[Code of Federal Regulations]
[Title 10, Volume 1]
[Revised as of January 1, 2001]
From the U.S. Government Printing Office via GPO Access
[CITE: 10CFR50.44]

[Page 704-708]
 
                            TITLE 10--ENERGY
 
                CHAPTER I--NUCLEAR REGULATORY COMMISSION
 
PART 50--DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES--Table of Contents
 
Sec. 50.44  Standards for combustible gas control system in light-water-cooled power reactors.

    (a) Each boiling or pressurized light-water nuclear power reactor 
fueled with oxide pellets withincylindrical zircaloy or ZIRLO cladding, 
must, as provided in paragraphs (b) through (d) of this section, include 
means for control of hydrogen gas that may be generated, following a 
postulated loss-of-coolant accident (LOCA) by--
    (1) Metal-water reaction involving the fuel cladding and the reactor 
coolant,
    (2) Radiolytic decomposition of the reactor coolant, and
    (3) Corrosion of metals.

This section does not apply to a nuclear power reactor facility for 
which the certifications required under Sec. 50.82(a)(1) have been 
submitted.
    (b) Each boiling or pressurized light-water nuclear power reactor 
fueled with oxide pellets within cylindrical zircaloy or ZIRLO cladding 
must be provided with the capability for--
    (1) Measuring the hydrogen concentration in the containment,
    (2) Insuring a mixed atmosphere in the containment, and
    (3) Controlling combustible gas concentrations in the containment 
following a postulated LOCA.
    (c)(1) For each boiling or pressurized light-water nuclear power 
reactor fueled with oxide pellets within cylindrical zircaloy or ZIRLO 
cladding, it must be shown that during the time period following a 
postulated LOCA, but prior to effective operation of the combustible gas 
control system, either:
    (i) An uncontrolled hydrogen-oxygen recombination would not take 
place in the containment; or
    (ii) The plant could withstand the consequences of uncontrolled 
hydrogen-oxygen recombination without loss of safety function.
    (2) If the conditions set out in paragraph (c)(1) of this section 
cannot be shown, the containment shall be provided with an inerted or an 
oxygen deficient atmosphere in order to provide protection against 
hydrogen burning and explosions during the time period specified in 
paragraph (c)(1) of this section.
    (3) Notwithstanding paragraphs (c)(1) and (c)(2) of this section:
    (i) Effective May 4, 1982 or 6 months after initial criticality, 
whichever is later, an inerted atmosphere shall be provided for each 
boiling light-water nuclear power reactor with a Mark I or Mark II type 
containment; and
    (ii) By the end of the first scheduled outage beginning after July 
5, 1982 and of sufficient duration to permit required modifications, 
each light-water nuclear power reactor that relies upon a purge/
repressurization system as the primary means for controlling combustible 
gases following a LOCA shall be provided with either an internal 
recombiner or the capability to install an external recombiner following 
the start of an accident. The internal or external recombiners must meet 
the combustible gas control requirements in paragraph (d) of this 
section. The containment penetrations used for external recombiners must 
either be:
    (A) Dedicated to that service only, conform to the requirements of 
Criteria 54 and 56 of appendix A of this part, be designed against 
postulated single failures for containment isolation purposes, and be 
sized to satisfy the flow requirements of the external recombiners, or
    (B) Of a combined design for use by either external recombiners or 
purge/repressurization systems and other systems, conform to the 
requirements of criteria 54 and 56 of appendix A of this part, be 
designed against postulated single failures both for containment 
isolation purposes and for operation of the external recombiners or 
purge/repressurization systems, and be sized to

[[Page 705]]

