ML20038A079

Status of Hydrogen Control Activities

ML20038A079
Person / Time
Issue date:	09/03/1981
From:	Butler W, Fleishman M, Larkins J Office of Nuclear Reactor Regulation, NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
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ML20038A080	List: ML20038A079 ML20040A601 ML20040A609 ML20040A617
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NUDOCS 8109280158
Download: ML20038A079 (27)

Text

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t STATUS OF HYDROGEN CONTROL ACTIVITIES September 3,1981 M. R. Fleishman, RES W. R. Butler, NRR

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M. A. Taylor, RES

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TABLE OF CONTENTS Pace 1.

Lessons Learned from TMI-2 Accident and Interim Rule on H2 Control..........................

1 1.1 Lessons Learned......................

1 1.1.1 Overview.......

1 1.1.2 Hydrogen Production and Release at THI-2......

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1. 2 Industry Meetings in 1979.................

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1. 3 TMI-2 Action Plan.....................

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1. 4 Long-Term Rulemaking on Degraded and Melted Cores (Severe Accident Rule).......................

3 1.5 Interim Rule Development..................

3 1.5.1 Proposed Rule....................

3 1.5.2 Proposed Requirements................

4 1.5.3 Commission Considerations..............

5 1.5.4 Current Status of Interim Rule...........

5 2.

Existing Regulatory Policy.........

6 2.1 Sections 50.44 and 50.46..................

6 2.2 Regul ato ry Gui de 1. 7....................

7 2.3 Standard Review Plan....................

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2. 4 Current Regulatory Posi tion................

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2. 5 Pre-TMI assumptions on H2 Generation............

8 3.

Risk-Based Considerations on the Value of Inerting.......

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i 3.1 History of Risk Considerations...............

8 3.2 Basis for Risk Perception.................

9 3.3 Differing Views of Risk Analyses..............

10 4.

Recent Case Policy.......................

11 4.1 OL Reviews.........................

12 4.1.1 Large Ory Containments...............

12 4.1.2 Ice Condensers and Mark IIIs............

12 4.1.3 Mark Is and Mark IIs................

13 4.2 NTCP Reviews.. "......................

13 4.2.1 Large Dry Containments...............

14 4.2.2 Ice Condensers...................

14 4.2.3 Mark IIIs......................

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TABLE OF CONTENTS (Continued)

Page 4.3 Operating Reactors.....................

15 4.3.1 Large Dry Containments...............

15 4.3.2 Ice Condensers and Mark IIIs............

15 4.3.3 Mark Is and Mark IIs................

15 5.

Research on Hydrogen......................

16 5.1 NRC's Research Program...................

16 5.2 Hydrogen Behavior and Control Program...........

17 5.2.1 Experimental Program................

17 5.2.2 Analytical Program.................

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5. 2. 2.1 Model Development..............

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5. 2. 2. 2 Detonation Calculations...........

18 5.2.2.3 MARCH Calculations..............

18 5.2.2.4 Hydrogen Transport Calculations.......

18 5.2.3 Hydrogen Compendium and Generic Operators Manual..

18 5.3 Hydrogen Combustion, Mitigation and Prevention Program...

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. 5. 3.1 Water Fog and Foams.................

19 5.3.2 Oxygen Oepletion and Post-Accident Inerting.....

19 5.3.3 Gas Turbine and High Capacity Recombiners......

19 5.3.4 Deliberate Flaring'from High Point Vents......

19 5.4 Combustible Gas in Containment Program...........

20 5.5 Core Melt Technology Program................

20 5.6 Equipment Survival Program.................

20 5.7 Safety Margins for Containment Program...........

21 5.8 Safety Margins for Category I Structures Program......

21 5.9 Effects of Hydrogen Explosions Program.

21 5.10 NRC Technical Assistance Contracts on Hydrogen Behavior..

21 5.11 00E, EPRI and Industry Programs on Hydrogen........

22 5.11.1 EPRI Research Program on Hydrogen.........

22 5.11.2 Owners Groups Sponsored Hydrogen Research.

24 5.11.3 IDCOR Hydrogen Program..............

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5. 11.4 00E Hydrogen Program 24 iii

1.

LESSONS LEARNED FROM THI-2 ACCIDENT AND INTERIM RULE ON He CONTROL 1.1 Lessons Learned 1.1.1 Overview The accident at Three Mile Island, Unit 2 (TMI-2), resulted in a severely damaged or degraded reactor core with a concomitant release of radioactive mate-rial to the primary coolant system and a large amount of fuel cladding metal-water (zirconium-oxygen) reaction in the core with hydrogen generation well in excess of the amounts required to be considered for design purposes by 10 CFR S 50.44, " Standards for. combustible gas control system in light water cooled power reactors." A review and evaluation of accident by the Lessons Learned Task Force (LLTF) revealed design and operational limitations that existed relative to mitigating the consequences of the accident and determining the status of the facility during and following the accident. The initial find-ings of the LLTF were published in NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations," dated July 1979, and NUREG-0585, "THI-2 Lessons Learned Task Force Final Report," dated October 1979.

1.1. 2 Hydrogen Production and' Release at TMI-2 A Special Inquiry Group (SIG), formed to evaluate the THI-2 accident, among other things developed an estimate of the hydrogen production and release during the accident. The SIG approached the problem of estimating the amount of hydro-gen produced and its release to the containment and to the atmosphere through internal calculations, outside contracts, and by reviewing the results of previous investigations. The SIG produced a best-estimate result and concluded that an upper limit could be defined though there remains a fairly large band of uncertainty.

The amount of hydrogen produced is a direct measure of tne amount of damage to the fuel assemolies (and vice versa) and one approacn is to begin by estimat-ing the dacage to the core and calculating the amount of hydrogen produced followed by subtracting away the known means of hycrogen removal and evaluating the difference.

