ML20127C490
| ML20127C490 | |
| Person / Time | |
|---|---|
| Issue date: | 02/21/1985 |
| From: | Houston R Office of Nuclear Reactor Regulation |
| To: | Butler W, Lear G, Sheron B Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20126E126 | List: |
| References | |
| FOIA-85-127 NUDOCS 8502270651 | |
| Download: ML20127C490 (2) | |
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tb UNITED STATES j
NUCLEAR REGULATORY COMMISSION
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February 21, 1985 9 ;Ra.G / ::7.
Addressees
- k. Wayne Houston, Assistant Director for Reactor Safety, F:.:d.:
Division of Systems Integration SUEJECT:
STAFF REVIEW OF HCOG PROGRAM PLAN As a rest,it of the recent series of meetings with HCOG, we are committed to prepare ar.c.ransmit to HCOG our evaluation of the program plan in essentially two chases.
ins first phase involves those tasks relating directly to the 1/4 s:aie tes
- *er-am, viz., Tasks 1, 7, 9 and 12.
Our evaluation is targeted fcr transristien to HCOG on or before March 22, 1985.
The second phase involves :ne en. ire plan and our evaluation is targeted for transmission in tic-April.
- would like to rneet with you or your designated representative in my office on Monaay, Fe.br k ry 25, at 1:00 p.m. to discuss the scope, content, and ourpose c' 07ese eyaluations.
.A Y
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., w R.' Wayne Houston, Assistant Director
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for Reactor Safety, Division of Systems Integration cc:
R. Bernero, DSI K. Parczewski, DE J. Rosenthal, DSI H. B. Clayton, DHFS M. Wigdor, DSI M. McCoy, DHFS J. Kudrick, DSI T. Novak, DL C. Tinkler, DSI G. Lainas, DL H. Gar;, CE J. P. Knight, DE C. P. Tar.. ~E W. V. Johnston, DE H. E. Polk, DE M. Fleishman, RES 75Mamp p
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Addr2ss22s FIS 21 1985 ADDRESSEES:
.e Walter Butler, CSB/DSI Brian Sheron, RSB/DSI George Lear, SGEB/DE Vince Noonan, EQB/DE Vic Benaroya, CMEB/DE Ashok, Thadani, RRAB/ DST Dennis Zeimann, PSRB/DHFS Carl Stahle, LB#4/DL 9
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-i MARK W CONTAINMENT HYDROGEN CONTROL OWNERS GROUP som H. Hoees. Cnoi, mon c/o Mississippi Power and Light e P.O. Dox 1640 e Jackson. Mssassippi 09205 601 969 2458 February 28, 1985 HGN-030 l
U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C.
20555 Attention :
Mr. Robert Bernero Subject
- Hydrogen Control Owners Group Program Plan HGN-030 Referenge 1: Letter HGN-024 from Mr. S. H. Hobbs to Mr. H. R. Denton dated December 14, 1984.
Reference 2: Letter HGN-028 from Mr. S. H. Hobbs to Mr. R. Bernero dated Feb. 14, 1985.
Reference 3: Letter HGN-029 from Mr. S. H. Hobbs to Mr. R. Bernero Feb. 19, 1985.
Dear Mr. Bernero:
During a meeting between members of your staff and the Hydrogen Control Owners Group (HCOG) on February 20, 1985, the HCOG couanitted to submit a revision to the Hydrogen Control Program Plan submitted in Reference 1.
Enclosed are 24 copies of Revision 2 to the Hydrogen Control Program Plan. This revision incorporates codifications to the task descriptions, subtask descriptions, and task networks. These modifications implement changes in the acceptance criteria which were transmitted to your staff in Reference 2 and correct editorial errors identified in reviewing the document.
The first page for each set of Revision 2 contains detailed instructions on replacing pages in the current program plan with Revision 2 inserts. These instructions are based on the assumption that Revision 1 of the program plan, which was formally transmitted to you by Reference 3 has been inserted into each copy of the program plan.
l The HCOG requests that the NRC staff provide a written evaluation of the i
program plan and acceptance criteria for tasks 1, 7, 9, and 12 by March 22, 1985. This schedule is important for supporting the start of testing at the 1/4 scale test facility. The HCOG further requests that the NRC yrovide a
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y Safety Evaluation Report documenting acceptance of the complete program plan by April 9, 1985.
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J0P12HGN85022801 - 1 J
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Page 2 of 2 HGN-030 This submittal was compiled by NCOG from the best information available for submittal to the Nuclear Regulatory Commission. The submittal is believed to be complete and accurate, but it is not submitted on any plant specific docket.
The information contained in this letter and its attachments should not be used for evaluation of any specific plant unless the information has been endorsed by the appropriate member utility. HCOG members may individually reference this letter in whole or in part as beinz applicable to their specific plants.
HCOG sppreciates the NRC staff's support to date in reviewing the program plan,
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Please contact me if you have any questions concerning this submittal.
t Sincerely.
Y S. H. Hobbs Chairman, Hydrogen Control Ownera Group SHH:rg Attachments cc:
Mr. Carl R. Stahle Hydrogen Control Program Manager U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C.
20555 Mr. Charles G. Tinkler Containment Systems Branch U. S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, D. C.
20555 Mr. John Cummings Project Manager Hydrogen Studies Division 4441 Sandia National Laboratory Albuquerque, NM 87185 J0P12HCN85022801 - 2 4
INSTRUCTIONS FOR INSERTING REVISION 2 OF THE HYDROGEN CONTROL PROGRAM PLAN 1.
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- 77. Replace page 1 Task 10 network with new network dated 2/28/85
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behind new Tab "Section 6".
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i An overall description of each of these major Tasks is provided in the following Sections.
Each Task is described with a brief statement of purpose followed by a description of each Subtask in the program.
After each section the o r g a rri z.a t i o n responsible for completing work associated with the subtask and j
the current status of the subtask is identified.
Generic work conducted by the Mark III Containment Hydrogen Control Owners
'l Group is identified as HCOG.
Plant specific work is identified i
as the Utility.
For generic work the HCOG is identified as the responsible organization where the HCOG has selected a i
subcontractor the subcontractor's name is shown.
However, the HCOG maintains overall responsibility for the subtask. These descriptions provide an outline of the work planned or completed to achieve the task goals.
For subtasks which are the
]2 responsibility of individual utilities, the status may be 12 indicated as "completa'! when one or more utilities have finished 12 the indicated subtask.
Following the Task and Subtask 12 descriptions is a statement of Acceptance Criteria which defines successful closure or completion of the major task.
HCOG plans to conduct the necessary research or analyses to satisfy the Task Acceptance Criteria and these criteria are considered a key part of each HCOG program for task closure.
Also included are the Task Networks which provide a visual representation of the planned work flow and the logical interaction between each Task, Subtask and Decision Point.
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d TASK 1 - ESTABLISH MOST PROBABLE HYDROGEN GENERATION EVENT J2
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The NRC's proposed interim rule, published for comment in December 1981, required consideration of a 75% metal-water reaction and the associated hydrogen release.
The rule
. j specified the quantity of hydrogen to be considered but did not j
specify the accident scenario postulated to produce the l
This was followed by the NRC's issuance of NUREG 0718 l
which recommended near term construction permit (NTCP) applicants to consider accidents generating hydrogen equivalent to a 100% metal water reaction of the fuel cladding.
1 The scenario postulated to produce hydrogen had impact on the mitigation system selection, design and analysis.
Inherent in the definition of a scenario was the determination of plant conditions at the time of hydrogen release, the functional availability of various accident mitigation systems, the response of operators, the loading conditions postulated for the containment structures and equipment, and the environmental conditions which equipment must be able to withstand.
Since there is no credible sequence which produces the specified hydrogen releases in the BWR/6, the hydrogen generation event j
(HGE) was based on a combination of deterministic and probabilistic criteria.
A probabilistic approach was used to identify the most representative accident initiating event for a HGE, and deterministic considerations were used in defining a realistic termination of the HGE, before significant core melt occured.
Plant and system conditions were specified consistent with HGE conditions and termination of the event.
This task is illustrat'ed as Task Network 1.0.
5 Two specific accident sequences will be evaluated to determine 12 if they make any significant contribution to the probability of
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i a recoverable HGE.
These events are anticipated transients
}2 without scram (ATWS) and Station Blackout (SDO).
Data will be
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evaluated to determine the contribution of these events to
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overall risk of HGEs.
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1.1 Define Hydrogen Requirements On October 2, 1980 the Nuclear Regulatory Commission published in the Federal Register a notice of proposed rule making to dictate new hydrogen control requirements for degraded core j
accidents.
The proposed rule required utilities to design systems capable of controlling hydrogen released from the interaction of seventy-five percent (75%) of the zircaloy fuel cladding with water, described an approach to qualify essentia) equipment before and after a hydrogen burn, and required design analyses to evaluate the effects of the hydrogen generat' ion event.
NUREG 0718 required NTCP applicants to meet the even more stringent requirement of designing systems capable of controlling hydrogen released from the interaction of one hundred percent (100%)..of the zircaloy fuel cladding with water.
The quantities of hydrogen described by the proposed rule and NUREG-0718 are substantially greater than previous design requirements.
Thus the proposed interim rule and NUREG 0718 have essentially mandated the design of a completely new hydrogen release control system.
The Hydrogen Control Owners Group has provided comments on the various stages of the rulemaking process in the following transmittals:
HGN-004 dated 4/8/82, HCOG to S. J. Chilk HGN-005 dated 4/8/82, HCOG to R. M. Bernero l
HGN-013 dated 11/7/83, HCOG to N. J. Palladino Responsibility - NRC Status
- Complete
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used to identify the most representative accident scenario for a hydrogen generation event (HGE) and deterministic considerations l
l were applied to define a realistic time period for ter,mination f
of the HGE before significant core melt occurred.
Using' this l
methodology a spectrum of possible accident scenarios were produced. This evaluation was completed for HCOG by GE and the i
results included in " Report on Hydrogen control Accident Scenarios, Hydrogen Generation Rates and Equipment Requirments".
This report estimated that the combined probability for all consolidated events which could lead to core melt was on the order of 1 x 10-6 per year.
The most probable HGE sequence was modeled as a turbine trip with bypass and loss of feedwater.
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Failure of all other makeup systems resulted in a drop of RPV level due to inventory loss through the bypass pressure control
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system until vessel iblation occurs'.
Reactor pressure is then
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controlled by SRV action until the RPV is depressurized by the j
operator per the emergency procedure.
Event recovery is started l
when the HPCS injection is recovered and water injection established.
Variations in the timing of operator actions to 12 recover core makeup systems were included in the study.
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i Responsibility - GE Status
- Complete l
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i 1.5 Submit Report To NRC l
The work for Subtask 1.4 was co:npleted in 1982 and submitted by i
HCOG to the NRC.
The results of this work were described by
" Report on Hydrogen Control Accident Scenarios, Hydrogen Generation Rates and Equipment R>equirements" and submitted as an 1
attachment to HGN-003.
A revision to this report to incorpora N~
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i additional work requested by the NRC staff was submitted to the I
l NRC as an attachment to HGN-006 dated September 9,1982.
i Responsiblity - HCOG Status
- Complete 1.6 Evaluate ATWS and 850 Accident Econarios
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The work under task 1.4 was completed in late 1982 and early
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1983.
Since that date, additional information within the 1
nuclear industry has been developed on the probability of
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Anticipated Transients Without Scram (ATWS) and Station Blackout
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l (SBO) Events.
The HCOG intends to re-evaluate the contribution
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of ATWS and SBO events to the overall probability of recoverable
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HGEs.