satisfy the flow requirements of the external recombiners or purge 
repressurization systems.
    (iii) To provide improved operational capability to maintain 
adequate core cooling following an accident, by the end of the first 
scheduled outage beginning after July 1, 1982 and of sufficient duration 
to permit required modifications, each light-water nuclear power reactor 
shall be provided with high point vents for the reactor coolant system, 
for the reactor vessel head, and for other systems required to maintain 
adequate core cooling if the accumulation of noncondensible gases would 
cause the loss of function of these systems. (High point vents are not 
required, however, for the tubes in U-tube steam generators.) The high 
point vents must be remotely operated from the control room. Since these 
vents form a part of the reactor coolant pressure boundary, the design 
of the vents and associated controls, instruments and power sources must 
conform to the requirements of appendix A and appendix B of this part. 
In particular, the vent system shall be designed to ensure a low 
probability that (A) the vents will not perform their safety functions 
and (B) there would be inadvertent or irreversible actuation of a vent. 
Furthermore, the use of these vents during and following an accident 
must not aggravate the challenge to the containment or the course of the 
accident.
    (iv)(A) Each licensee with a boiling light-water nuclear power 
reactor with a Mark III type of containment and each licensee with a 
pressurized light-water nuclear power reactor with an ice condenser type 
of containment issued a construction permit before March 28, 1979, shall 
provide its nuclear power reactor with a hydrogen control system 
justified by a suitable program of experiment and analysis. The hydrogen 
control system must be capable of handling without loss of containment 
structural integrity an amount of hydrogen equivalent to that generated 
from a metal-water reaction involving 75% of the fuel cladding 
surrounding the active fuel region (excluding the cladding surrounding 
the plenum volume).
    (B) Containment structural integrity must 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. This method could include the use of actual material 
properties with suitable margins to account for uncertainties in 
modeling, in material properties, in construction tolerances, and so on. 
Another method could include a showing that the following specific 
criteria of the ASME Boiler and Pressure Vessel Code are met:
    (1) That steel containments meet the requirements of the ASME Boiler 
and Pressure Vessel Code (Edition and Addenda as incorporated by 
reference in Sec. 50.55a(b)(1) of this part), specifically in Section 
III, Division 1, Subsubarticle NE-3220, Service Level C Limits, 
considering pressure and dead load alone (evaluation of instability is 
not required); and
    (2) That 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.
    (C) Subsubarticle NE-3220, Division 1, and Subsubarticle CC-3720, 
Division 2, of Section III of the ASME Boiler and Pressure Vessel Code, 
referenced in paragraphs (c)(3)(iv)(B)(1) and (c)(3)(iv)(B)(2) of this 
section, have been approved for incorporation by reference by the 
Director of the Office of the Federal Register. A notice of any changes 
made to the material incorporated by reference will be published in the 
Federal Register. Copies of the ASME Boiler and Pressure Vessel Code may 
be purchased from the American Society of Mechanical Engineers, United 
Engineering Center, 345 East 47th Street, New York, NY 10017. It is also 
available for inspection at the NRC Technical Reference Library, Two 
White Flint North, Room 2B9, 11545 Rockville Pike, Rockville, MD.
    (D) If the hydrogen control system relies on post-accident inerting, 
the containment structure must be capable of withstanding the increased 
pressure:
    (1) During the accident, where it is acceptable to show that it does 
not exceed Service Level C Limits or the

[[Page 706]]

Factored load Category (as described in paragraph (c)(3)(iv)(B) of this 
section; and
    (2) Following inadvertent full inerting during normal plant 
operations, where it is acceptable to show that it does not exceed 
either the Service Level A Limits of Subsubarticle NE-3220 (for a steel 
containment) or the Service Load Category of Subsubarticle CC-3720 (for 
a concrete containment.)
    (3) Modest deviations from the criteria in paragraph (c)(3)(iv)(D) 
of this section will be considered by the Commission if good cause is 
shown.
    (E) If the hydrogen control system relies on post-accident inerting, 
the systems and components required to establish and maintain safe 
shutdown and containment integrity must be designed and qualified for 
the environment caused by such inerting. Furthermore, inadvertent full 
inerting during normal plant operations must not adversely affect 
systems and components needed for safe operation of the plant.
    (v)(A) Each licensee with a boiling light-water nuclear power 
reactor with a Mark III type of containment and each licensee with a 
pressurized light-water nuclear power reactor with an ice condenser type 
of containment issued a construction permit before March 28, 1979, for a 
reactor that does not rely upon an inerted atmosphere to control 
hydrogen inside the containment, shall provide its nuclear power reactor 
with systems and components necessary to establish and maintain safe 
shutdown and to maintain containment integrity. These systems and 
components must be capable of performing their functions during and 
after exposure to the environmental conditions created by the burning of 
hydrogen. Environmental conditions caused by local detonations of 
hydrogen must also be included, unless such detonations can be shown 
unlikely to occur.
    (B) The amount of hydrogen to be considered is equivalent to that 
generated from a metal-water reaction involving 75% of the fuel cladding 
surrounding the active fuel region (excluding the cladding surrounding 
the plenum volume).
    (vi)(A) Each applicant for or holder of an operating license for a 
boiling light-water nuclear power reactor with a Mark III type of 
containment or for a pressurized light-water nuclear power reactor with 
an ice condenser type of containment issued a construction permit before 
March 28, 1979, shall submit an analysis to the Commission as specified 
in Sec. 50.4.
    (B) The analysis required by paragraph (c)(3)(vi)(A) of this section 
must:
    (1) Provide an evaluation of the consequences of large amounts of 
hydrogen generated after the start of an accident (hydrogen resulting 
from the metal-water reaction of up to and including 75% of the fuel 
cladding surrounding the active fuel region, excluding the cladding 
surrounding the plenum volume) and include consideration of hydrogen 
control measures as appropriate;
    (2) Include the period of recovery from the degraded condition;
    (3) Use accident scenarios that are accepted by the NRC staff. These 
scenarios must be accompanied by sufficient supporting justification to 
show that they describe the behavior of the reactor system during and 
following an accident resulting in a degraded core.
    (4) Support the design of the hydrogen control system selected under 
paragraph (c)(3)(iv) of this section; and,
    (5) Show that for those reactors described in paragraph (c)(3)(iv) 
of this section that do not rely upon an inerted atmosphere to control 
hydrogen inside the containment:
    (i) The containment structural integrity as described in paragraph 
(c)(3)(iv) of this section will be maintained; and
    (ii) Systems and components necessary to establish and maintain safe 
shutdown and to maintain containment integrity will be capable of 
performing their functions during and after exposure to the 
environmental conditions created by the burning of hydrogen, including 
the effect of local detonations, unless such detonations can be shown 
unlikely to occur.
    (vii)(A) By June 25, 1985, each applicant for or holder of an 
operating license subject to the requirements of paragraphs (c)(3) (iv), 
(v) and (vi) of this section shall develop and submit to the Commission 
a proposed schedule