Adjustments may then be made to effect a balance between pro-duction and removal.

Other approaches have been the reverse--determine how much hydrogen was removed (by the containment fire and the hydrogen bubole) 1

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and calculate backwards to determine the original amount. Neither approach is fully satisfactory because of the uncertainties in the amcunt of core damage, in the size and composition of the hydrogen bubble, and in the size of the fire in the containment.

It was estimated that the total amount of H2 generated was 990 pounds which was equivalent to that amount generated from the reaction of 45% of the fuel cladding with water. Of the 990 pounds of H2 that was generated, 770 pounds was estimated to be released to the containment building primarily during the 6th thru 10th hours; of this, 594 pounds was burned and 176 pounds remained in the containment building. Of the 220 pounds remaining in the primary system (during the time of the fire) 58 pounds was dissolved in the water and 162 bubble. At about the 16th hour, when the reactor coolant pounds was in the H2 pump was started, most of the H2 remaining in the primary system (70%) was released through the letdown line to the makeup tank and vented to the stack or leaked to the auxiliary building.

It is noted that there still exists questions concerning the amount of H, both generated and involved in the burn, and the source of the H.

H.F.

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Ring of Dupont has suggested that the amount of H2 generated could be less than the. generally accepted amount by as much as a factor of 10.

Furthermore, he suggests that the H2 was generated by radiolysis induced by the release of radioactive fission products.from failed fuel elements.

Finally, he suggested burn involved a much smaller amount of H2 than ' assumed previously, that the H2 and consisted of a series of small local burns rather than one containment wide burn.

This matter is still undergoing technical discussions but final resolu-tion will have to await a detailed examination of the TMI-2 core and the interior of the containment.

1. 2 Industrv Meetinos in 1979 The Office of Nuclear Reactor Regulatica sent letters to all licensees of operating nucle'ar power plants, operating license applicants, licensees of plants under construction and pending construction permit applicants, inform-ing them of the followup actions that should be taken in light of the lessons learned from TNI-2.

Specifically for all operating nuclear power plants, letters were sent on Septemoer 13, 1979 concerning " Followup Actions Resulting from the NRC Staff Reviews Regarding the Three Mile Island Unit 2 Accident,"

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l in which a set of recommendations was presented that was to be implemented.

Regional meetings were held during the week of September 24, 1979 to explain in more detail each of the recommendations. On October 30, 1979, letters were again sent to all licensees of operating nuclear power plants to provide addi-tional clarification of the NRC staff recommendations (Discussion of Lessons Learned Short Term Requirements).

1. 3 TMI-2 Action Plan The Commission reviewed and endorsed the "TMI-2 Action Plan," NUREG-0660, I

published in May 1980.

In the Action Plan, Task II.B., " Consideration of Degraded or Melted Cores in Safety Review," identified a number of actions that involve developing and implementing a phased program to include, in safety reviews, consideration of core degradation and melting beyond the design basis.

One of these actions, Task II.B.8, was a "Rulemaking proceeding on degraded core accidents." This was to comprise two separate actions, a long-term rule-making and an interim rule.

1.4 Long-Term Rulemaking on Georaded and Melted Cores l

The long-term rulemaking was td be preceded by an advance notice of pro-poseo rulemaking.

This long-term rulemaking was to consider to what extent, if any, nuclear power plants should be designed to deal effectively with degraded-core and core-melt accidents and to mitigate te ccnsequences thereof.

The advance notice was forwarded to the Commission for approval via SECY-80-357,

" Degraded Cooling Rulemaking." The advance notice was published on October 2, 1980.

This rulemaking has evolved into what is now being referred to as the Severe Accident Rule.

1. 5 Interim Rule Develoament 1.5.1 Procosed Rule The other action identified in Task II.B.8 involved an Interim Rule which was to be based on a number of recommendations specifically related to accidents that involve severely damaged or degraded reactor cores.

The recommendations were determined by the staff to be of such safety significance that they should 3

be codified by regulation in order to provide assurance that the public health and safety was adequately protected.

The staff believed that the changes result-ing from these requirements would improve the capability of nuclear power plants to deal with TMI-2 type accidents.

The proposed Interim Rule was forwarded to the Commission via SECY-80-399, was approved, and subsequently published on October 2,1980 along with the advance notice for the long-term rulemaking.

1.5.2 Prooosed Recuirements The requirements proposed in the Interim Rule involved hydrogen management in light-water reactors and specific design and other requirements to mitigate the consequences of degraded-core accidents. The staff position on hydrogen management was pr=sented to the Commission in SECY-80-107, 80-107A, and 80-107B,

" Proposed Interim Hydrogen Control Requirements for Small Containments." The proposed Interim Rule was consistent with these papers and also represented the rule:naking mentioned in Task II.B.7, " Analysis of hydrogen control" of the Action Plan. The other specific requirements were described previously in NUREG-0578, NUREG-0585, and in the September 13, 1979 and October 30, 1979 letters to licensees of operatir.g plants. The list of items covered by the proposed. Interim Rule was:

i 1.

Inerting of Mark I,and II 3WRs 2.

Design Analyses Control Penetrations 3.

Dedicated H2 Recombiner Capability 4.

H2 5.

High Point Vents 6.

Post-Accident Protection of Stfety Equipment and Areas -

7.

In-Plant Iodine Instrumentation 8.

Post-Accident Sampling 9.

Leakage Integrity Outside Containment 10.

Accioent Monitoring Instruments 11.

Detedtion of Inadequate Core Cooling 12.