The results of this evaluation will be provided to the
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NRC staff for review.
1 Responsibility - HCOG
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Status
- Not Started
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4-6 2/28/85 J
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production to manually initiate the system.
HCOG has 12 recommended that the ignitors should be actuated when the
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reactor pressure vossol water level reachos the top.of active
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fuel (TAF).
Thoro should be a minimum of 10 minutos atra'ilable 1
betwoon the time when the reactor pressure vossel water level
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reaches TAF and the time when significant hydrogen production
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commences.
In addition, llCOG recognized that operation of the 1
ignition system when hydrogen was not present would not af fect the plants.
Rosponsiblity - IICOG Status
- Comploto 3.7 Location, Redundancy and Soparation Critoria Design critoria for the ignitor systems woro developed generically by llCOG.
Critoria for locatic.n, redundancy, and separation, woro developed.
Each member utility has also considorod high energy pipo-whip and jet impingement in the ignitor locations.
The ignitor location critoria also specify that hydrogon burns in the drywell, watwoll, upper containment and equipment rooms must be assumod.
The ignitors wore spaced by each member utility to minimize the potential for hydrogen accumulation which could load to local detonations assuming a single failure.
Ignitors will not be separated by more than 30 feet when all amorgency safoguard featuro (ESP) power supplios are operable.
Operable ignitors will not bo separated by moro 12 than 60 foot when one ESP power supply is inoperable.
Two exceptions exist to those requirements.
Ignitors are moro 12 widely spaced in the largo open regions above the refueling floor where hydrogon pocketing cannot occur and in the lower portions of drywell which are subject to flooding.
Each member ilIS in also designed to be tasted during normal plant operation Responsiblity - IIC00 Status
- Completo 4-22 2/20/05
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t 3.8 Locate Ionitors l
Using the design criteria that were developed by the, Hydrogen l
Control Owners Group, in Subtask 3.7, general igniter locations l
were developed by each utility member.
This assured a high
,j degree of consistency in the approach used by the individual utility members to finalize igniter locations in subtask 3.9.
l
'i Responsibility - Utility l }
Status
- Complete 3.9 Exact Toniter Locations and Support Design Each utility member has made final igniter location decisions based on a detailed review of plant specific geometry and support member availability.
Due to availability of support 12 members and construction interferences, it was occasionally not
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possible to separate ignitors with both ESF divisions operable
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by only 30 feet.
However, final igniter locations assure that
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with both ESF divisions operable, ignitors will not be separated
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l by more than 35 feet.
With one ESF division inoperable,
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I operable ignitors will not be separated by more than 70 feet.
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The two exceptions to thin are identified under subtask 3.7.
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Supports for ignitors were designed to withstand seismic, hydrodynamic, and pool swell loads.
Final locations have been determined by each member utility.
Responsibility - Utility Status
- Complete 3.10 Desian of Ioniter Power Sucolv l
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I Each utility has reviewed the intial design of the igniter power supply that was prepared for HCOG by MPAL.
The igniter devices 12 t
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are supplied with nominal 120 VAC power which is converted by a
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transformer to 12.0 V before being applied to the glow plug.
1 Each utility has determined what failure modes are possible and I
considered the effects in the power supply design.
In addition
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each member utility has added the igniter loads to appropriate
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ESF emergency power supplies.
Responsibililty - Utility I
i Status
- Complete l
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3.11 Install Ionitors
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Each utility member is responsible for igniter and associated
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power supply and control system installation.
This installation f
process is dependent on each plants projected startup and construction schedule.'
System testing and confirmation of operability will be completed prior to full power operation for each plant or as allowed by specific plant license conditions.
I Responsibility - Utility 3
Status
- In Progress l
3.12 Develop Generic Tech Snecs i
Generic technical specifications will be developed as a f
guideline for each member to use to develop plant specific l
technical specifications.
The technical specifications approved for MPAL will be considered in developing a guideline for the i
other HCOG members.
Responsibility - HCOG
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Status
- In Progress
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4.3 Burn Static or Dynamic MP&L has completed an analysis for the GGNS which indicates that the pressure loadings produced by deflagrations occur byer a period of several seconds.
Since the pressure loading occura over a relatively long period or time, the HCOG elected to use l
this analysis as the basis for treating these loads as static loads.
Responsibility - HCOG Status
- Complete 4.4 Containment Ultimate Capacity Analysis Each utility has completed a containment ultimate capacity analysis.
This analy's'is was conduct'ed using specified material properties or information from certified material test reports for the containment structure, the drywell, airlocks, and major penetrations.
The capability of all local components such as
]2 airlock and hatch seals, penetrations, etc. has been evaluated
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to assure that these components have the capability to withstand
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at least the pressure determined to be limiting with respect to
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the overall structure.
The containment ultimate capacity
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analyses also establish that the Mark III drywell will withstand
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large internal pressures.
Either the ultimate capacity of the
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drywell structure has been established, or the drywell has been
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shown to have a pressure retaining capability which considerably
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exceeds the peak pressure which might be produced by combustion
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in the drywell.
Individual reports for assumptions and 1
methodology has been submitted by member utilities to the NRC
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staff.
Responsibility - Utility Status
- Complete 4-30 2/28/85
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4.5 Containment Negative Capacity Analysis Each utility has conducted and completed a negative. capacity analysis to determine the maximum external pressure load'on the containment and drywell which could result from hydrogen l'
combustien or has documented the very large external pressure i
capacity of the containment. In cases utilities have shown that the containment and drywell can withstand external pressure loads which exceed the maximum external pressure which could result from hydrogen combustion.
Individual reports of assumptions and methodology has been submitted by member utilities to the NRC staff.
Rosponsibility - Utility Status
- Complete 4.6 Consider Load Due to Local Detonation The hydrogen igniters were located to preclude the accumulation of detonable concentrations of hydrogen in equipment rooms or other enclosed areas.
The open geometry of the Mark III and turbulence further minimizes the possibility of local detonations.
HCOG has concluded that local detonations will not occur in the Mark III containment.
This conclusion was based upon extensive literature research completed by MP&L.
This conclusion was documented in HCOG's response to an NRC request for additional information (RAI).
The response to this RAI was 12 submitted by letter HGN-011 dated May 11, 1983.
I Responsibility - Utility /HCOG
]2 Status
- Complete
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4.7 Document Exclusion of Local Detonation Each utility has the responsibility to document to the NRC that 4-31 2/28/85
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I local detonations are not sufficiently prob $ble to warrant consideration or that no significant containment pressure effect occurs. This will include a discussion of plant specific l
features which preclude the accumulation of hydrogen concentrations to detonable levels.
l Responsibility - Utility Status
- Complete l
4.8 Verify Containment Canability
't Each utility has verified that the peak containment pressure produced by postulated hydrogen combustion is below the containment ultimate capacity.
The reports demonstrate that each element of the containment pressure boundary is capable of withstanding the peak')ressures prod'uced by hydrogen combustion.
The reports also demonstrate that the drywell is capable of 12 withstanding the peak pressure which might be produced by
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combustion in the drywell.
1 Responsibility - Utility Status
- Complete 4.9 Submit Ultimate Canacity Analysis to NRC Each utility member has prepared an Ultimate capacity Analysis report for the NRC.
This report contained appropriate details of analytical methods, assumptions and evaluation results.
The reports documented that the ultimate containment capacity exceeded the peak pressure which would be produced by hydrogen combustion.
1 Responsibility -
Utility Status Complete 4-32 2/28/85 J
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5.1 Survey Available Code l
A survey of potential codes which could model the Mark III containment and hydrogen combustion phenomena was conducted by HC00.
This survey determined that no codes which existed at the i
.i time could adequately treat the postulated combustion of l }
hydrogen in a Mark III Containment.
l Responsibility - HCOG Status
- Complete t
5.2 Select Code l
This subtask required selection of a code to analyze the postulated hydrogen combustion in the Mark III containment.
l Selection of an existing code without modification was not a viable option, as determined in Subtask 5.1.
During the survey conducted in Subtask 5.1, the HC00 found that of fshore Power Systems (OPS) had used a code titled CLASIX for performing analysis of hydrogen combustion in the Sequoyah, McGuire, and D.C. Cook containments.
In addition, OPS used CLASIX to analyze 12 the floating nuclear reactor plant design.
It was determined that CLASIX could model many elements common to the Mark III containment and OPS had prepared a plan for verifying this code.
Based on these facts, the HCOG determined that it would be more cost effective to modify the CLASIX code to correctly simulate
{
the Mark III containment rather than develop a new code.
Responsibility - HCOG Status
- Complete 5.3 Modify CLASIX to Include Mark III reatur g The CLASIX code was developed to model hydrogen burn transients 4-36 2/28/85 1
t i
.-~
..~.~.---. -.~.-....
.~
l l
5.7 submit Tonical Recort Verifying CLASIX to NRC l
i l
The CLASIX-3 code is an adaptation of the ice condenser containment code, CLASIX.
The CLASIX code was used in the containment response analysis of the sequoyah, McGuire and and.
L D.C. Cooke plants.
Formal verification of the base code was completed and has been submitted to the NRC in a topical report.
ll Verification of the CLASIX-3 code, has also been completed and ll submitted to the NRC in the form of a topical report as 1
described in subtask 5.9.
Responsiblity - HCOG Status
- Complete 5.8 Determine if Additional Verification is Required The Hydrogen Control Owners Group has reviewed the verification work completed to date and has determined that additional analysis is required. Comparison with similiar codes for 12 containment response indicates that CLASIX-3 provides reasonable results for accidents which can be analyzed with other containment response analysis codes.
In addition, a number of sensitivity studies have been completed which show that the code predicts reasonable variations in program output for variations in code input.
Although all verification work completed to date 12 indicates that CLA8IX-3 adequately predicts deflagration type
)
combustion in the Mark III containment, HCOG intends to evaluate
]
CLASIX-3 predictions against large scale data relevant to the
)
Mark III containment geometry.
This data is limited to data
)
which will be obtained in the Mark III '/4 Scale Test Program.
1 The specific CLAs!X-3 comparisons will be made as part of Task
)
y 12 in this program.
1 I
)
Responsibility - HCOG l
Status
- Complete i
4-39 2/28/85 4
... ~
~-----...E.~I~~~-...'~...~-~--.T' t
l l
5.9 Prepare CLASIX-3 Verification Report The efforts which have been completed to verify the CLASIX and CLASIX-3 programs provide substantial assurance that these codes reasonably predict deflagration combustion in the containment following an accident which releases large quantities of hydrogen.
Complete details of this verification process were
'I prepared by OPS and included in the CLASIX and CLASIX-3 Topical Reports.
Verification reports included numerous hand calculations, comparisons of particular code models with other i
accepted codes and an independent verification of the suppression pool model.
Reponsibility - OPS Status
- Complete 5.10 Submit CLASIX-3 Verification Report to NRC l
l A report describing the CLASIX-3 code was prepared for the I
j Hydrogen control owners Group by Westinghouse and submitted to the NRC.
This report, titled, "The CLASIX-3 Computer Program for the Analysis of Reactor Plant Containment Response to l
Hydrogen Release and Deflagration", was submitted to the NRC as I
an attachment to HGN-009 dated March 19, 1983.
i Responsibility - HC00 status
- Complete i
l t
4-40 2/2e/as 1
i i
.......;........~............._,_,,.-
r t
I requirements, and flame velocity.
The burn parameters were determined based on COMBEX's* experience and upon a review of the results from the Electric Power Research Institute's testing program.
