents.
Rev. 3 of RG 1.7, Page 2


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.    To obtain this margin, the assumed amount of hydrogen should be no less than five times that calculated in accordance with 5 50.46.
(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 paragraph apply only to water-cooled reactor designs with characteristics (e.g., type and quantity of cladding materials) such that the potential for production of combustible gases is comparable to light-water reactor designs licensed as of October 16, 2003.


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
(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:
          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.      This is because the hydrogen generation from metal-water reaction is a surface phenomenon.    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.    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.
        (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.


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
The combustible gas control systems, the atmosphere mixing systems, and the provisions for measuring and sampling that are required by Section 50.44 are risk-significant, as they have the ability to 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.
                                                1.7-3
... .:...                    .    .  .


.....
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.
                                                .  .. ...  ... .
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)
Rev. 3 of RG 1.7, Page 3
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.
==C. REGULATORY POSITION==
1.      Combustible Gas Control Systems The following design guidance is applicable to combustible gas control systems installed to 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 not be 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 50.44. 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 Monitors
2.1      Hydrogen 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:
1 Section 50.44 does not require the deliberate ignition systems used by BWRs with Mark III type containments and 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 staffs review of Generic Safety Issue 189, Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen Combustion During a Severe Accident.
 
Rev. 3 of RG 1.7, Page 4
 
(1) Equipment Survivability: The hydrogen monitoring equipment need not be qualified in 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 environment for 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 5
 
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 be from 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 function. 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 requirements can 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 to Containment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2 (Ref. 15), the staff developed a 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.


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
Rev. 3 of RG 1.7, Page 6
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.
2.2      Oxygen 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 qualified in 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 metal-water reaction is of significance when establishing system per- formance requirements for containment designs that employ time-dependent hydrogen control features.    The staff recognizes that it would be unrealistic to assume a' instantaneous release of hydrogen from an assumed metal-water reaction.    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. 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
The evaluation of survivability should consider the effects of the post-accident environment for 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.
                                    1.7-4


7.        .
Acceptable approaches for demonstrating equipment survivability are described in Chapter 19 of the ABWR FSER (Ref. 9) and the AP1000 FSER (Ref. 11).
"lower volume" of ice condenser containments, and the full volume of "dry"
(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.
containments.        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.


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.          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.
(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.


Table I defines conser, tive values and assumptions that may be used to Cv.Iluate the production of cu:.-,)ustible gases following a LOCA.
(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 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 flam=.*ble 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.
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.


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


The Bureau of Mines has conducted experiments at its        facilities with initial      hydrogen volume concentrations on the order of 4 to 12 volume percent.
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.


On the basis of these experiments and a review of other reports, the NRC
Rev. 3 of RG 1.7, Page 7
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.         For initial  concentrations of hydrogen greater than about 6 volume percent, it is possible in the presence of sufficient ignition
                                            1.7-5


*  -..          .     *~..i'*
(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.
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.
(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.


In small containments, the rduount of metal-water reaction postulated in Table 1 may result in hydrogen concentrations above acceptable limits.
(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.


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) atmospheres.      This measure, the
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.
"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.      Hydrogen recombiners can process the containment atmosphere at a rate of only 100 scfm per recombiner.


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


There are presently no other methods of combustible gas control except for purge systems, which release radioactive materials.
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.


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.
(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.


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.
(9)    Human Factors: The instrumentation should be designed to facilitate the recognition, location, replacement, repair, or adjustment of malfunctioning components or modules.


Care should be taken in its design to ensure that the system itself      does not introduce safety problems that may affect containment integrity; for example, if a flame recombiner is used, propagation of flame into the containment should be prevented.
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.


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.      An example of an acceptable arrangement would be to provide at least one hydrogen control
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.
                                    1.7-6


system per site with the provision that a redundant unit would be available from a nearby site.
(10)    Direct Measurement: To the extent practicable, monitoring instrumentation inputs should be from 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.


==C. REGULATORY POSITION==
3.      Atmosphere Mixing Systems Section 50.44 requires that all containments have a capability for ensuring a mixed atmosphere.
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.
 
This capability may be provided by an active, passive, or combination system. 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.
 
Rev. 3 of RG 1.7, Page 8
 
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 gases that 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 Production Materials 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.
 
2 The 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 analyses through sensitivity studies, validation with data, or other means.
 
Rev. 3 of RG 1.7, Page 9
 
==D. IMPLEMENTATION==
The purpose of this section is to provide information to applicants and licensees regarding the NRC staffs 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 NRCs 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.