Training to Mitigate Degraded Core Accidents '

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1.5.3 Commission Considerations The Commission has been considering the ability of all light-water reac-tors, particularly pressurized light-water reactor facilities with ice con-denser type containments and boiling light-water reactor facilities with Mark III type containments, to withstand an accident with the concomitant generation of large amounts of hydrogen, such as the type which occurred at Three Mile Island, Unit 2 (TMI-2). As a result, it was decided to add two new items to the list, namely H2 Control for Mark III BWRs and Ice. Condenser PWRs and Equipment Survivability During H2 Burning, and to change the scope of the required Design Analysis. On April 23, 1981, the staff briefed the Commission on a Final Interim Rule which included the above 12 items plus the 2 new items (14 items total).

During the Policy Session on April 30, 1981, the Commission reviewed SECY-81-246 and approved for publication in the Federal Register a notice of proposed rulemaking that would incorporate into 10 CFR Part 50 a set of TMI-2 requirements. for operating license applications (Distilled from NUREG-0737,

" Clarification of TMI Action Plan Requirements").

This proposed OL rule was published in the Federal Register on May 13, 1981, however, a similar rule with respect to operating reactors was disapproved by the Commission (SECY-81-422) on August 6, 1981.

Of the 14 items covered by the Interim Rule in the April 23, 1981 presenta-tion, 9 of the items were also included in the proposed OL rule (items 3, and I

5 to 12).

These items have been deleted from the Interim P.ule.

1.5.4 Current Status of Interim Rule The remaining items included in the Interim Rule are primarily related to H2 management and control aspects.

It has been broken up into a final rule covering 2 items which were originally proposed in October 1980 and a proposed rule covering 3 new items, not originally proposed, as follows:

Final Rule

- Inerting of Mark I and II BWRs

-H2 Recombiner Capability Control for Mark IIIs and Ice Condensers Proposed Rule

-H2

- Equipment Survivability

- Analyses l

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The Interim Rule has been sent to the Commission for its consideration via a paper designated as SECY-81-245A dated August 19, 1981.

2.

EXISTING REGULATORY POLICY 2.1 Sections 50.44 and 50.46 Section 50.44 " Standards foc combustible gas control system in light-water cooled power reactors," requires a licensee or license applicant to show that, during the time immediately following a postulated loss-of-coolant accident (LOCA) but before effective operation of the combustible gas control system, either:

(1) an uncontrolled hydrogen-oxygen recombination will not take place in the containment, or (2) the plant can withstand the consequences of uncon-trolled hydrogen-oxygen recombination without loss of safety.

If neither of these conditions can be shown, the containment must be provided with an inerted atmosphere or an oxygen deficient condition in order to provide protection against hydrogen burning and explosions during this time.

Secion 50.44 gives credit to performance of the emergency core cooling system (.ECCS) by specifying that the amount of hydrogen assumed to be contri-buted by the metal-water reaction shall be either five times the total amount of hydrogen calculated in demonstrating compliance with the ECCS acceptance criteria (S 50.46, " Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors") or an amount related to a specific depth of fuel clad reacted (approximately one percent of the fuel clad), whichever amount is greater: This generally results in a calculated fuel-cladding water reaction of less than 1 1/2 percent.

1. s a result, if a licensee or license applicant can show that the calculated metal-water reaction is well within the ECCS acceptance criteria (e.g., less than 0.5 percent of the fuel clad reacts which would be well within the 1.0 percent specified in the ECCS criteria), it is a straightforward task to demonstrate that there is no need to inert the containment of'its plant, even if the plant has a small containment volume such as in a Mark I or Mark II boiling water reactor (BWR).

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1 2.2 Raquiatory Guide 1.7 i

Most operating BWR plants with Mark I cbntainments have been inerted due to guidance provided in an early version of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Acci-t dent," in which the designs were required to accommodate five percent metal-water reaction during the LOCA blowdown. This guide was in effect prior to promulgation of 5 50.44. The combination of containment volume and zirconium inventory of the BWR Mark I containments made it necessary to inert these con-tainments to ensure that the hydrogen concentration in containment following a LOCA would not exceed the lower flammability limit of 4 volume percent in air.

2.3 Standard Review Plan (SRP)

The NRC staff review of H2 control is basically performed in accordance with the procedures set out in SRP section 6.2.5, " Combustible Gas Control in L

Containment." Section 6.2.5 provides guidance for implementing the require-ments of Sections 50.44 and 50.46, General Design Criteria 41, 42, 43 and 50 of Appen' ix A to 10 CFR Part 50, and Regulatory Guide 1.7.

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2.4 Current Regulatory Position Section 50.44 and associated changes tn Regulatory Guide 1.7 allowing credit for ECCS performance potentially reduced the amount of metal water reaction that is required to be considered in the containment design. This has permitted Hatch 2, a.4rk I BWR plant located in Baxley, Georgia, and operated by the Georgia Power Co., to operate without inerting.

In addition, it would permit Mark II BWR plants now in the operating license (OL) review process to operate without inerting.

By a ruling of the Appeal Board (ALAB-229, September 18, 1974; 8 AEC 425(1974)), inerting was not required for the Vermont Yankee plant, located in Vernon, Vermont, and operated by the Vermont Yankee

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I Nuclear Power Corporation.

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Generation

2. 5 Pre-TMI Assumptions on H2 Regarding the H2 generation assumptions prior to the THI-2 accident, the primary H2 sources were:

a.

A maximum of about 1 1/2 percent fuel-cladding water reaction b.

Radiolytic decomposition c.

Corrosion of paints and metals.

These sources are relatively slow generators and can be adequately dealt with by the use of H2 recombiners.

3.

RISK-BASED CONSIDERATIONS ON THE VALUE OF INERTING 3.1 History of Risk Considerations On several occasions, the views of the Probabilistic Staff (PAS) (now part of the Division of Risk Analysis) have been sought about the potential value of inerting in reducing the overall accident risks from BWRs with the Mark I and II containment designs.