I 1
The burn parameters which have been used are flame velocity of 6 l
f t/second, combustion initiates at a bulk average hydrogen concentration of St and 456 of the hydrogen is burned.
j For 12 1
oxygen limited or steam limited conditions, it is assumed that
{
combustion will be initiated with a bulk average oxygen j
i concentration of 54.
100% of the oxygen will be consumed in the i
resulting burn.
Hydrogen combustion should not occur when steam concentrations exceed 504 by volume.
i l
Responsibility - HCOG Status
- Complete 6.10 Investicate Mark III Uniaue Phenomer,a l
Associated with the Mark III containment is the possibility of inverted diffusion flames in the drywell.
This could occur when hydrogen and steam displace the oxygen in the drywell, followed by a gradual re-introduction of oxygen through the drywell vacuum breakers, the drywell purge system, or other air flow paths into the.rywell.
This phenomena will be evaluated by Suhtaaks 10.12 through 10.20 to determine the effects on the drywell thermal environment.
l Another phenomena unique to the Mark III containment was identified as the potential for hydrogen combustion to occur et-l the surface of the pressure suppression pool.
Degraded core hydrogen generation scenarios established that conditions in the upper containment and wetwell involved the release of hydrogen 4-49 2/28/85 J
] U J.....T._..
_.....e..........~.
.....m......
l I
a i
1 6.24 Lareer Scale Testino Required?
i In comparison of test results from the 1/20 scale f acility and l
1/5 scale single sparger tests, it was observed theti flame heights in the larger scale were much lower than the flame l,3 heights which would have been predicted from 1/20 scale
! J measurements.
The 1/20 scale data appears to overpredict flame heights and subsequent thermal environments.
The 1/20th scale i
was not adequately tristrumented to provide a definition of the l
j thermal environment for all locations.
To obtain more i :
definitive data on the thermal environments, the NC00 decided to l
conduct additional tests of the Mark III containment response in a larger scale facility.
Responsibility - HC00,,
Status
- Complete 6.25 Initiate 1/4 Scale Test Froaram As a result of the NCOG's decision to conduct large scale testing, several organisations were evaluated for their ability to conduct a 1/4 scale, 3600 diffusion flame test program.
The Electric Power Research Institute was previously selected as the 1/4 scale test program manager.
The NCOG selected Factory Mutual Research corporation as the major subcontractor.
Selection of these organisations initiated Task 9.0, " Diffusion Flame Thermal Environment 1/4 Scale Test Program".
Responsibility - HCOG Status
- Complete t
6.26 Evaluate TVA Ioniter scray Tests 12 i
l The Tennessee Valley Authority (TVA) completed a series of tests
]
l r
4-56 2/2s/as
~.....
~
i to evaluate igniter performance in a spray environment.
These
]2 tests monitored surface temperature of Tayco A.C.
igniter and
)
GMAC model 7G glow plug ignitors in a spray environment.
HCOG
)
will evaluate applicability of these tests to the Ma'rk III
)
containment and to the NCOG igniter system design.
HCOG will
)
l :
report the results of this evaluation to the NRC.
]
I t Responsibility - HCOG
)
, I 1
j Status
- Not Started
)
1 I
t t
I t
M 4-56a 2/2s/ss e
.,,. ~......
..s.
._.s..
n,,..
.a TASK 7 - GENERATION OF HYDROGEN RELEASE HISTORIES Early work completed by the Hydrogen Control Owners Group utilized the MARCH code for calculating hydrogen production from degraded core accidents.
The MARCH results predicted a large burst of hydrogen late in the transient due to massive slumping of meltdd fuel into the lower plenum which was not
!]
. representative of recoverable degraded core accidents.
To provide a more realistic time history for hydrogen release, the MARCH output-was modified by increasing the hydrogen production rate late in the time history to a constant rate which ended when hydrogen production equaled the amount produced by oxidizing'75% of the fuel cladding.
The MARCH code contains a number of modeling simplifications and assumptions which make it exces'sively conservative for predicting hydrogen release rates from BWR cores.
In order to determine more realistic hydrogen production rates for degraded core conditions in a BWR core, the Hydroger$ Control Owners Group has investigated the BWR Core Heatup Code developed for the Industry Degraded Core Working Group (IDCOR).
This code was found to be capable of accurately modeling the BWR/6 core and producing hydrogen release time histories which are more characteristic for recoverable degraded core accidents.
1 i
The BWR Core Heatup Code does not predict recoverable accidents 32 for sequences which lead to oxidation of large amounts of fuel I
cladding.
Deterministic hydrogen release rates must be
]
integrated with mechanistic predictions in order to achieve an 1
integrated-hydrogen production equal to oxidation of 75% of the
]
active fuel cladding.
]
.?
4-59 2/28/85 t
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1 J*
3/4 uncovered and core heatup is assumed to start approximately 2000 seconds after scram.
Any changes in accident sequence assumptions from Subtask 7.14 or from emergency procedure development in Subtask 13.10 which will affect the hydrogen release rates will be reflected in the final release histories.
l Responsibility - HCOG j
Status
- Complete f
7.8 Calculate Degraded Core Hydrogen Release Histories Using the modified BWR Core Heatup Code produced by Subtask 7. 6 and the accident scenario assumptions defined in Subtask 7.7, calculations of various hydrogen release rates were performed.
These analyses investigated the effect of various reflood flow rates and reflood initiation time. This information was provided to Subtask 9.7 to aid in generation of a draft test matrix.
9
]2 Responsibility - EPRI Status
- Complete i
_7. 9 Complete BWR Core Heatup Code Sensitivity Study The sensitivity of the BWR Heatup Code to input parameter variation and modeling assumptions has been assessed.
The change in hydrogen release rates due to variation of reflood flow rates, reflood initiation timing, vessel pressure, initial 1,
core water level, core wide radiant heat transfer modeling, core nodalization and fuel-clad gap conductance was determined.
The
]2 sensitivity of code predictions to timing for depressurizing the
]
reactor vessel will be evaluated.
The effect of varying the
]
amount of zircaloy inventory melt considered to be recoverable
]
j will be evaluted.
Finally, the initial temperature assumed in
]
the reactor vessel for the cladding will be varied.
These
]
4-64 2/28/85
.a
._.,n.,.._
~ > " -
- ~~~
' ' * * * ~ ~ ~ ~ ' ~ ' ~ '
~
~~
~~
sensitivity runs should demonstrate that the controlling
]2 parameters for peak hydrogen generation rate, duration of hydrogen production, and total hydrogen production are the
]2 reflood injection rate and reflood initiation timing.
Resolution of questions in Subtask 7.12 will be considered in the sensitivity studies.
Responsibility - EPRI Status
- In Progress
]2 7.10 Submit BWR Core Heatup Code Details to NRC The Nuclear Regulatory Commission staff requested additional details on the BWR Core Heatup Code prior to a review meeting on the use of the code.
The Hydrogen Control Owners Group has submitted the " User's-Manual and Details of Modeling for the BWR Heatup Code" to the Nuclear Regulatory Commission staff for review.
This manual was submitted to the sta'ff by letter HGN-020 dated September 5, 1984.
The manual discusses assumptions used in the code, equations solved by the code, required input, available output, and solution schemes employed.
Responsibility - HCOG Status
- Complete 7.11 HCOG/NRC Meeting to Review Code Application
?
j The Hydrogen Control Owners Group has reviewed the modifications to the BWR Heatup Code implemented in Subtask 7.6 and the input assumptions for recoverable accidents with the Nuclear Regulatory Commisssion staff.
The results of sensitivity studies performed with the BWR Core Heatup Code version proposed-for generation of final hydrogen release histories for input y
into the 1/4 scale testing program have also been reviewed with 4-65 2/28/85
....-~......
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._....a._
.m ~
I 4
l the staff.
In addition, this meeting permitted the Nuclear Regulatory Commission staff to identify their concerns with the BWR Core Heatup Code modeling.
This meeting was held on October 3 and 4, 1984.
~
Responsibility - HCOG Status
- Complete 7.12 Resolve Questions on BWR Heatup Code The Nuclear Regulatory Commission staff review of the modeling and input assumptions implemented in the BWR Heatup Code and the meeting conducted as part of Subtask 7.8 has generated questions concerning the use of the code.
Questions were identified relating to steam in the bypass region of the core, decay energy
+
ef fect of initial core water level, effect of the variation in fuel-clad gap conductance and oxidation cutoff temperatures.
Clarifications or changes to achieve resolution should be completed before final hydrogen release histories are selected
-for the 1/4 scale test facility in Subtask 7.17.
Any changes or clarifications will be input to Subtasks 7.6 and 7.9 and their effect determined.
Responsibility - HCOG i
Status
- In Progress
?
7.13 HCOG/NRC Meeting to Review Accident Sequences
- 1 4
j The hydrogen release histories which will be used in the 1/4 scale test program must be specified.
In order to accomplish this task, the accident sequences which will be evaluated for the containment and containment systems must be determined.
5 This involves defining the timing for system depressurization, I
timing for reflood initiation, systems available to provide 1
1 4-66 2/28/85 l
e
- I
-.....,.... - u n c
.... ~.. ~ - - - - -
. ~. " - ~ - -. - - ~. ~. - -
- ~ " " * ' ' ' ' - * ~
reflood, timing for use of containment coolers or containment sprays, and effect of plant emergency procedures.
The assumptions for these accident sequences must be consistent with the conditions that lead to hydrogen production and recoverable degraded core conditions.
The Hydrogen Control Owners Group reviewed the proposed accident sequences with the Nuclear
]2 Regulatory Commission staff.
The reflood flow rates which
]
1 t
should be considered in defining the input to the 1/4 scale test
]
program were also discussed. The meeting was held on January
]
I 30, 1985.
]
Responsibility - HCOG Status
- Completed 12 7.14 Resolve Questions on Accident Sequences It is anticipated that the Nuclear Regulatory Commission staff's review of the accident sequences proposed for use in the 1/4 scale test facility could generate questions regarding the assumptions and reasoning which supports the sequences.
Any changes or clarifications must be complete before use of the hydrogen-release histories in the 1/4 scale test program.
Any effect on Subtask 7.7 will be addressed to determine if any assumptions used to determine hydrogen generation release histories are affected.
Responsibility - HCOG Status
- In Progress
]2 1
7.15 Calculate 75% MWR Hydrogen Release History
]2 A hydrogen release history which results in total hydrogeaL_ ]
production equivalent to oxidizing 75% of the active core
]
zircaloy cladding will be calculated.
This release history will
]
4-67 2/28/85 b
...,...w_.
. ~.... ~ _ -...... -. -.. =.
~....
... ~..., -.........-.
l be calculated as mechanistically as possible given the existing
]2 limits on ability to mechanistically predict large amounts of
]
i hydrogen in a Boiling Water Reactor core while maintaining the
]
core in a recoverable geometry.
The hydrogen release history
]
will also attempt to provide some representation of the hydrogen
]
production which might occur in a BWR core analogous to the
]
hydrogen production on which the rule is based.
This hydrogen
]
a i
release history will be used in any additional containment
]
response analysis under subtask 8.9 and in analysis of the
]
1 drywell response to degraded core accidents in subtasks 10.8 and
]
10.9.
]
Responsibility - HCOG
]
Status
- In Progress
]
PrepareHydrogenhteleaseHistoryReport l
7.16 1
A report detailing the methodology and assumptions used to j
generate hydrogen release histories using the BWR Core Heatup Code will be prepared.
The BWR Core Heatup Code sensitivity runs will be included in this report along with a discussion of the effect on code output from varying initial parameters.
An open question identified at subtask 7.13 related to the basis 12 for the irreversible oxidation cutoff used to terminate zircaloy
]
oxidation in a given node.