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


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 combustible gas control system should be designated Seismic Category I
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.
(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.
Rev. 3 of RG 1.7, Page 10


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.                 These values may be changed on the basis of additional experimental evidence and analyses.
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, 2003.3
2. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission, Washington, DC.4
3. 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.4
4. 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), U.S. Nuclear Regulatory Commission, Washington, DC,
  September 14, 2000.5
5. NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, December 1990.6
6. 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
3 All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are available electronically 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 NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs 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.   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.
4 All NRC regulations listed herein are available electronically through the Public Electronic Reading Room on the NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD;
  the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;
  fax (301) 415-3548; email PDR@nrc.gov.


1.7-7
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 NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs mailing address is USNRC PDR, Washington, DC 20555; telephone (301) 415-4737 or (800) 397-4209;
                                                ... ..............................
  fax (301) 415-3548; email PDR@nrc.gov.
                                                            ..............................


- pV~T'~~Nfl. .. .. - .,.
6 Copies are available at current rates from the U.S. Government Printing Office, P.O. Box 37082, Washington, DC
                                  D. IMPLEXENTATION
  20402-9328 (telephone 202-512-1800); or from the National Technical Information Service (NTIS) by writing NTIS
        This guide will be used by the staff in the evaluation of l.icensut.s'
  at 5285 Port Royal Road, Springfield, Virginia 22161, online at http://www.ntis.gov, by telephone at (800) 553-NTIS
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.
  (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 NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs 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 NRCs Public Web site http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/.
                                            Rev. 3 of RG 1.7, Page 11


1.7-8
7.  Regulatory Guide 1.155, Station Blackout, U.S. Nuclear Regulatory Commission, Washington, DC.7
8.  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 1994.6
9.  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 1994.6
10. NUREG-1512, Final Safety Evaluation Report Related to the Certification of the AP600
    Standard Design, Docket No. 52-003, U.S. Nuclear Regulatory Commission, Washington, DC,
    September 1998.6
11. NUREG-1793, Final Safety Evaluation Report Related to the Certification of the AP1000
    Standard Design, Docket No. 52-006, U.S. Nuclear Regulatory Commission, Washington, DC,
    September 2004.6
12. Regulatory Guide 1.118, Periodic Testing of Electric Power and Protection Systems, U.S. Nuclear Regulatory Commission, Washington, DC.7
13. SECY-00-0191, High-Level Guidelines for Performance-Based Activities, U.S. Nuclear Regulatory Commission, Washington, DC, September 1, 2000.5
14. NUREG-0737, Clarification of TMI Action Plan Requirements, U.S. Nuclear Regulatory Commission, Washington, DC, November 1980.6
15. 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, 1998.8
16. Regulatory Guide 1.28, Quality Assurance Program Requirements (Design and Construction),
    U.S. Nuclear Regulatory Commission, Washington DC.7
17. 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
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 NRCs 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.


Table I.  Acceptable Assumptions for Evaluating the Production of Combustible Gases Following a Loss-of-Coolant Accident (LOCA)
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 NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs mailing address is USNRC PDR, Washington, DC 20555-000
            Parameter                                      Acceptable Value Fraction of fission product                (a)  Beta radiation energy absorbed by                   (1) Betas from fission products the coolant*                                          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
Hydrogen yield rate G(H2 )               0.5 molecule/l0OEv Oxygen yield rate G(O2 )                 0.25 molecule/100Ev Extent and evolution time of              Hydrogen production is 5 times the initial core metal-water reaction        extent of the maximum calculated hydrogen production from the              reaction under 10 CFR Part 50,
cladding surrounding the fuel            §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.


Aluminum corrosion rate for              200 mils/yr (This value should be aluminum exposed to alkaline              adjusted upward for higher solutions                                temperatures early in the accident sequence.)
===1. The PDR===
Fission product distribution model        50% of the halogens and 1% of the solids present in the core are intimately mixed with the coolant water.
    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.


1.7-9
8 Copies are available electronically through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession #ML021270103.
                      .                                             ,   *.       "-


Wý-.                                             "M":  .  .- -    I "POMIWPTMWý    .
Rev. 3 of RG 1.7, Page 12
                                        Table 1 (Continued)
                          Parameter                                            Acceptable Volue All noble gases are released to the containment.