In mid 1976, PAS views were sought when OSD was proposing inclusion of Section 50.44, concerning hydrogen control matters, into the Commission rules and regulations.

The net effect of Section 50.44 would have been to eliminate inerting requirements for nearly all of the BWRs.

The PAS view in mid-1976 was that inerting of the BWR Mark I con,tainment had very small value in terms of overall accident risk reduction.

The question of inerting and its value to risk reduction was again raised after the hydrogen burn experience at TMI-2.

PAS views were also sought again in the early part of 1980.

This time PAS views were incorporated into SECY 107B which provided the licensing staffs' interim requirements for hydrogen control.

The overall conclusions by PAS remained essentially unchanged from those given in mid-1976.

These were now shared by some of the licensing staff in the Risk and Reliability Branch of NRR.

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.t 3.2 Basis for Risk Perceotion The principal reasons why inerting would have small value in terms of reducing the overall BWR accident risk derives from available risk-based studies.

These studies indicate that the overall BWR risk would be very much dominated by accident sequences involving overpressure failure of the Mark I containment prior to core damage occurring.

Inerting yould therefore not logically offer any benefit to these risk dominating sequences since hydrogen would not be evolved until after loss of containment integrity had already occurred.

Inerting could, however, have some small value in those sets of sequences of lesser importance to risk where core degradation or melt would take place prior to containment overpressure failure.

However even for these sequences, inerting might only buy some increment of time prior to containment I

overpressure failure should the core damage or melt become extensive.

The PAS views stated in SECY-80-107B are reproduced below for information:

"The views of the PAS are contained in an internal men.,randum, "Value of Inerting to Overall Accident Risk Reduction," dated J ane 10, 1980, a copy of which is provided as Enclosure 1.

In summary, the PAS view is that "...inerting has small value in terms of overall accident risk reduction and it is believed that means exist that could have equal or greater value.

If an' urgency presently exists for inertin'g the Mark I and II containments, the bases are not found in risk-based studies of which the PAS is aware.

It should also be said that the PAS can presently offer no ove mhelming argument against I

an inerting decision exception for those views described above."

Some preliminary analyses by the Risk and Reliability Branch of NRR with probabilistic risk assessment (PRA) methodology leads to the same conclusion (see Enclosure 2).

However, this conclusion is subject to a fair measure of uncertainty because of the inability of the methodology to treat intermediate states for each event in the various scenarios. Operator intervention, including correct actions, incorrect actions, and delayed actions, similarly, cannot be ade-quately treated with the methodology."

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3.3 Differing Views of Risk Analyses While the results of probabilistic analysis methods are valuable, licens-ing decisions should not be. based entirely on the results of probabilistic analysis methods. There are a number of other factors that need to be consid-ered in arriving at a balanced judgment for licensing decisions.

Depending on 4

the issue at hand, these factors can include:

1) the uncertainties associated with the probabilistic analysis models; 2) the extent to which operator inter-vention could ameliorate or exacerbate the accident sequences; 3) the impact-i benefit ratios for the various mitigation measures; and 4) overall agency l

policy.

Thus, for example, in the case involving a transient followed by the total loss of pool cooling (RHR failure), which is one of the dominant contributors to risk, current PRA methodology would find that the containment would fail l

due to overpressure before any metal-water reaction can occur in the core.

For this case, having an inerted containment would not lead to any reduction I

in the off-site dose consequences.

However, the methodology used to arrive at the above conclusion did not give credit for the amount of energy transfer through the relatively thin shell of the steel drywell and torus. Calculations indicate that at temperatures ;

corresponding to near failure pressdras for the containment, i.e., about 350*F, the heat transfer from the Mark I steel containment due to natural convection and radiation would serve to limit the pressure rise of the containment atmos-phere. Additionally, if operator action with mechanically held fire hoses could lead to a dousing of the containment (torus) steel shell, then heat transfer rates would be increased by tio orders of magnitude, at the peak condi-tions, promoting sufficient heat removal to prevent containment failure from overpressurization.

The rate of containment pressurization for th.is scenario is sufficiently slow that ample time will be available for decision and imole-mentation of appropriate operator action. An inerted containment, in this l

scenario, would improve the plant's capability to tolerate subsequent and further degradation of the ECCS.

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4.

RECENT CASE POLICY New licensing requirements for dealing with hydrogen releases during postu-lated degraded core accidents have been under development since early 1980.

Item II.B.7 of the TMI Action Plan (NUREG-0660) stated that certain analyses would be performed in this area.

The results of these staff analyses were reported in the SECY 80-107 series of Commission Memoranda, " Proposed Interim Hydrogen Control Requirements for Small Containments." Commission guidance has been provided on these new licensing requirements in conjunction with its reviews of the various licensing cases and the various rulemaking activities.

In summary: form, pending the Interim Rule on hydrogen control, the staff is requiring that a significant fraction,(about 75%) of the fuel cladding be assumed to undergo metal-water reaction and that the design previde means for dealing with the hydrogen thus generated as follows:

(1) For all ice condenser and Mark III containments, additional hydrogen control measures have been required to assure safe shutdown and main-tain containment integrity; and (2) For the Mark I and Mark II containments, an inerted atmosphere is required.

A discussion of these new licensing requirements and how they have been implemented in recent operating license (OL) reviews, near term construction permit (NTCP) and manufacturing license (ML) reviews, and for Operating Reac-tors (OR) is provided below.

The format will be to discuss the cases in the inverse order of containment size; i.e., large dry, ice condenser and Mark III, and Mark I and Mark II.