The HCOG committed to evaluate the
]
a buildup of an oxide layer on the cladding during the oxidation
]
5 transient.
As long as this oxidation layer remains thin, an
]
I irreversible oxidation cutoff temperature of 2400 F should be
]
0 acceptable.
HCOG's evaluation of oxide layer buildup will be
]
included in this report.
This report will be submitted to the
]
Nuclear Regulatory Commission staff in Subtask 7.17 to documenL ]
h 4-67a 2/28/85 d
- ---: ' r.
,_.. _...r,......
.. - ~ -.
~ - - - - - -
u -f the hydrogen release histories used in the 1/4 scale test facility.
Responsibility - HCOG Status
- In Progress
]2 1
7.17 Submit Hydrogen Release History Report to NRC i
The report prepared as part of Subtask 7.16 will be submitted to 12 the Nuclear Regulatory Commission staff to document the assumptions and modeling used to generate the hydrogen release histories used in the 1/4 scale test program.
This submittal will include a description of input parameters and assumptions used in the generic analysis.
The report will provide final documentation for the sensitivity of code results to variation in input assumptions.
i a
Responsibility - HCOG Status
- Not Started 7.18 Select Hydrogen Release Histories For Input To 1/4 Scale Testinq I!
Based on the work completed for Subtasks 7.1 through 7.10 preliminary selection of hydrogen release rates for input to the 1/4 scale test facility has been completed.
As a result of the completion of Subtasks 7.11 through 7.14, information necessary l
to either confirm the previously selected hydrogen release rates or to select final hydrogen release time histories will be available.
This process will insure the individual members ci the Hydrogen Control Owners Group and the Nuclear Regulatory Commission staff understand the basis for hydrogen releas e j
histories which will be used in the 1/4 scale test facility.
]
4-68 2/28/85
'8
.... ~. -..
..-.m..=.
. - ~~ <
t t
a One hydrogen release history used in the 1/4 scale test facility 12 i
shall represent the most probable hydrogen generation event.
]
I This event entails loss of core inventory, failure of all makeup
]
systems, core heatup, hydrogen production and terminatibn of the
]
event by recovery of a large flow ECCS.
Another hydrogen
]
release history used in the 1/4 scale facility shall be a
]
I release history which results in a limiting diffusion flame
]
l thermal environment.
This release history must be produced by a
]
-l system which would be available during a degraded core event.
]
The event shall produce sustained diffusion flames.
A third
]
hydrogen release history shall be used to validate the
]
containment response analysis code selected under task 5.
]
1 Responsibility - HCOG Status
- In Progress 12 2
7.19 Provide Basis For Selection To NRC The basis for selecting final hydrogen release histories will be provided to the Nuclear Regulatory Commission staff before scoping tests begin.
Previous information exchanges in Subtasks 7.11 through 7.14 should have resolved most Nuclear Regulatory Commission staff questions and this submittal is to confirm the Hydrogen Control Owners Group plans just prior to initiation of Scoping Tests in Subtask 9.16.
Responsibility - HCOG e
Status
- In Progress
]2 I
i em 4-69 2/28/85
\\
j
.)
. _.... ~. -.. +,... - -. ~... ~...
-~
.c......s 7.20 Resolve Questions on Selection Basis It is anticipated that the Nuclear Regulatory Commissio' n, staff review of the final hydrogen release histories may generate some additional questions.
Resolution of any questions which would affect the Scoping Test Matrix or Subtask 7.14 will be given first priority and all remaining questions will be resolved 1.;
before production testing begins.
This will assure Nuclear Regulatory Commission staff agreement on the accident sequences considered, test matrix and hydrogen release histories before major testing is started.
Responsibility - HCOG S ta.tus
- In Progress
]2 b
e 4-70 2/28/85 J
.._ _-..:.. c. -.
~ ~ - - ~ ~ ~ * ' ~ " ~ ~ ~ ~
that CLASIX-3 provides consistent results.
Responsibility - OPS Status
- Complete j
8.6 Submit CLASIX-3 Report to NRC A report summarizing the sensitivity studies completed with the i
CLASIX-3 computer code was prepared.
This report detailed the input to the CLASIX-3 computer code for each case and contained I
the output from each sensitivity run including pressures, 3 -
temperatures, and gas concentrations for the drywell, wetwell and containment.
General conclusions regarding the effect of 4
individual parameter changes were identified.
This report was 1
submitted to the NRC by HCOG letter HGN-001 dated January 2, 1982.
Responsibility - HCOG Status
- Complete 8.7 Resol're Questions on Sensitivity Studies The Nuclear Regulatory Commission staff has requested additional information to document the use of CLASIX-3 and the selection of hydrogen burn parameters used in the base case analyses submitted by the Hydrogen Control Owners Group.
The Nuclear Regulatory Commission staff Requests for Additional Information
]2 (RAIs) were intially answered by the Hydrogen Control Owners Group's submittal HGN-011 on May 11, 1983.
Additional questions
]2 exist on the effect of several parameters on temperature
]
predicted by CLASIX-3 in the wetwell.
Parameters of concern
]
include spray carryover fractions, assumed beam lengths, heat - ]
y
. transfer effectiveness of sheet flow, use of mass mean spray
]
droplet size and characteristic length used to determine burn
]
time.
Questions on the heat transfer methods used in CLASIX-3, J
4-78 2/28/85 W4
~
_._.L........._.....i~~~.-.,_
... _., ~ _ _
~. I the ef fects of sensible heat loss from the RPV, and the effects
]2 of bypass leakage which were submitted to each utility member of HCOG are being addressed generically by HCOG.
In addition, testing results from Subtask 6.21 has provided informat. ion to answer portions of the NRC concerns.
)
Responsibility - HCOG Status
- In Progress 1
j 8.8 Determine If Additional Generic Analysis Is Required
?
t i
In response to Nuclear Regulatory Commission staff comments, CLASIX-3 has been modified to include NUREG-0588 heat transfer correlations as detailed in Subtask 10.1.
Also the Nuclear Regulatory Commission staff has questioned other parameters and assumptions as delinea_ted in subtask.8.7.
The results from new
]2 analyses from Subtask 10.9 may also provide results which affect previous work.
Spray initiation timing and the effect of other operator actions evaluated in Subtask 13.10 as part of the emergency procedures may.also necessitate completion of new generic analyses.
In addition, the HCOG has agreed to review deflagration data from burns conducted at the Nevada Test Site as part of Subtask 14.5.
These tests may provide a basis for revising some of the assumptions in the CLASIX-3 analyses.
Other questions concerning the hydrogen release histories could affect results of the previous CLASIX-3 base case and sensitivity runs.
Responsibility - HCOG Status
- Not Started 8.9 Define Additional Analyses l
.?
If additional analyses are determined to be required at decision point 8.8, then the extent of these analyses will be identified.
4-79 2/28/85 A
..m
.. ~.....
.. ~.,,.,
- -. ~...
Sensitivity runs which exercise the parameter (s) to be investigated will also be identified.
For any new analyses
]2 which are completed, the hydrogen and steam source terms used in
]
the CLASIX-3 analyses will correspond to the hydrogen release
]
history calculated at subtask 7.15.
Any additional analysis
]
which is completed will be performed with the version of the j
CLASIX-3 code that has been modified to include NUREG-0588 heat
.'D t transfer methods.
The results of any new analyses will be input
]
into Subtask 11.8 to generate the containment deflagration
]
thermal environment.
n i
Responsibility - HCOG Status
- Not Started 8.10 Plant Specific Runs Required Each utility will determine if a plant specific hydrogen deflagration analysis is required.
This will invo'lve making a comparison of plant specific systems, containment geometries, release rates and other plant specific technical assumptions with those used in the generic runs.
If the utility determines a new base case is required, this work will be completed in Subtask 8.12.
If no additional runs are determined to be necessary, then this position will be justified in subtask 8.13.
3 Responsibility - Utility j
Status
- In Progress 8.11 Establish Common CLASIX-3 Assumptions i
HCOG will prepare a list of common assumptions for use by utilities completing plant specific hydrogen deflagration runs--
,9 This will insure that parameters which have been investigated by HCOG are treated in a conservative and consistent manner in analyses performed by the individual utilities.
Each utility 4-80 2/28/85 ia
- L., ~. - ~
~
. iJ.T
_.n..-,.,..
1 l
anticipated the Nuclear Regulatory Commission staff would request additional information on the f acility's goals, design and use.
Several Requests for A,dditional Information (RAI) were received by HCOG on December 8, 1983 and responses were provided in letter HGN-016 dated April 2, 1984.
In addition, other
- j questions relating to the facility have been addressed to the 1
Hydrogen Control Owners Group during the meeting identified in
'1 Subtask 9.4.
The HCOG is committed to evaluate the comparitive 12
'i heat losses between the test facility and the full scale Mark
]
l III plants.
The purpose of this study is to assure that the gas
]
l temperatures measured in the 1/4 scale facility will be
]
conservative or comparable to the temperatures which would be
]
produced in the full scale plant.
The results from this study
]
will be documented to the NRC.
This activity reflects part of
]
the generic efforts to answer these questions and provide responses to the Nuclear Regulatory Commission.
The development of responses to the Nuclear Regulatory Commission staff has had impact on the facility goals, design and use.
Therefore, feedback from this activity to program elements 9.1 and 9.2 is shown.
Responsibility - HCOG Status
- In Progress t
9.6 Construct Test Facility The 1/4 scale test facility is being constructed for HCOG by the Electric Power Research Institute (EPRI) and Factory Mutual Research Corporation (FMRC).
Initial construction was started in August, 1983 and was completed in January, 1985.
The 1/4 12 Scale Test Facility is located at FMRC's remote test site E West Gloucester, Rhode Island.
Planning and construction
]2 4-91 2/28/85 J
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.m
__..m.
m
)
control was provided by FMRC with EPRI retaining overall
]2 management and budget authority.
Responsibility - EPRI/FMRC Status
- Complete
]2 i
i 9.7 Draft Test Matrix
{
' ]
The 1/4 scale testing will be divided into shakedown, scoping l
and production tests.
The shakedown tests are intended to verify proper operation of all systems and instruments in the I
test facility.
Scoping tests will evaluate the effect of a number of' parameters which might affect the production tests.
Parameters which will be evaluated during scoping tests include test repeatability, the effect of simultaneous hydrogen and steam discharge, the 4ffect of release through both the LOCA vents and the SRV spargers, effect of the presence of grating near the suppression pool surface, the effect of' containment cooling system dif ferences, the effect of geometry differences
]2 between two HCOG member plants, and the effect of changing the
]
blockage fraction in the facility.
The production tests will be
]
{
used to define the full scale thermal environment produced by diffusion flames in the wetwell.
The production tests will include variation on the location of the sparger which is assumed to be open, a test without containment sprays in i
operation and two tests with a hydrogen release rate below the threshold for steady diffusion flames.
Production tests will be completed for at least three containment geometries.
The final production test matrix will be dependent upon the outcome of the scoping tests.
The hydrogen release rates used for the various tests are provided by Subtask 7.8.
I This draft test matrix is complete and has been discussed with y
the Nuclear Regulatory Commission Staff.
However, it is 4-92 2/28/85 4
,....,,.n..-..--.....-,.
- ~. - - - - - ~ ~ ~ -
~ ~
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.. ~...
anticipated that resolution of Nuclear Regulatory Commission I
questions in Subtask 9.9 may affect the final test matrix.