[[Page 707]]

for meeting these requirements. The schedule may be developed using 
integrated scheduling systems previously approved for the facility by 
the NRC.
    (B) For each applicant for an operating license as of Febuary 25, 
1985, the schedule shall provide for compliance with the requirements of 
paragraph (c)(3)(iv)(A) of this section prior to operation of the 
reactor in excess of 5 percent power. Completed final analyses are not 
necessary for a staff determination that a plant is safe to operate at 
full power provided that prior to such operation an applicant has 
provided a preliminary analysis which the staff has determined provides 
a satisfactory basis for a decision to support interim operation at full 
power until the final analysis has been completed. However, the record 
in this rulemaking shows that such preliminary analyses are not 
necessary for a staff determination that a plant is safe to operate at 
full power if the staff has determined for similar plants, referenced in 
this notice of rulemaking, that similar systems provide a satisfactory 
basis for a decision to support operation at full power until the 
preliminary analyses have been completed.
    (C) For those holders of operating licenses containing license 
conditions on Hydrogen Control Measures covered by this section, the 
schedule shall be consistent with those license conditions, or approved 
amendments thereto.
    (D) For those facilities not having an NRC approved integrated 
scheduling system, a final schedule for meeting the requirements of 
paragraphs (c)(3) (iv), (v) and (vi) of this section shall be 
established by the NRC staff within 90 days of receipt of a proposed 
schedule from the licensee or applicant, taking into account the current 
status of efforts at the facility to comply with the requirements; 
analyses that may be provided by applicants or licensees regarding the 
impacts of these requirements on other scheduled plant modifications, 
including any NRC-mandated safety modifications, and their relative 
importance to safety; and the Commission's objective that these 
requirements be complied with without undue delay.
    (d)(1) For facilities that are in compliance with Sec. 50.46(b), the 
amount of hydrogen contributed by core metal-water reaction (percentage 
of fuel cladding that reacts with water), as a result of degradation, 
but not total failure, of emergency core cooling functioning shall be 
assumed either to be five times the total amount of hydrogen calculated 
in demonstrating compliance with Sec. 50.46(b)(3), or to be 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 the plenum volume) to a depth of 0.00023 inch (0.0058 mm), 
whichever amount is greater. A time period of 2 minutes shall be used as 
the interval after the postulated LOCA over which the metal-water 
reaction occurs.
    (2) For facilities as to which no evaluation of compliance in 
accordance with Sec. 50.46(b) has been submitted and evaluated, the 
amounts of hydrogen so contributed shall be assumed to be that amount 
resulting from the reaction of 5 percent of the mass of metal in the 
cladding cylinders surrounding the fuel, excluding the cladding 
surrounding the plenum volume.
    (e) For facilities whose notice of hearing on the application for a 
construction permit was published on or after November 5, 1970, purging 
and/or repressurization shall not be the primary means for controlling 
combustible gases following a LOCA. However, the capability for 
controlled purging shall be provided. For these facilities, the primary 
means for controlling combustible gases following a LOCA shall consist 
of a combustible gas control system, such as recombiners, that does not 
result in a significant release from containment.
    (f) For facilities with respect to which the notice of hearing on 
the application for a construction permit was published between December 
22, 1968, and November 5, 1970, if the incremental radiation dose from 
purging (and repressurization if a repressurization system is provided) 
occurring at all points beyond the exclusion area boundary after a 
postulated LOCA calculated in accordance with Sec. 100.11(a)(2) of this 
chapter is less than 2.5 rem to the whole body and less than 30 rem to