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4.1 OL Reviews 4.11 Large Dry Containments The staff completed its reviews for the full power licen11ng of North Anna 2, Salem 2, Farley 2, and THI-1 (Restart) without imposing any new require-ment for hydrogen control.

This position was based on the staff's view that large dry containments had sufficient capability to accommodate hydrogen combus-tion so that no new requirements needed to be imposed, pending the long-term rulemaking proceeding on degraded core ~ cooling.

In the proposed rule portion of the Interim Rule, (SECY-81-245A), the staff proposes to require that certain specific analyses be performed for all large dry containments to assure that containment integrity and safe shutdown will not be jeopardized by hydrogen releases from the postulated degraded core accidents; i.e, involving 75% fuel cladding reaction with water. The staff is presently conside-ino w ether and when to begin implementatian of this new requirement on a large dry containment OL application, pending completion of the rulemaking on the Interim Rule.

4.1,2 Ice Condensers and Mark IIIs Full _ power licenses for ice ~ condenser plants are bei.ng issued with a license condition pertaining.to hydrogen control.

It requires that by Jinuary 31, 1982 a hydrogen control system be designed and installed wnich can accommo-date the hydrogen released as a result of 75% cladding-water reaction with adequate safety margins.

Sequoyah, Untt 1 and McGuire, Unit I were licensed on this basis; and all subsequent ice condenser plants are expected to be licensed on the same basis.

The subsequent plants include Sequoyah, Unit 2; McGuire, Unit 2; Watts Bar, Units 1 and 2; and Catawba, Units 1 and 2.

The situation with 0.C. Cook, Units 1 and 2, operating plants, will be discussed in Section 4.3.2.

These ice condenser units are all being equipped with a distributed igni-tion (glow plug) system (DIS).

The DIS is being designed to accommodate the 5 0 accident sequence, which is similar to the THI-2 accident, with up to 75%

2 cladding-water reaction and a hydrogen release rate as high as 70 lbs/ min.

The resultant peak combustion pressure is required to be less than the contain-ment failure pressure.

Peak temperatures are required to be less than that which would cause failure of certain essential equipment.

The response to 12

local detonations must be within the structural capacity of the containment.

All of these conditions must be met with adequate safety margins.

The first Mark III containment to be licensed for operation is expected to be Grand Gulf, Unit 1, now scheduled f'or the end of 1981.

The applicant, Mississippi Power & Light, has proposed use of a distributed ignition system (DIS).

The licensing basis for a DIS in a Mark III containment is currently being developed.

It is expected to be similar to the licensing basis used in the ice condenser containments.

Since this licensing basis probably won't be fully established until some time in 1982, and since this will be the first application of the OIS to a Mark III containment, interim licensing with a license condition similar to that imposed for Sequoyah, Unit 1 is likely.

4.1.3 Mark Is and Mark IIs The only Mark I and Mark TI boiling water reactors to be considered for full power licensing since the TMI-2 accident are the Fermi-2 and LaSalle, Unit 1 plants, respectively. There are no operating Mark II plants at tnis time. Our licensing basis for these plants is that they be operated with an inerted containment. With an inerted containment, these plants can accommo-date the 75% cladding-water reaction mentioned above.

4.2 NTCP Reviews There are only seven CP/ML applications covered by our NTCP requirements dealing with hydrogen control. They include plants with large dry containments, ice condenser containments and Mark III containments.

Our licensing requirements for these NTCP applications are d'etiailed in a recent proposed change to the Commission's rules published on March 23, 1981 and based on NUREG-0718.

A summary of these requirements is that:

(1) The minimum pressure capacity of the containment shall be no less than 45 psig; (2) degraded core scenarios involving as much as 100". cladding-water reaction are to be considereo in the containment design; (3) the maximum hydrogen con-centration uniformly distributed throughout the containment must be less than 10% by volume; (4) the hydrogen control system must keep the containment pres-sure and, therefore, the containment structural stresses within Service Level C limits; and (5) equipment essential to assuring safe shutdown and maintaining containment integrity must be able to survive the environmental consequences 13

of the degraded core accident.

In addition, containment. penetration (s) are to be provided for possible future use with a filtered vent system. These require-ments were made more severe than those established for OL and OR cases so as to minimize the extent to which construction of these plants would foreclose the ability to impose those new requirements that might develop as a result of the long-term rulemaking proceeding on degraded core cooling.

4.2.1 Large Dry containments Pilgrim, Unit 2 is the only large dry containment for which a hydrogen control system has been proposed by its applicant.

The other NTCP applicants for large dry containments have not yet filed their proposed approaches.

Pilgria, Unit 2 will be using the DIS.

Preliminary analyses show that the associated peak pressures and temperatures will satisfy the staff's accept-ance criteria. Design details and functional capability of the system will be furnished to the staff two years after issuance of the CP.

4.2.2 Ice condensers The only NTCP ice condenser plants are those associated with the manufactur-ing licenses for Offshore Power Systems (OPS).

Like Sequoyah and McGuire, the OPS plants will be using a DIS.

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If the DIS is found to be an appropriate long term solution to the hydrogen control issue for Sequoyah and other operating ice condenser plants, and it appears that it will, then satisfying the NTCP requirements for OPS with the DIS should be a straightforward task. The applicant will be given two years after issuance of'the ML to furnish the required supporting information.

4.2.3 Mark IIIs There are three Mark III plants among the NTCP applications. Two have proposed use of the OIS and one has proposed use of a post-accident inerting

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approach.

The DIS for the Mark III plants (Skagit and Black Fox) are likely to be similar to that proposed for the Grand Gulf plant.

The post-accident inerting approach proposed for the Allens Creek plant will involve the prompt injection of carbon dioxide by a system of sprays to achieve an 70% concentration some 14

15 minutes after actuation of the system.

Preliminary analyses by GE have indi-cated both approaches are feasible.