Responsibility - HCOG Status
- Complete i
9.8 Submit Test Matrix To NRC 3;
i l
The draf t test matrix was submitted to the Nuclear Regulatory Commission by HGN-018 dated July 6,1984.
The purpose of this submittal was to provide the Nuclear Regulatory Commission staff l
with the technical basis for the draft test matrix.
This matrix was reviewed in a meeting between HCOG and the Nuclear 4
Regulatory Commission staff on August 28 and 29, 1984.
It is 3
anticipated that questions on the draft test matrix will be identified by the NRC 's'taff.
Resolu' tion of questions will be completed as part of subtask 9.9.
Responsiblity - HCOG Status
- Complete 9.9 Resolve Questions on Test Matrix i
i i
i Resolution of Nuclear Regulatory Commission staff questions on
{
the draft test matrix assumptions and methodology must be complete before the beginning of scoping tests in the 1/4 scale facility.
This is to insure agreement on the important parameters which should be established by the various scoping and production tests.
This will establish test conditions which will yield realistic diffusion flame thermal environments.
l Resolution of Nuclear Regulatory Commission staff questions will provide input to Subtask 9.7.
In addition, input from Subtask I
4-93 2/28/85 f
I
...,..~....
.~..~.__~.-.
.~.- - - -
. - ~ - - - + -
- - > ~ ~ ~ ~ - ~ ~ -
~ - - -
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l 13.10 may provide effects from the EPG review which will affect 1
the draft test matrix.
Responsibility - HCOG Status
- In Progress l
l 1
9.10 Develon Test Procedures
..)
Test procedures which define the logical progression of each test and the use of the 1/4 scale test facility will be generated.
These procedures will specify the values of appropriate parameters, test conditions, and test acceptance criteria so that a set of data sufficient to accomplish the individual test goals will be assured before the facility l
configuration is significantly altered. The procedures will assure consistency of the test data and that the facility is properly aligned before, during and after test'ing.
The procedures will also insure the safety of facility personnel is 4
maintained by identifying personnel precautions and limitations.
- l subtask 13.9 will provide input on the use of containment sprays.
t Responsiblity - EPRI/FMRC Status
- In Progress i,
3 9.11 HCOG/NRC Final Facility Walkdown i !
- l After construction is essentially complete, the Hydrogen Control Owners Group offered the Nuclear Regulatory Commission staff the 12 I
opportunity to conduct a final facility walkdown and review I
instrument locations.
This walkdown and review allowed the
]2 utility sponsors and the Nuclear Regulatory Commission staff tT
.?
inspect the finished facility to assure it is designed and i
constructed as previously described.
Also the facility's 4-94 2/28/85 i
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r i
ability to exercise various parameters (such as hydrogen release rate) was reviewed to assure the test matrix can be implemented.
]2 The detailed instrumentation location was reviewed.
The meeting
]
was held on December 18, 1984,
)
i Jj Responsibility - HCOG
]
Status
- Complete
]2 ii 9.12 Complete Shakedown Testing Af ter construction of the test facility is complete, shakedown testing will be started to initially calibrate, test and place various mechanical, electrical, and instrumentation systems in the facility in working order.
This will assure that the
{
facility will respond as planned to the test sequence and that instrumentation required to record the facility response is in working order.
Any instrumentation or other equipment which does not respond as planned will be repaired, replaced or shown to be unnecessary for testing to begin.
Integrated tests will j
be completed to assure that all of the facility systems j
including containment sprays, instruments, and hydrogen injection system can function in an, integrated manner. Any necessary changes during shakedown will be documented and explained to the Hydrogen Control Owners Group.
j Responsibility - EPRI/FMRC Status In Progress
]2 9.13 Identify Any Facility Changes to NRC Any necessary facility changes or problems found durinL shakedown in subtask 9.12 which requires the facility to be substantially modified or the test matrix to be changed will be discussed with the Nuclear Regulatory Commission staff.
The 4-95 2/28/85 l
4
.~y.....
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start of Scoping Tes:s, Subtask 9.17, will not be delayed for this activity.
Responsibility - HCOG
~
Status
- In Progress i
9.14 Prepare Final Design Report To document the 1/4 scale facility design, (which has been modified since the submittal of the draft design report) a final 1/4 Scale Test Facility design report will be prepared.
Significant changes in the facility geometry, containment cooling systems modeled and instrumentation will be incorporated into the final design report.
The role of calorimeters provided by Sandia National Laboratory and a complex geometry calorimeter provided by the Hyd'rogen Control' Owners Group will also be discussed in the final document. This activity will.be completed concurrently with other Subtasks.
The start of scoping tests will not be delayed for this activity.
Responsibilty - HCOG Status
- Not Started 9.15 Submit Final Design Report to NRC Staff The report documenting the design and configuration of the f acility as used for testing will be submitted by the Hydrogen Control Owners Group to the Nuclear Regulatory Commission staff.
The Nuclear Regulatory Commisssion staff has been apprised of all facility changes by various HCOG/NRC meetings.
This Subtask will provide a reference for facility design features.
7 Responsibility - HCOG Status
- Not Started I
4-96 2/28/85 4
. ~.
~- --"*--- ~
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9.16 Finalize Scoping Test Matrix Prior to the beginning of scoping tests, all questions relating to the effects which could alter the production test matrix must be identified.
The hydrogen release history which will be used for the repeatability tests and all other scoping tests must be finalized.
Hydrogen release histories finalized from Program Task Element 7.17 may have a late input on the scoping test j
matrix.
Completion of this element will be established when i
Program Task Elements 9.9 and 7.17 are complete.
i Responsibility - HCOG Status
- Not Started 9.17 Complete Scoping Tests There are currently 15 scoping tests planned to confirm the effect'of important parameters which affect the definition of the diffusion flame thermal environment.
The following parameters will be addressed by scoping tests:
- Data repeatability
- Threshold for establishing continuous diffusion flames
- Effect of concurrent steam and hydrogen injection
- Simultanecus discharge through LOCA vents and spargers
- Effect of grating near suppression pool surface on the 4
combustion v.ransients
- Effect of changing the blockage fraction in the
]2 j
facility
]
- I
- GGNS & PNPP geometry similarities A complete set of test data will be recorded for each scoping test.
~
Responsibility - FMRC 4
Status
- Not Started 4-97 2/28/85 J
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i 1
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subtasks 12.6,12.8,12.9 and 12.10 to aid in validation of the
]2 analytical methods used in the survivability analysis program.
l Responsibility - EPRI/FMRC t
Status
- Not Started 1
j 9.24 Submit Preliminary Scoping Test Results The scoping test data will be organized and correlated into a scoping test report and submitted to the Nuclear Regulatory Commission.
This will serve to document the Hydrogen Control Owners Group's conclusions regarding the parameters investigated during scoping tests.
Any changes to the production test matrix to reflect the observed behavior of the facility will also be addressed.
Final scoping test data evaluations will be included in the Final Test RepoYt prepared und'r Subtask 9.28.
e Responsibility - HCOG Status
- Not Started 9.25 Resolve Questions on Scoping Test Results It is anticipated that the Nuclear Regulatory Commission staff's J
detailed review of the scoping test data report may generate i
questions which could affect the production test matrix.
Feedback to Subtask 9.19 to reflect any late changes in the production test matrix is considered a part of this task element.
The Hydrogen Control Owners Group will also generate 1
-j responses to any Nuclear Regulatory Commission staff questions regarding the scoping tests as part of this activity.
Responsibility - HCOG
~
J Status
- Not Started i
1 4-1 01 2/28/85
,i
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t transfer to the containment sprays.
The sprays shall be demonstrated to produce bulk atmosphere mixing patterns which are representative of the bulk atmosphere mixing patterns expected at full scale.
The total heat transfer characteristics of the scaled test 11 I
facility shall be conservative with respect to the heat transfer l
characteristics of all full scale Mark III containment or i
representative of the actual heat transfer characteristics.
It shall be demonstrated that the heat losses to the gratings, 11 walls and other heat sinks in the scaled test facility shall not
]
exceed the heat losses to gratings, walls, equipment and other
]
3 heat sinks 'in the full scale facility.
]2
]
The scaled test facility must have the capability to simulate variable hydrogen a-nd steam flow into the facility.
The hydrogen flow rate into the facility must be variable over the the range of expected hydrogen production rates.
~ he facility 12 T
shall have the capability to simulate steam flows associated
]
with the hydrogen release history which produces the limiting
]
diffusion flame thermal environment throughout the hydrogen
]
generation event.
The hydrogen shall be injected into the facility at locations which correspond to the points of hydrogen release in the full scale plant.
4 4.
The repeatability of the test data shall be evaluated.
A 11 set of tests with comparable initial conditions and identical
]
geometry shall be completed.
Acceptable repeatability for
]
s i
the test data will be determined by comparing the gas
]
t i
temperatures, velocities, and radiant heat fluxes.
g 4-105 2/28/85 i
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number of different locations in the wetwell and containment.
The mixing data will be evaluated to demonstrate that mixing of the containment gases prevents significant accumulation of hydrogen.
Hydrogen concentrations shall be shown to not. exceed 8 v/o at elevations above the first row of igniters and that 11 detonatable mixtures do not exist.
]
Y 9.
Data shall be obtained from the test facility to validate
-[
analytical methods which are used by HCOG.
An instrumented calorimeter with complex geometry shall be included in the i
facility.
This calorimeter shall be used to validate the heat transfer methods used in calculating the temperature response of the component.
Data shall also be obtained to validate the methodology used to
]2 define the thermal environment for evaluating equipment
]
survivability.
Convective heat flux shall be determined and
]
i compared with values calculated using measured gas velocities
]
I and temperatures.
Radiative heat flux shall be measured and
]
compared with values calculated using standard text book
]
methods.
]
i a
0 i
4-107a 2/28/85
>- -~-
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TASK 10 - EVALUATION OF DRYWELL RESPONSE TO DEGRADED CORE ACCIDENTS A drywell line break which leads to degraded core' accident will result in hydrogen being introduced into the drywell.
Current accident sequences indicate very little oxygen will be 1
present at the time of hydrogen release due to prior steam purging of the drywell atmosphere into the containment through the LOCA vents.
As core recovery progresses, the introduction of oxygen into the drywell can occur through the drywell vacuum breakers, operation of the drywell hydrogen mixing system, or
]2 reverse vent clearing.
The re-introduction of oxygen could
]
result in either a standing flame at the point of entry known as an inverted diffusion flame or will mix with the drywell atmosphere until ignition of a deflagration occurs.
Both scenarios will add hydrogen / oxygen re-combination energy to the drywell environment.
This task will evaluate the drywell thermal environment before hydrogen combustion occurs and determine the additional effects of hydrogen combustion in the drywell.
i 1
4-108 2/28/85
'i --
..a 10.7 Define ADS Timing Flow through a postulated drywell break will continue until the reactor is depressurized by the operator using the IDS The time that ADS actuation will occur will depend on the operator's response to action limits in the emergency procedures based on drywell temperature, reactcr vessel level, suppression pool heat capacity or other limits in the emergency procedures.
The i
i operator is expected to maximize the time available before hydrogen production using steam cooling.
Before hydrogen o
production commences, the vessel will be fully depressurized.
This establishes the point where hydrogen and steam production from the degraded core can be predicted by the BWR Core Heatup I
Code.
The hydrogen and steam flow will be split between the i
open ADS SRV's and the drywell break after the vessel is depressurized using the ADS.
Responsibility - HCOG Status
- Not Started 10.8 Calculate Drywell Break-SRV Flow Split The division of steam and hydrogen between the drywell break and the open SRV's will be affected by the drywell pressure history after ADS, the number of ADS valves open, suppression pool level, the reactor vessel pressure, and the timing of ADS.