[[Page 708]]

the thyroid, and if the combined radiation dose at the low population 
zone outer boundary from purging and the postulated LOCA calculated in 
accordance with Sec. 100.11(a)(2) of this chapter is less than 25 rem to 
the whole body and less than 300 rem to the thyroid, only a purging 
system is necessary, provided that the purging system and any filtration 
system associated with it are designed to conform with the general 
requirements of Criteria 41, 42, and 43 of appendix A to this part. 
Otherwise the facility shall be provided with another type of 
combustible gas control system (a repressurization system is acceptable) 
designed to conform with the general requirements of Criteria 41, 42, 
and 43 of appendix A to this part. If a purge system is used as part of 
the repressurization system, the purge system shall be designed to 
conform with the general requirements of Criteria 41, 42, and 43 of 
appendix A to this part. The containment shall not be repressurized 
beyond 50 percent of the containment design pressure.
    (g) For facilities with respect to which the notice of hearing on 
the application for a construction permit was published on or before 
December 22, 1968, if the combined radiation dose at the low population 
zone outer boundary from purging (and repressurization if a 
repressurization system is provided) and the postulated LOCA calculated 
in accordance with Sec. 100.11(a)(2) of this chapter is less than 25 rem 
to the whole body and less than 300 rem to the thyroid, only a purging 
system is necessary, provided that the purging system and any filtration 
system associated with it are designed to conform with the general 
requirements of Criteria 41, 42, and 43 of appendix A to this part. 
Otherwise, the facility shall be provided with another type of 
combustible gas control system (a repressurization system is acceptable) 
designed to conform with the general requirements of Criteria 41, 42, 
and 43 of appendix A to this part. If a purge system is used as part of 
the repressurization system, it shall be designed to conform with the 
general requirements of Criteria 41, 42, and 43 of appendix A to this 
part. The containment shall not be repressurized beyond 50 percent of 
the containment design pressure.
    (h) As used in this section: (1) Degradation, but not total failure, 
of emergency core cooling functioning means that the performance of the 
emergency core cooling system is postulated, for purposes of design of 
the combustible gas control system, not to meet the acceptance criteria 
in Sec. 50.46 and that there could be localized clad melting and metal-
water reaction to the extent postulated in paragraph (d) of this 
section. The degree of performance degradation is not postulated to be 
sufficient to cause core meltdown.
    (2) A combustible gas control system is a system that operates after 
a LOCA to maintain the concentrations of combustible gases within the 
containment, such as hydrogen, below flammability limits. Combustible 
gas control systems are of two types: (i) Systems that allow controlled 
release from containment, through filters if necessary, such as purging 
systems and repressurization systems, and (ii) systems that do not 
result in a significant release from containment such as recombiners.
    (3) A purging system is a system for the controlled release of the 
containment atmosphere to the environment through filters if needed.
    (4) A repressurization system is a system used to dilute the 
concentration of combustible gas within containment by adding inert gas 
or air to the containment. Dilution of the combustible gas results in a 
delay in time until a flammable concentration is reached and permits 
fission product decay. Operation is limited to a containment 
repressurization to 50 percent of the containment design pressure. A 
purging system is normally part of the repressurization system.

[43 FR 50163, Oct. 27, 1978, as amended at 46 FR 58486, Dec. 2, 1981; 50 
FR 3504, Jan. 25, 1985; 50 FR 5567, Feb. 11, 1985; 51 FR 40308, Nov. 6, 
1986; 53 FR 43420, Oct. 27, 1988; 57 FR 39358, Aug. 31, 1992, 61 FR 
39299, July 29, 1996; 64 FR 48951, Sept. 9, 1999]