However, a number of issues will require further analyses and tests by the applicants.

We expect to give these applicants until two years following CP issuance to furnish the required supporting information.

4.3 Ooerating Reactors 4.3.1 Large Dry Containments The proposed rule portion of the Interim Rule includes a provision that licensees of plants with large dry containments perform certain analyses to show that safe shutdown and containment structural integrity can be maintained; i.e. essential equipment will survive the resulting environment. The require-ments are the same as those being considered for this type of containment at the OL stage of licensing (see 4.1.1).

These analyses would have to be sub-mitted by two years following the effective date of the rule.

Plants that cannot show these capabilities would be backfit, if necessary, under the provisions of 10 CFR S 50.109, "Backfitting."

4.3.2 Ice Condensers and Mark IIIs The only ice condenser plant that was operating prior to the TMI-2 acci-dent is 0.C. Cook, Units 1 and 2.

No Mark III plant has yet been licensed for operation.

Acting on Commission guidance, the staff has required the D.C. Cook licensee to install a hydrogen control system.

The licensee has proposed,.

installed, and received interim approvfl to operate a OIS similar to that of Sequoyah, Unit 1.

The proposed rule portion of the Interim Rule would also require these plants to perform analyses to justify the hydrogen control system selection and to assure containment structural integrity and continual core cooling.

These analyses'would have to be submitted by one year after the effective date of the rule.

4.3.3 Mark Is and Mark IIs There are presently 22 operating Mark I plants, two of which are not required to be inerted, i.e., Vermont Yankee and Hatch, Unit 2.

The final rule i

~T portion of the Interim Rule will require these units to be operated with an inerted containment.

There is no operating Mark II plant at this time.

For many years, the Vermont Yankee licensee has vigorously opposed our efforts to promulgate a rule requiring the inerting of all Mark I and Mark II containments.

They hold the view that the hazard to the operating crew and j

the tendency to defer in-containment inspections outweigh any safety benefits that might be gained by inerting the Vermont Yankee containment. They propose deferral of any inerting requirement until they can complete a study of alter _

native measures.

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5.

RESEARCH ON HYDROGEN 5.1 NRC's Research Program i

i The overall NRC research > ogram consists of eight major research programs.

Four research programs are aimed at developing an understanding of the phenomena associated with hydrogen burns, the methods for severe accident prevention and mitigation, and the effects of burns on equipment.

These programs are:

a.

Hydrogen Behavior and Control b.

Hydrogen Combustion, Mitig'ation and Prevention c.

Combustible Gas in Containment d.

Equipment Survival i

Three other research programs use the results of the above programs to assess the effects of hydrogen burn pressures on containment and finally one program which will include the effects of hydrogen generation from molten core-concrete interactions. This last group of four programs respectively are:

a.

Safety Margins for Containment b.

Safety Margins for Category I Structures c.

Effects of Hydrogen Explosions i

d.

Core Melt Technology Following is a brief description of these programs which, except as noted in the text, are being conducted at Sandia National Laboratory.

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5. 2 Hydrogen Behavior and Control Program The Hydrogen Behavior and Control Program is the largest of the eight programs (approximately $2M in FY 82) and is concerned with understanding the phenomena associated with hydrogen combustion and assessing the threat posed by hydrogen released during degraded core LWR accidents.

This program will generate information, procedures and equipment concepts which can be used to help prevent or mitigate the hydrogen threat.

5.2.1 Exce-imental Program This program includes experimental projects which are directed at determin-ing hydrogen deflagration and detonation limits in air and steam and noting how the location and strength of the ignition source affects those limits.

These experiments will be determining temperature and pressure profiles as a function of time, thus providing information which will be needed to develop and validate analytical models.

Additionally, tests are being planned to understand autoignition of hydrogen, particularly as it relates to hydrogen and steam jet releases, similar to what might occur from a pipe break.

A key area of-study is the work on the transition from deflagration to detonation which can occur as a flame propagates from one chamber to another or through a concentration gradient or accelerates around structures.

Flame acceleration in the upper plenum structure of an ice condenser containment was a concern which was raised relative to the Sequoyah distributed ignition system and sub-sequently in the McGuire hearings.

Flame acceleration occurs when a flame front bends around a structure.and begins to break up, inducing turbulence and developing more surface area, causing fhe flame to burn faster with higher temperatures and pressures.

Flame acceleration can cause lean mixtures to reach temperatures and pressures exceeding the theoretical adiabatic-isochoric limits.

Currently being planned are some engineering scale experiments to mock up the upper plenum structure of Sequoyah to assess this effect.

Addi' tionally, contractors at McGill University are perfonning some laboratory scale tests on flame acceleration and the transition from deflagration to detonation.

5.2.2 Analytical Prooram The analytical portion of the Hydrogen Behavior and Control Program involves several different aspects.

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5.2.2.1 Model Develcoment. A deflagration and detonation model (named OEFLAG and DETON respectively) is being developed to more accurately predict temperature and pressure histories in containment during and after a hydrogen combustion. The work on deflagrations will include the effects of CO, CO and 2

water fog evaporation.

The detonation model will be able to predict Chapman-Jouguet pressures and temperatures including increases after normal reflection.

Current effort is to include in the model heat transfer mechanisms such as radiation, convection with or without condensation, conduction into surfaces, and the evaporation of sprays.

5.2.2.2 Detonation Calculations.

A number of detonation calculations have been performed using the code CSQ in regards to the Sequoyah analysis l

(NUREG/CR-1762) and for the Zion study.

5.2.2.3 MARCH Calculations.

MARCH calculations have been performed for a number scenarios in which hydrogen is released from the.primar/ system.