Based on these factors, a realistic flow split of hydrogen and steam will be determined.
Using the steam and hydrogen release history generated as part of subtask 7.15, a blowdown history of
]2 i
steam and hydrogen through the break into the drywell will be calculated.
This blowdown history of steam and hydrogen and the pre-ADS steam blowdown history from either Subtask 10.4 or 10W l
8 will be used to define the steam and hydrogen release history to the drywell from the time of break until core recovery is 4-112 2/28/85
.J
._ c
..c _...
j.
4 4
completed.
This composite release history will be used as input i
to analyze the drywell thermal response in Subtask 10.9.
l Responsibility - HCOG Status
- Not Started i
h 10.9 Analyze DrYwell/ Containment Pressure Using CLASIX-3 The CLASIX-3 program is a multi-compartment containment response analysis code which will predict temperature, pressures and concentrations of gases.
To represent a drywell break, a time history of.the steam, hydrogen and fission product energy along i
with associated enthalpies must be input into the code.
This is i
L provided by Subtasks 7.15 and 10.8 for this analysis. Emergency
]2 procedure actions determined from Subtask 13.10 will also be To determine if deflagrations in the used for this analysis.
4 drywell are possible, the CLASIX-3 code or other acceptable i
deflagration analysis code will be run for the drywe'll using the blowdown history from Task 10.8 and appropriate combustion parameters for the drywell.
The analysis will account for the effects of drywell bypass leakage on the wetwell and upper
]'
containment response.
If deflagrations occur then the effect of this drywell pressure spike will be determined.
This analysis will also allow the Hydrogen Control Owners Group to determina j
if inverted diffusion flames can be established at the exit of the CGCS compressor or from other oxygen sources in the drywell.
The results from this analysis may define the environment which l
equipment in the drywell must survive if the Hydrogen Control Owners Group concludes that inverted diffusion flames do not occur in the drywell.
Thus, this subtask provides input to Subtask 11.6 in the equipment survivability analysis program.
The resolution of Nuclear Regulatory Commission staff questionT from Subtask 10.11 concerning this analysis will be assessed to determine their impact on this subtask.
Since previous CLASIX-3 4-113 2/28/85 1
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r i
i ACCEPTANCE CRITERIA FOR TASK 10 EVALUATION OF DRYWELL RESPONSE TO DEGRADED CORE ACCIDENTS 11 1.
Accident scenarios which introduce hydrogen into the drywell shall be described. Scenarios will be based on line breaks that
'I are consistent with a recoverable degraded core.
The largest i
j break considered shall be limited to a size equivalent with the i
throat of a single safety relief valve.
i, i
2.
Vessel blowdown and drywell response prior to vessel 4
q depressurization shall be predicted with a recognized analysis code.
Realistic assumptions shall be used in calculating the drywell's response to vessel blowdown.
3.
Vessel blowdown and core heatup following depressurization of the reactor coolant'~ system will be' predicted with a degraded core analysis meeting the acceptance criteria for Task 7.
]1 Vessel blowdown to the drywell shall include the period of
]
j recovery from the degraded condition.
i t
4.
The drywell response shall be calculated using an analysis code meeting the acceptance criteria for Task 5.
Parameter studies shall be completed to determine variations in plant t'
unique features such as the hydrogen mixing system or vacuum 5
breakers.
1 I&
5.
The potential for existence of combustion phenomena unique 3
to the drywell shall be evaluated. Criteria for the existence of
,. i inverted dif fusion flames in the drywell shall be established.
These criteria shall include definition of oxygen inflow rates,
]1 bulk compartment hydrogen concentration, and air inlet nozzle
]
I geometries required to sustain an inverted diffusion flame. IT
]
these criteria are satisfied, the effect of inverted diffusion flames on the drywell environment shall be defined using l
l 4-124 2/14/85 i
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- la
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existing experimental data and analytical techniques or from a suitable test. Drywell essential equipment exposed to a i
potential inverted diffusion flame environment will be shown to 4
meet the acceptance criteria of Task 11.
1 1',
j 6.
The pool swell transient shall be defined based upon 6
lj expected combustion in the drywell.
Drywell and containment 9
.4-l structures and components shall be evaluated to determine that
.Y pool swell does not impose structure, equipment or support 1
.~ s s
loadings greater than previously analyzed.
This may be i
j ;
accomplished by demonstrating that pool swell loads do not exist i
or that pool swell loads are enveloped by the present design loads, or that essential structures and components survive the I
pool swell event. The LOCA design basis drywell to containment
]1 pressure differential will be compared to the differential
]
].
pressure transient pr'6duced by hydrogen combustion.
No pool
]
swell loadings will be evaluated if the drywell to containment
]
differential pressure for a design basis event exceeds the
]
5 hydrogen combustion differential pressure for the length of the
]
i transient.
]
I e
j i
i t
i 1
4 4
J
.t" II 4-125 2/14/85 l
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- i 11.1 Criteria For Equipment Survivability The criteria used to determine if a piece of essential equipment in the containment or drywell survives a degraded core' hydrogen combustion event are based on the current equipment qualification program conducted in accordance with NUREG-0588.
l
,l If the pressure spike from a hydrogen deflagration is below the static qualification pressure, the equipment is deemed
)
acceptable.
If the surface temperature response of a component a
to hydrogen combustion is above the qualification temperature 4
,2 and the temperature rise of the critical component is below its
=,
qualification temperature, the equipment is expected to survive.
t.
i If the temperature qualification envelope does not bound the temperature rise of the critical component due to hydrogen combustion, other measures must be applied as determined in subtask 11.18 to assure survivability.
l Responsibility - HCOG 4
Status
- Complete i
11.2 Identify Equipment Required To Survive A generic list of equipment required to survive a degraded core accident resulting in hydrogen production has been prepared by i
the Hydrogen Control Owners Group.
Five criteria were used in
]2 the selection of this equipment:
(1)
Systems and components required to maintain the core
]2 l
in a safe shutdown condition 3
]
(2)
Equipment and structures required to maintain the
]
integrity of the containment pressure boundary
]
I (3)
Equipment and systems which must function to mitigate
]
the consequences of the event
]
y (4)
Instrumentation and systems which will be used to
]
l monitor the course of the event and provide guidance
]
i 4-128 2/28/85
. _. n
,...-,....-.-....~s
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~ ~ ~ - ~ ~ ~ ~ ~ " ~ ~ - ~ ~ ~ ' ~ ~ ~ ~
~-
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to the operator for initiating actions in accordance
]2 with the Emergency Procedures Guidelines
]
(5)
Components whose failure could preclude the ability,of
]
the above systems to fulfill their intended fbnction
]
~i The generic equipment list has been given to each Hydrogen
,7 Control owners Group member utility for guidance in developing
.k plant specific equipment lists.
a Responsibility - HCOG Status
- Complete 11.3 Select Survivability Analysis Code A computer code capable of analyzing the thermal response of a piece of equipment subjected to transient heat fluxes due to thermal radiation, convective heat transfer and conductive heat transfer will be selected.
The code will be capable of solving l
complex geometries in various coordinate systems.
The ability to input the variable heat flux at component boundaries due to i j the dynamic thermal environment will be required.
Responsibility - HCOG Status
- In Progress 11.4 Establish Components To Be Analyzed 1
1
. jl Once plant specific lists of equipment required to survive these
}
transients are established, the lists will be reviewed to identify common components.
In addition, the physical geometry of similar components will be reviewed to determine if a single heat transfer model can be used to represent a variety o4--
j similar components.
Finally, if two similar pieces of equipment are included on the survivability lists of HCOG utilities, it will be determined if the more thermally limiting piece of 4
(
l 4-129 2/28/85 i
i
%,......,.........+.-~s------~~----~~-~~~~~~-~-
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i' ACCEPTANCE CRITSRIA FOR TASK 11 EQUIPMENT SURVIVABILITY ANALYSIS PROGRAM 11 1.
A list of equipment required to survive hydrogen g'entration
]
events shall be prepared for each plant.
Equipment meeting the following criteria shall be included on this list:
}
- Equipment and systems which must function to mitigate the consequences of the event i
i
- Equipment and structures required to maintain the
{
integrity of the containment pressure boundary
- Systems and components required to maintain the core in a safe shutdown condition
- Instrumentation and systems which will be used to monitor the course of the event and provide guidance to the 11 operator for initiating actions in accordance with the
]
Emergency Procedure Guidelines
]
- Components whose failure could preclude the ability of 11 the above systems to fulfill their intended function
]2 The effects of hydrogen combustion are limited to the containment and drywell.
Only equipment located in these two compartments shall be evaluated for inclusion on the survivability list.
Degraded core accidents evolve over a relatively long period of time before zircaloy oxidation begins.
Many components will have performed their safety function before hydrogen combustio r j
,9 can begin.
If these components are not required to function during or after hydrogen combustion, and if failure of the 4-139 2/14/85 g
_...~.m
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.. _.... _.. -...... ~,,.......
.. ~....... _.. -
component will not compromise plant safety or adversely affect the performance of equipment required to survive hydrogen
]1 combustion, then the component will not be required to survive
]
these accidents.
~
]
i 2.
The equipment and its internal component temperature
]1
]
responses will be calculated using an accepted heat transfer 1
code.
This code shall be capable of solving steady-state and ai transient heat conduction problems including radiant heat
!' 3 transfer in one, two and three dimensional cartesian or cylindrical coordinates.
The analysis code shall be capable of analyzing time dependent boundary conditions.
Equipment models shall be based on equipment drawings and
]1 manufacturer's data which account for the as-installed
]
~
orientation and mounting arrangement.
Models shall be
]
constructed considering the most appropriate coordinate system,
)
component materials, internal heat generation, internal volumes
]
or air spaces, and specific thermal properties of the materials
]
of construction.
Boundary conditions shall be established for
]
all conducting surfaces.
]
3.
The number of components to be modeled and/or analyzed may
]1 l
be limited if one of the following criteria is met:
A.
The identical component model has been previously analyzed with a more limiting thermal environment and i
i found to be acceptable.
'l B.
A similar, more thermally responsive component model, has been determined to provide conservative thermal response results which meet the survivability criteria.
l Components may be judged to be similiar if the thermal
]1
{
mass of two components, materials for two components,
)
4-140 2/14/85
-i
l__
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or overall geometry for two components can be shown to be comparable or conservative.
4.
The thermally limiting component shall be a component
]1 or subassembly of a piece of equipment required to survive
]
f.
i hydrogen combustion which is determined most likely to fail 1
during a hydrogen combustion temperature transient.
5.
Thermal environments produced by deflagrations, diffusion
]1 flames and inverted diffusion flames shall be defined for the locations of equipment required to survive these transients.
The deflagration thermal environment shall be defined based on containment response analysis produced in Task 8.
The diffusion flame thermal environment shall be defined by scaling up of test
]1 data from appropriate tests in the 1/4 scale test facility and
]
meeting the acceptance criteria identified in Task 9.
The
]
inverted diffusion flame thermal profile for the drywell shall be defined based upon experimental data or analyses using the acceptance criteria identified in Task 10.
All of the thermal profiles shall be defined based on realistic experimental data or analyses.
Factors of conservatism need not be applied to the definition of the thermal environments.
6.
Equipment and components shall have demonstrated the ability
]1 to survive a hydrogen burn temperature transient if one of the following criteria is met:
1; A.
The equipment surface temperature is equal to or below the equipment qualification temperature.
B.