All other fission products remain lit fucl rods.
18. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation),
    U.S. Nuclear Regulatory Commission, Washington, DC.7
19. Regulatory Guide 1.176, An Approach for Plant-Specific, Risk-Informed Decisionmaking:
    Graded Quality Assurance, U.S. Nuclear Regulatory Commission, Washington, DC.7
20. NUREG/CR-4905, Detonability of H-Air-Diluent Mixtures, prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC, June 1987.6
21. 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 1987.6
22. 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 1989.6
23. 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 1990.6
24. ASME Boiler and Pressure Vessel Code, Section III, Rules for Construction of Nuclear Power Plant Components, American Society of Mechanical Engineers, New York, NY, 1992.9
25. Final Regulatory Analysis for 50.44, Attachment 4 to SECY-03-0127,U.S. Nuclear Regulatory Commission, Washington, DC, July 2003.10
26. Draft Regulatory Guide DG-1117, Control of Combustible Gas Concentrations in Containment.


Hydrogen concentration limit              4 v/o***
U.S. Nuclear Regulatory Commission, Washington, DC, August 2002.11
                                                          P.AO-d~id It IMAIiUj Limit (I V/0 ttt- geii Is present. )
9 Copies may be purchased from the American Society of Mechanical Engineers, Three Park Avenue, NewYork, NY
                For water, borated water, and borated alkaline solutions; for other solutions, data should be presented.
    10016-5990; phone (212) 591-8500; fax (212) 591-8501; www.asme.org.


This fraction is thought to be conservative; further analysis may show that it should be revised.
10
    This regulatory analysis is available electronically under Accession #ML031640482 in the NRCs 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 NRCs 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 NRCs Public Document Room (PDR), which is located at
    11555 Rockville Pike, Rockville, Maryland; the PDRs mailing address is USNRC PDR, Washington, DC 20555-0001.


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.
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.


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). 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.
11 Draft Regulatory Guide DG-1117 is available electronically under Accession #ML022210067 in the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html.


Applicants and licensees should demonstrate such capability by suitable analyses and qualification test results.
Copies are also available for inspection or copying for a fee from the NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs 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.


1.7-10
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Revision as of 19:59, 13 March 2020

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


U.S. NUCLEAR REGULATORY COMMISSION March 2007 Revision 3 REGULATORY GUIDE

OFFICE OF NUCLEAR REGULATORY RESEARCH

REGULATORY GUIDE 1.7 (Draft was issued as DG-1117, dated August 2002)

CONTROL OF COMBUSTIBLE GAS CONCENTRATIONS

IN CONTAINMENT

A. INTRODUCTION

In September 2003, the U.S. Nuclear Regulatory Commission (NRC) issued a revision of Section 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 required under 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 Design Certifications; 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.

The U.S. Nuclear Regulatory Commission (NRC) issues regulatory guides to describe and make available to the public methods that the NRC staff considers acceptable for use in implementing specific parts of the agencys regulations, techniques that the staff uses in evaluating specific problems or postulated accidents, and data that the staff need in reviewing applications for permits and licenses. Regulatory guides are not substitutes for regulations, and compliance with them is not required. Methods and solutions that differ from those set forth in regulatory guides will be deemed acceptable if they provide a basis for the findings required for the issuance or continuance of a permit or license by the Commission.

This guide was issued after consideration of comments received from the public. The NRC staff encourages and welcomes comments and suggestions in 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 comments may 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 through the NRCs public Web site under the Regulatory Guides document collection of the NRCs Electronic Reading Room at http://www.nrc.gov/reading-rm/doc-collections/ and through the NRCs Agencywide Documents Access and Management System (ADAMS) at http://www.nrc.gov/reading-rm/adams.html, under Accession No. ML070290080.

B. DISCUSSION

Section 50.44 provides requirements for the mitigation of combustible gas generated by a 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 associated 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 2

(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 paragraph apply only to water-cooled reactor designs with characteristics (e.g., type and quantity of cladding materials) 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 provisions for measuring and sampling that are required by Section 50.44 are risk-significant, as they have the ability to 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.

Rev. 3 of RG 1.7, Page 3

C. REGULATORY POSITION

1. Combustible Gas Control Systems The following design guidance is applicable to combustible gas control systems installed to 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 not be 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 50.44. 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 Monitors

2.1 Hydrogen 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:

1 Section 50.44 does not require the deliberate ignition systems used by BWRs with Mark III type containments and 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 staffs review of Generic Safety Issue 189, Susceptibility of Ice Condenser and Mark III Containments to Early Failure from Hydrogen Combustion During a Severe Accident.