These analyses have been done for Zion (large dry PWR), Sequoyah (ice condenser),

and are currently planned for Grand Gulf (BWR MARK III) and Surry (Subatmospheric PWR).

These analyses have provided useful information to NRR in their assess-ment of the Sequoyah interim distributed ignition system and will likewise be used in the evaluation of the hydrogen control system for Grand Gulf.

5.2.2.4 Hydrogen Transcort Calculations.

A difficult area of analysis which is being addressed deals with hydrogen transport. An attempt is being made to modify or' develop a code to predict the, concentrations of hydrogen, l

air and steam in containment as functions of position and time for hypothetical LWR accidents.

The German code RALOC is being assessed to determine its pre-sent and potential ability to handle transport and mixing analysis.

5.2.3 Hydrogen Comoendium and Generic Ooerators Manual The progr'am has generated and published a compendium (NUREG/CR-1561) of information concerning the behavior of hydrogen during a hypothetical LWR acci-dent.

The report addresses the questions of hydrogen generation, solubility, detection, combustion and recombiners.

This compendium will be updated as needed.

A generic LWR hydrogen manual is also being developed to provide information on the handling of hydrogen during and after an accident.

This 18

o manual can be used by reactor power plant designers and operators as a basis for preparing their own plant-specific operation and emergency manuals.

5.3 Hydrogen Combustion. Mitigation and Prevention Program The Hydrogen Coabustion, Mitigatior and Prevention Program was originally a part of the Hydrogen Behavior and Control Program but was separated with the RES reorganization in order to allow more emphasis to be placed in studying the design criteria and feasibility of proposed prevention and mitigation schemes.

Although under a separate Fin the programs are still closely coordi-nated and tied together.

5.3.1 Water Foq and Foams Experimental facilities and tests-being run under the Hydrogen Behavior and Control Program to characterize hydrogen deflagration and detonation are also including the effects of water fogs and foams as a mitigative approach to controlling temperatures and pressures.

A.05 percent volume fraction of sus-pended water droplets can reduce the temperature from a stoichicmetric mixture of hydrogen and air from approximately 2700 K to less than 1200*K with a pro-portional reduction in pressure.

5.3.2 0xygen Oeoletion and Post-Accident Inerting Planned also are sitigation experiments on the effectiveness and feasi-bility of oxygen depletion, pre-inerting and post-accident inerting with carbon dioxide and halon's.

5.3.3 Gas Turbine and High Caoacity Recombiners Other prevention and mitigation schemes, such as gas-turbines and high capacity recombiners, will also be assessed.

5.3.4 Deliberate Flarina from High point Vents An experimental investigation of the feasibility of a deliberate flaring technique in conjunction with a high point vent as a method for controlling hydrogen release during an LWR accident will also be performed.

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5.4 Combustible Gas in Containment Program Other potential sources of hydrogen are from the corrosion of zinc, alumi-num and other coatings and from the interaction of molten core material with concrete.

The first area is being addressed in the Combustible Gas in Contain-ment Program which is determining the rates of hydrogen generation from galva-nized steel and will be looking at the kinetics for other materials.

Another facet of this program is to quailtify and characterize the type of precipitate that can be formed in these reactions in order to assess a potential problem of sump blockage.

5.5 Core Melt Technology Program The second area of hydrogen generation mentioned above is included as part of the Core Melt Technology Program wherein a number of field-scale tests are being perfonned to characterize and model the thermal, chemical and mechanical interaction of molten fuel and steel with concrete and with potential core-retention materials, both in the presence and absence of water.

This program will generate information on hydrogen evolution rates as a function of time, penetration rate, type of material and other paramete~rs.

The area is important because large releases of gases (H, CO, H O vapor) can place significan't loads 2

2 on containment.

5.6 Eouioment Survival Program The combustion of hydrogen in containment can lead to the failure of impor-tant safety equipment or equipment necessary for plant isolation during an acci-dent.

The Experimental Equipment Survival Program was initiated in order to experimentally assess the effects of hydrogen combustion on equipment.

This program also provides a data base in order to develop analytical models to assess equipme'nt survivability. NRR is sponsoring the other half of this program which is to develop analytical methods to calculate the effects of hydrogen combustion on equipment. The experimental facilities used in the Hydrogen Behavior and Control Program will b'e used to actually test equipment in hydrogen burning environments. Testing of equipment will not begin until i

20

FY 82.

There are larger facilities such as the Radiant Heat Transfer Facility available in the event of a need to test large components.

5. 7 Safety Margins for Containment Program All of the above programs (except equipment survivability work) provide data which will be used in assessing the threat to containment from hydrogen combustion.

This information will be used in the Safety Margins for Contain-ment Research Program which is aimed at developing an understanding of failur.e modes and margins of failure for various types of containment.

The objectives of the program are to develop reliable methods of predicting containment behavior, identifying limit states due to accidents and natural phenomena and to perform correlations of tests and analytical results.

This program will address both uniform internal pressure loadings and unsymmetrical dynamic pressure loads (combustion loadings) for various containment types.

5.8 Safety Margins for Category I Structures Program In %ddition to the work on containment buildings a similar program exists for Category I Structures.

5.9 Effects of Hydrogen Exolosions Program A small program is being empleted I'n FY 82 which was aimed at developing simplified behavior models for containment response to H expl sions for use 2

~

in risk assessment calculations.

This program is being conducted at MIT and l

provides a link between the Hydrogen Behavior and Control Program with proba-bilistic risk assessment analysis.

l 5.10 NRC Technical Assistance Contracts on Hydrogen Behavior There are a number of NRC technical assistance contracts providing case specific information on ::jdrogen behavior.

As part of an independent assess-ment of the efficacy of the Sequoyah hydrogen control system, NRR is sponsor-ing a series of tests on the GM glow plug installed in Sequoyah.