If the surface temperature exceeds the equipment qualification temperature, then the equipment oT component will survive a hydrogen burn if the temperature response of the most thermally limiting 4-1 41 2/14/85
~........._-...,...
i i
component is equal to or below the component qualification temperature.
C.
The equipment surface temperature is equal to 6r below the equipment survivability temperature.
The i
survivability temperature shall be defined as a j
temperature, higher than the qualification temperature, at which the equipment has been demonstrated to operate by analyses or testing.
Component qualification temperature shall consider the period of
]1 time that a component is maintained at a specific temperature.
]
7.
Equipment and components shall demonstrate the ability to 11 survive a hydrogen burn pressure transient by meeting one of the following criteria:
A.
The equipment experiences a peak pressure or 11 differential pressure, as determined from containment
]
deflagration analysis acceptable per criteria
]
identified in Task 8, below the equipment qualification
]
]
pressure.
B.
The equipment can be shown to be insensitive to pressure increases 8.
If a piece of equipment or critical component cannot be
]1 l
shown to survive, then measures shall be identified to assure survivability.
These measures may include but are not limited to:
A.
Protecting the component by use of:
1)
Shields 2)
Insulation 3)
Cooling 4-142 2/14/85
.)
..... -., ~..
1 s.
m,.,_
.~
r 4
B.
Replacing the component with equipment which will 11 survive the hydrogen burn environment.
]
C.
Relocating the component to a milder environment
-4 4
9
.o 9
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12.1 Develop CLASIX-3 Model of 1/4 Scale Test Facility A specific CLASIX-3 input case using the 1/4 scale test facility will be developed.
The modeling of the 1/4 scale test facility
]2 may be considered an extension of the code verification
]
completed under subtask 5.8.
The specific treatment of
}
geometry, compartment volumes, heat sinks, spray flow, spray carry over, inter compartment connections and other features of the 1/4 test facility must be determined in order to obtain an accurate model.
This information will be used in Subtask 12.2 to specify an appropriate input case.
Responsibility - HCOG Not Started Status 12.2 Specify CLASIX-T Input The steam and hydrogen flows, compartment initial conditions, burn parameters, flow path parameters, spray system parameters, heat sinks, and suppression pool level, must be determined to define an input case for CLASIX-3.
Where appropriate, the same assumptions used in previous containment response analyses will a
be used in the 1/4 scale test predictions.
For example, hydrogen combustion in a compartment will be initiated when bulk compartment hydrogen concentration reaches 84 and 85% of the hydrogen in the compartment will be assumed to burn.
At
]2 least one analysis will also be completed with best estimate
]
3 combustion parameters such as assuming combustion is initiated
)
I when compartment hydrogen concentration reaches 64 with 65% of
)
the hydrogen being burned.
The spray carryover fraction used in
]
the CLASIX-3 analysis will be measured in the 1/4 scale test.
]
Responsibility - HCOG Status Not Started 4-144 2/28/85
......_..._._..._.....-.__.........,s._..
w..,-,..--
)
i 1
12.3 Complete CLASIX - 3 Prediction l
Using the CLASIX-3 model of the 1/4 scale test facility developed in Subtask 12.1 and the input data file from Subtask j
12.2 with appropriate test specific flows, a test prediction of lj the 1/4 scale test facility response will be made.
This run
- ' i will predict the containment gas temperatures, constituent gas concentrations, and containment pressure response for the deflagration event in the 1/4 scale test faclity.
i e
l
'l Responsibility - HCOG Status
- Not Started 3
12.4 Design Complex Calorimeter l
A complex calorimeter will be designed and installed in the test i
facility.
The complex calorimeter will represent ~differant
~
types of equipment geometries such as rectangular and cylindrical components.
It will also have several different l
materials with a variety of coupled thermal masses.
This device will be sufficently instrumented to measure its response to the 1
I convective and radiant heat flux present in the f acility.
The ability to move the calorimeter to various locations in the 1/4 i
j scale test facility will be provided in order to measure the jj change in total heat flux as the distance from hot gas f
{
plumes is increased. Design details on the complex calorimeter,
]2 l
}
its location in the facility, and instrumentation near the
]
j calorimeter were supplied to the NRC staff in letter HGN-027 1
i i
4 t
dated February 13, 1985.
]
4 Responsibility - EPRI Status
- Complete
]2
~
)
.9 i
4-145 2/28/85 i
i J
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--~..-.. - -
z.=.= -. ~ -
- + - ~ - - - -
~ ~ ~ * ~ -
. - - - ~
12.5 Prepare Model of Complex Calorimeter k
Using the modeling instructions and format specified by the survivability analysis code established at subtask 11.3,.a heat i
transfer model of the complex calorimeter will be prepared.
This model will be used in Subtask 12.7 and 12.9 along with the l
appropriate thermal environments to predict the response of the i
complex calorimeter.
Responsibility - HCOG Status
- Not Started 1
12.6 Compare CLASIX-3 Predicted Results with Measured Results
]2 The CLASIX-3 predictions of temperatures, pressures and gas
]
l concentrations in the.wetwell and upper containment shall be
]
compared with measured test results.
The amount of combustion
]
j which is predicted to occur in each compartment will'be compared
)
with the combustion observed in test data and on video tapes.
]
An attempt will be made to relate temperatures measured in the
]
facility locally to global temperatures predicted by CLASIX-3.
]
4 Responsibility - HCOG
]
Status
- Not Started
]
i l
12.7 Apply CLASIX-3 Temperatures to Ca= plex Calorimeter Model i
Using the mathematical model of the complex calorimeter prepared
.j in subtask 12.5 and the thermal response code selected in l
Subtask 11.3, a prediction of the complex calorimeter behavior will be made.
The deflagration environment predicted for the l
1/4 scale facility by CLASIX-3 will be used to make t h ie--
i prediction.
This information will be used in Task 12.8 to i
I 4-146 2/28/85 s
i j
n..~.
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u demonstrate conservatism in the use of CLASIX-3 output and thermal response models to predict the peak temperatures in equipment and components.
Responsibility-- HCOG l
Status Not Started i
t j
12.8 Compare Measured Results with CLASIX-3/ Heat Transfer Code 4
Predictions The data from Subtask 9.23 which defines the measured response of the complex calorimeter in the 1/4 scale test facility will be compared with the predicted response of the complex calorimeter model produced by Subtask 12.7.
If the predicted response is conservative compared to the measured response of the complex calorimeter, then the analytical methodology previously used to evaluate equipment survivability for deflagrations will have been shown to be conservative.
This result will be documented in Subtask 12.13.
If the predicted response does not envelope the measured response of the complex calorimeter, then a review of modeling assumptions and techniques will be conducted to determine what revisions are necessary.
Changes in modeling assumptions or methodology will be documented in Subtask 12.13.
This information will be considered in the equipment survivability analysis completed at Subtask 11.11.
Responsibility - HCOG Status Not Started i
i 12.9 Apply Measured Diffusion Flame Environment to Complex l
Calorimeter Model
?
Using the data from Subtask 9.23, the diffusion flame thermal 4-147 2/28/85 i
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environment in the vicinity of the complex calorimeter will be determined.
This thermal environment will be used as input to the thermal response model of the complex calorimeter prepared in Subtask 12.5.
This will yield a prediction of the' response of the complex calorimeter in the diffusion flame environment of i
the 1/4 scale test facility.
This information will be used in Subtask 12.9 to validate the heat transfer modeling assumptions and techniques.
i j.
Responsibility - HCOG Status
- Not Started 12.10 Compare Measured Results With Thermal Response Predictions The data from Subtask 9.23 which defines the measured response of the complex calor-im'eter in the 1/4 scale test facility will be compared with the predicted response of the complex calorimeter model produced by the Subtask 12^. 9.
If the predicted response from the model is conservative compared to the measured response of the complex calorimeter, then validation of the techniques and assumptions used to construct 1
the.model and the assumptions used in defining the thermal environment will be achieved.
This result will be documented in the report prepared at Subtask 12.13.
If the predicted response does not envelope the measured response of the complex calorimeter then a review of modeling assumptions and techniques will be conducted to determine what revisions are necessary.
Changes in modeling assumptions or methodology will be documented in the report prepared as part of Subtask 12.13.
This information will be used in the equipment survivability analysis conducted at Subtask 11.11.
Responsibility - HCOG Status
- Not Started 4-148 2/28/85 e
-.. _..,....... -.._..,....-.....c...1
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12.11 HCOG/NRC Meeting to Discuss Methodology Validation The Hydrogen Control Owners Group will meet with the Nuclear Regulatory Commission staff to discuss the result 5 Qf the modeling validation process.
The model predictions to demonstrate that the modeling yields conservative thermal response results when compared to the measured thermal response will be discussed with the Nuclear Regulatory Commission staff.
The results from Subtasks 12.8 and 12.10 will be presented.
Responsibility - IICOG Status
- Not Scheduled i
I 12.12 Resolve Questions of Methodology Validation It is anticipated the Nuclear Regulatory Commission staff review of the modeling process and validation techniques presented in Subtask 12.11 may generate questions or requests for additional information regarding the methodology and assumptions employed.
This subtask will provide responses and resolution of the staff's questions.
Any changes in the modeling process will be documented in the report prepared as part of Subtask 12.13 and reflected in the methods used for Subtask 11.11 to analyze the thermal response of essential equipment.
Responsibility - HCOG Status
- Not Started 12.13 Prepare Final Rep 6rt on Methodology Validations A report documenting the validation of the methodology for analyzing equipment response in a diffusion flame environment--
j and demonstrating the conservatism in previous deflagration 4-149 2/28/85
- .4
...._.~..,..........:_...
..,.,..,_..m.,
thermal environment definitions will be prepared.
This report will describe the techniques and assumptions used in Subtask 12.5 to develop an analytical heat transfer model of the complex calorimeter used in the 1/4 scale test facility.
The predicted response of this model to the diffusion flame th'ermal environment produced in the facility will be described and l
compared with the actual environments measured in the test j
facility.
In addition, the measured response of the complex i
calorimeter to a low hydrogen release rate which leads to j
deflagrations will be compared to a prediction of the calorimeter response using CLASIX-3 with the analytical heat transfer model of the calorimeter.
It will be shown that CLASIX-3 and the heat transfer model of the calorimeter provide conservative predictions of the response of the calorimeter to deflagration thermal environment measured in the 1/4 scale test facility. Information from the NTS data evaluation in Subtask 14.5 will be reviewed for the preparaton of this report.
Any changes to the modeling techniques resulting from the validation process will be incorporated into the models prepared for the analyses of equipment required to survive degraded core accident in Subtask 11.11.
Responsibility - HCOG Status
- Not Started 12.14 Submit Final Report on Methodology Validation to NRC The final report documenting the results from the validation process will be submitted to the Nuclear Regulatory Commission staff.
This report will include a description of any changes made in modeling techniques to analyze equipment for Subtask 11.11.
The report will also contain responses to any NRC questions as developed at Subtask 12.12.
Based on previous interactions with the Nuclear Regulatory Commission staff, it is 4-150 2/28/85
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not. anticipated that any further questions will be identified following submittal of this report.
Responsibility - HCOG j
Status Not Started l.
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ACCEPTANCE CRITERIA FOR TASK 12 VALIDATION OF ANALYTICAL METHODS 1.
To validate the methodology used to construct" equipment thermal response models, the predicted response of a mathematical model of a complex calorimeter in known thermal 3
1 environments resulting from hydrogen dif fusion flames shall be
'i.
compared with the measured response of the actual calorimeter.
'f The methodology will be verified if the predicted response is conservative compared to the measured response.
This validation i
process shall be completed in two thermal environments with different radiative and convective contributions to the total surface heat flux.