Rev. 3 of RG 1.7, Page 4

(1) Equipment Survivability: The hydrogen monitoring equipment need not be qualified in 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 environment for 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 5

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 be from 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 function. 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 requirements can 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 to Containment Hydrogen Monitors for Arkansas Nuclear One, Units 1 and 2 (Ref. 15), the staff developed a 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 6

2.2 Oxygen 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 qualified in 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 environment for 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 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment (Ref. 17)

(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 7

(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 be from 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 system. 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.

Rev. 3 of RG 1.7, Page 8

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 gases that 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 Production Materials 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.

2 The 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 analyses through sensitivity studies, validation with data, or other means.

Rev. 3 of RG 1.7, Page 9

D. IMPLEMENTATION

The purpose of this section is to provide information to applicants and licensees regarding the NRC staffs 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 NRCs 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 ANALYSIS

A separate regulatory analysis was not prepared for this guide. The regulatory analysis prepared for the revision of 10 CFR 50.44, Standards for Combustible Gas Control System in Light-Water-Cooled Power 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.

Rev. 3 of RG 1.7, Page 10

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, 2003.3

2. 10 CFR Part 50, Domestic Licensing of Production and Utilization Facilities, U.S. Nuclear Regulatory Commission, Washington, DC.4

3. 10 CFR Part 52, Licenses, Certifications, and Approvals for Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC.4

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

September 14, 2000.5

5. NUREG-1150, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, U.S. Nuclear Regulatory Commission, Washington, DC, December 1990.6

6. 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

3 All Federal Register notices listed herein were issued by the U.S. Nuclear Regulatory Commission, and are available electronically 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 NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs 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 NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD;

the PDRs 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 NRCs 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 NRCs Public Document Room at 11555 Rockville Pike, Rockville, MD; the PDRs 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 NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs 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 NRCs Public Web site http://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/.

Rev. 3 of RG 1.7, Page 11

7. Regulatory Guide 1.155, Station Blackout, U.S. Nuclear Regulatory Commission, Washington, DC.7

8. 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 1994.6

9. 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 1994.6

10. NUREG-1512, Final Safety Evaluation Report Related to the Certification of the AP600

Standard Design, Docket No.52-003, U.S. Nuclear Regulatory Commission, Washington, DC,

September 1998.6

11. NUREG-1793, Final Safety Evaluation Report Related to the Certification of the AP1000

Standard Design, Docket No.52-006, U.S. Nuclear Regulatory Commission, Washington, DC,

September 2004.6

12. Regulatory Guide 1.118, Periodic Testing of Electric Power and Protection Systems, U.S. Nuclear Regulatory Commission, Washington, DC.7

13. SECY-00-0191, High-Level Guidelines for Performance-Based Activities, U.S. Nuclear Regulatory Commission, Washington, DC, September 1, 2000.5

14. NUREG-0737, Clarification of TMI Action Plan Requirements, U.S. Nuclear Regulatory Commission, Washington, DC, November 1980.6

15. 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, 1998.8

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

U.S. Nuclear Regulatory Commission, Washington DC.7

17. 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

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 NRCs 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 NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs 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 NRCs 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 12

18. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation),

U.S. Nuclear Regulatory Commission, Washington, DC.7

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

Graded Quality Assurance, U.S. Nuclear Regulatory Commission, Washington, DC.7

20. NUREG/CR-4905, Detonability of H-Air-Diluent Mixtures, prepared by Sandia National Laboratory for the U.S. Nuclear Regulatory Commission, Washington, DC, June 1987.6

21. 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 1987.6

22. 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 1989.6

23. 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 1990.6

24. ASME Boiler and Pressure Vessel Code,Section III, Rules for Construction of Nuclear Power Plant Components, American Society of Mechanical Engineers, New York, NY, 1992.9

25. Final Regulatory Analysis for 50.44, Attachment 4 to SECY-03-0127,U.S. Nuclear Regulatory Commission, Washington, DC, July 2003.10

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

U.S. Nuclear Regulatory Commission, Washington, DC, August 2002.11

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 NRCs 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 NRCs 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 NRCs Public Document Room (PDR), which is located at

11555 Rockville Pike, Rockville, Maryland; the PDRs 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 NRCs 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 NRCs Public Document Room (PDR), which is located at 11555 Rockville Pike, Rockville, Maryland; the PDRs 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 13


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