The tests 21

are designed to examine the performance of the igniter under a spectrum of con-ditions (principally varying hydrogen and steam concentration). This program j

includes a modification of the CCMPARE Code to predict the atmosphere pressure and temperature response to hydrogen burning. An assessment of the Grand Gulf hydrogen control system is being perfo'rmed in coordination with the Hydrogen Behavior and Control Program to assess the operability of the ignition system under various environmental conditions.

As noted earlier model development for equipment survivability is being performed under an NRR technical assist-ance contract. Models developed in this program will be validated with the results of the Equipment Program.

A technical assistance contract to assess the effects of hydrogen detonation in Sequoyah has been expanded to develop models for five steel containments.

A prcgram expected to be completed in early 1982 will develop requirements for installation of high point RCS and reactor vessel head vents that are remotely operable from the control room.

Additionally, NRR is planning work to evaluate the design basis for hydrogen control systems for certain construction permit and operating license appli-cations. This work will include analysis of the most probable accident scenarios that can lead to 100% clad / water reaction, the effects of recovery of various failed systems in subsequent stages of the scenarios and timing of-the actuation of the hydrogen control system relative to accident scenarios.

5.11 00E, EPRI and Industry Programs on Hydrogen A number of research programs on hydrogen are being sponsored by 00E, EPRI and the nuclear industry (IDCOR) with the majority of the work being done by EPRI.

~

5.11.1 EPRI Research Program on Hydrogen EPRI will be spending approximately a million dollars a year for the next two years on hydrogen research.

The program elements of the EPRI program are as follows:

(1) Development and preliminary testing of deliberate ignition devices.

l l

(AECL-Whiteshell and Acurex Corp.).

22

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(2) Analyses and experiments on basic hydrogen combustion phenomena including the effects of steam, turbulence and the potential for transition to detonation (AECL-Whiteshell).

(3) Experiments on hydrogen control methods including water spray and fog (Factory Mutual Research, Acurex).

(4) Measurement and analysis of hydrogen mixing and distribution with natural and forced circulation (HEDL-W).

(5) Demonstration of hydrogen combustion and management techniques on a large scale (Nevada Test Site).

(6) Development and validation of analytical tools for hydrogen mixing and burn velocity, pressure ar.d t.emperature (HEDL, AECL, FMRC, EPRI).

In addition to the above program elements, EPRI is also investigating the possibikity of performing some hydrogen burn equipment survivability tests in conjunction with the combustion experiments.

The large scale tests at the Nevada Test Site are currently being costed and a final decision on this-program element has not been made.

RES staff has reviewed the EPRI program and has had preliminary discussions.

There are a number of potential areas for joint cooperation (e.g., the large scale Nevada tests), however, nothing has been finalized as of yet.

A number of the EPRI tests'will provide reason-ably good comparisons with the NRC worff and will be useful in deducing the effects of scale and different geometries.

The large scale mixing tests in the Containment Systems Test Facility (CSTF) vessel at HEDL (these tests are sub-sidized by 00E) are very useful for comparison of transport and mixing analysis and the results will be made available to RES for code comparisons.

A number of the models on hydrogen combustion phenomena being developed by EPRI will be simplified thermodynamic calculations (as opposed to kinetics) and their program does not address in any detail flame acceleration phenomena.

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5.11.2 Owners Groues Soonsored Hydrogen Research.

The Ice Condenser Owners Group has sponsored a series of experiments by Fenwal, Inc. to evaluate the efficacy of igniters. The testing was conduct M using a single igniter assembly in a spherical vessel approximately 6 feet in diameter (134 ft.3).

~

Tests were conducted as a function of the concentration of hydrogen (4-12%),

pressure, temperature, turbulence (induced by fans) and steam concentration.

Additionally, tests were run to assess the effects of sprays and to note the effect of hydrogen burn on typical equipment, e.g., limit switch and solenoid valve, by measuring the equipment temperature response.

The results of the Fenwal test program were. generally consistent with the applicable published information on hydrogen combustion.

TVA has also conducted tests to obtain pertinent performance characteristics of the glow plug igniters installed in Sequoyah.

There is a BWR MARK III Owners Group which is currently planning to sponsor some research on hydrogen behavior and control in coordi-nation with the EPRI Program. Their plan will be available in early-mid FY 82.

5.11.3 IDCOR Hydrogen Program. Technology for Energy Corporation (TEC) as part of its IDCOR program will, over the next 16-18 months, develop review papers j

on:

Hydrogen Generation Rate, Hydrogen Distribution Combustion Limits of H -Air-2 Steam-C0, Survey of H2 and 02 Detectors, Evaluation of Pre-Inerting, Evaluation 2

of Fogging / Spray Suppression and Hydrogen Burn Control Systems Evaluatich.

These papers will be based on available information and current programs (NRC, EPRI),

and will make recommendations fo'r the need for further work, develop best esti-mates and assist the industry in assessing the risk reduction for proposed mitigs-tion schemes.

Currently no new work on hydrogen is proposed by TEC.

5.11.4 00E Hydrogen Program.

DOE's effort on hydrogen safety research is limited to an assessment of hydrogen monitoring equipment for various environ-mental conditions.

This is a follow on to an NRC program to review the avail-able and currently used hydrogen instrumentation.

DOE picked up this program at NRC request.' As noted earlier 00E is providing its facility (CSTF) at HEDL for the hydrogen transport work.

Additionally, as part of its Plan For Improv-ing Safety of Nuclear Power Plants, DOE proposes to provide an internal organi-Iation that can coordinate a National Hydrogen RD&D Program in cooperation with industry, NRC, other Government bodies, and foreign countries to understand hydrogen behavior in LWR plants, and develop methods of handling hydrogen in potential reactor accidents.

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