2.
The methodology used to predict the equipment thermal response using mathematical models of the equipment and thermal environments derived from containment deflagration response predictions from CLASIX-3 analysis shall be validated.
Validation can be accomplished by showing the predicted response of a mathematical model of the complex calorimeter, using a predicted thermal environment from CLASIX-3 analysis of a known condition in the 1/4 scale facility, is conservative compared to the measured response of the complex calorimeter on the 1/4 scale facility.
3.
Combustion parameters for CLASIX-3 predictions as follows shall be acceptable for validating CLASIX-3:
11 e
{
A.
Hydrogen volume percent required for ignition 6 v/o B.
Hydrogen volume percent required for propagation 6 v/o C.
Hydrogen fraction burned
.65
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D.
Minimum oxygen volume percent for ignition 5 v/o E.
Minimum oxygen volume percent to support 0 v/o combustion F.
Flame speed 6 ft/sec
,j Heat removal from the 1/4 scale facility shall be consistent 11 j
with the methodology used for full scale containment analysis.
]
l The containment spray carryover fraction in the facility shall
)
j be determinad.
1.
A CLASIX-'3 prediction shall be completed using the same 11 l i assumptions as used in previous licensing analysis.
1 Specifically, combustion'shall be initiated when hydrogen
)
concentration reaches.8 v/o with 85 % of the hydrogen burned.
]
4.
The CLASIX-3 predictions of 1/4 scale test temperatures and 11 pressures shall be compared with measured temperatures and
]
pressures.
The intent of this comparison shall be to 1
demonstrate that CLASIX-3 conservatively predicts compartment
]
average peak temperatures and pressures.
Temperatures produced
)
by any localized hydrogen combustion shall be compared with the
]
compartment averaged temperature response.
]
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ACCEPTANCE CRITERIA FOR TASK 13 COMBUSTIBLE GAS CONTROL EPG 11 A symptom based emergency procedure guideline which pr'ovides
]
1.
j guidance to the reactor operator on utilization of hydrogen
. l control equipment shall be developed.
The guideline shall
']
specifically provide guidance on use of the hydrogen igniters,
'i the drywell hydrogen mixing systems and the hydrogen l
recombiners.
Operator actions shall be initiated based upon
'S plant symptoms which are independent of a specific accident sequence.
11 l
2.
The operator actions specified in the emergency procedure
]
guidelines shall preserve containment integrity and equipment function to the greatest extent possible.
The guideline shall indicate limits for securing equipment in order to preserve 4
equipment function and maintain containment integrity.
4 I
3.
The emergency procedure guideline shall provide guidance to the operator for all postulated accidents and transients including accidents and transients which are outside the existing design basis.
This is in accordance with the requirements of NUREG-0737.
All accident scenarios and plant conditions considered in developing the emergency procedure guidelines need not be considered in licensing analysis of I
i hydrogen generation events.
M
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l 14.1 Identify All NTS Tests Completed During Program
+
A review of all Nevada Test Site (NTS) tests has been 12 conducted.
This review identified each test for addi,tional
]
{
evaluation.
All NTS tests which might provide potential l
};
information concerning hydrogen combustion assumptions used in
)
licensing analyses or concerning behavior of equipment used in Mark III containments in hydrogen burn environments were noted
]2
]
for further evaluation. These tests were identified from HCOG's monitoring of EPRI research activities in Subtask 6.8 and this review was initiated to resolve an open item with the NRC staff.
]2 1
Responsibility - HCOG i
Status
- Complete 12 14.2 Summarize Test Data A summary of test data from all applicable tests identified in
)
Subtask 14.1 will be prepared. A list of instruments with i
available data for each test with a graph or chart of the instruments response in the facility will be prepared.
i For j
tests where equipment was included, the information which exists regarding the equipment before, during and af ter the test will be summarized.
1 i
j Responsibility - HCOG i
Status
- In Progress
]2 i
i*
14.3 Identify Acolicable Tests and Eauipment Used in Mark III 4
Plants The tests identified and summarizing in subtasks 14.1 and 14.2_
will be examined to determine if test conditions are applicable i
to Mark III containment conditions.
Tests where hydrogen 4-163 2/28/85 4
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concentrations exceeding the values allowed by the hydrogen ignition system may have produced thermal environments harsher than expected in the Mark III containment after a degraded core hydrogen generation event.
Also equipment installed,in the j
hydrogen test dewar will be reviewed to determine if it is representative of any equipment used by the HCOG members.
14
-l Responsibility - HCOG Status
- In Progess 12 e
14.4 Determine Equipment Which Failed and Cause The NTS equipment data will be examined to determine which items j
of equipment applicable to HCOG member plants failed and the apparent failure mechanisms.
If a report addressing a particular equipment failure is available, then this report will be consulted.
If no information is available, the installation of the component, the stated capabilities of the component and other relevant data will be analyzed to determine if equipment i
failure occurred as a result of hydrogen combustion.
i i
Responsibility - HCOG Status
- In Progress
]2 14.5 Identify Any Differences Between Licensing Assumptions and NTS Results l
The conclusions derived from the NTS test series will be reviewed against the assumptions used for the Hydrogen Control Owners Group CLASIX-3 analysis.
Specific differences identified in subtask 14.5 will be evaluated for their effect on previous analytical work performed by the Hydrogen Control Owners Group 3
This information will be used in Subtask 8.8 to determine if i
additional generic CLASIX-3 containment response analyses with 4-164 2/28/85 4
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- 7
- m7.. f. c....a.w modified assumptions are required. Any effect on generic assumptions will be considered in the methodology validation process and documented in Subtask 12.13.
It is also possible
-]2 that resolution of Nuclear Regulatory Commission staff questions from Subtask 14.8 may also affect previous assumptions.
4 Responsibility - HCOG
- ]
Status
- In Progress
]2 4
14.6 Submit Evaluation Results to the NRC 1
The Hydrogen Control Owners Group committed to review NTS data and provide an assessment of this data to the Nuclear Regulatory Commission staff for review.
A letter report documenting the review conducted as part of subtask 14.1 - 14.5 will be prepared and submitted to,the staff for review and information.
This report will identify all tests completed, the tests which are applicable to Mark III containment plants, and the basis for concluding that other tests are not applicable to the Mark III containment.
The report will also summarize conclusions regarding causes of equipment failure as established in Subtask 14.4.
Responsibility - HCOG Status
- Not Started I
(
j 14.7 HCOG/NRC Meetings to Discuss Results The Hydrogen Control Owners Group will review the NTS data evaluation with the Nuclear Regulatory Commission staff. Any apparent differences between the NTS test results and assumptions used by the Hydrogen Control Owners Group in _
containment response and equipment survivability analysis will 4-165 2/28/85 J
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ACCEPTANCE CRITERIA FOR TASK 14 NEVADA TEST SITE DATA EVALUATION
]1 1.
Data obtained by EPRI from a series of tests conducted in a
]
large scale hydrogen dewar and intended to provide generic information on the performance and thermal response of selected i
nuclear plant equipment under a range of hydrogen burn l
environments shall be evaluated.
1 2.
Nuclear plant equipment used in the Nevada Test Site (NTS) tests will be reviewed and equipment and cables which are similar in manufacture and design to equipment utilized by HCOG
)
member utilities shall be identified.
Equipment and cables not applicable to HCOG member utilities shall also be identified.
i 3.
Equipment and components used i,n the NTS test series and similar to equipment and components used by HCOG member utilities shall be evaluated to determine all failures which occurred in tests where the hydrogen concentration was less than 10 volume. percent.
The cause of failure and, if available, the manufacturer's evaluation of the failure, shall be identified.
4.
Premixed combustion tests for hydrogen concentrations at 11 or below 10 volume percent shall be evaluated for equipment
]
performance and thermal response.
Since the distributed igniter i
system provides reliable ignition for hydrogen concentrations at 6 volume percent, concentrations above 10 volume percent are not realistic for recoverable degraded core accidents, i.
5.
Data from premixed and continuous hydrogen injection tests shall be reviewed to provide a comparison between assumptions ll l
?
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used in licensing analysis and test results.
The following items will be compared with the NTS data results:
A)
The concentration at which ignition occurs B)
Apparent flame speeds in the facility
,]
C)
Burn completeness for various cor.ditions
'I D)
Effects of steam injection on combustion E)
Effects of fans and sprays on combustion
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A 6.0 IMPLEMENTATION OF FINAL HYDROGEN CONTROL RULE
]2 A revision to Title 10 of the Code of Federal Regulations (10CFR) Part 50.44 (C) (3) was published in the Federal Register on January 25, 1985.
This revision constitutes the final hydrogen control rule for ice condenser and Mark III containments.
This section discusses how the Hydrogen Control Program Plan implements the requirements contained in the final j
hydrogen control rule.
Paragraph (iv) (A) requires each licensee with a boiling light water nuclear power reactor and a Mark III type of containment to provide a hydrogen control system justified by a suitable program of experiment and analysis.
The program outlined in this program plan constitutes a suitable program for assuring that a distributed ignition system fully complies with the final hydrogen control rule.
This program integrates generic work completed by the HCOG with plant specific work completed by HCOG
?
member utilities.
Tasks 2 and 3 in the Hydrogen Control Program Plan define HCOG's approach in selecting and designing a hydrogen control system.
Paragraph (iv) (B) requires demonstration of containment structural integrity.
Task 4 in the Hydrogen Control Program Plan defines the approach used by HCOG and HCOG members to establish the ultimate pressure capacity of the containment structure.
This task also involves comparing the peak pressures produced by hydrogen combustion as calculated in Task 8 against 1
the calculated containment ultimate capacity.
The ultimate i
capacity must exceed the peak pressure predicted by the 12 containment response analyses completed under Task 8.
]2 Paragraph (v) (A) requires each licensee with a boiling ligE water nuclear power reactor to provide the nuclear power reactor 6-1 2/28/85 4
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F-with systems and components necessary to establish and maintain safe shutdown and containment integrity.
These systems and components must be capable of performing their function during and after exposure to environmental conditions produced by hydrogen combustion.
Task 11 in the Hydrogen Control Program Plan involves identifying all equipment which must survive hydrogen combustion conditions produced by operation of the i
igniter system during degraded core accident conditions.
Task 11 also defines the app 2.oach which will be used to assure that equipment required to survive these accidents survives the environments which could be produced by hydrogen combustion.
Paragraph (v) (A) requires consideration of the environmental conditions produced by local detonations unless such detonations can be shown to be unlikely to occur.
Task 4 in the Hydrogen Control Program P_1_an includes a,n evaluation of the need to consider local detonations based on the results of this task, HCOG has concluded that the probability of local-detonations is I
sufficiently low to warrant excluding consideration of local detonations.
Paragraph (v) (B) requires that the amount of hydrogen to be considered in designing the hydrogen control system shall be equivalent to that generated from a metal water reaction involving 75% of the fuel cladding surrounding the active fuel.
Task 7 in the Hydrogen Control Program Plan includes a task to evaluate the hydrogen production from accidents which result in
~this amount of hydrogen production.
Tasks 8 and 10 in the Hydrogen Control Program Plan involve analyses of the containment response to accidents which result in hydrogen production equivalent to oxidize 75% of the fuel cladding i
surrounding the active fuel.
k 6-2 2/28/85
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6 Paragraph (vi) (A) requires each licensee to complete an analysis.
Paragraph (vi) (B) delineates the requirements for this analysis.
The analysis must include:
]2 A.
an evaluation of the consequences of large amounts of
]2 hydrogen generated after an accident,
)
I
]
B.
the period of recovery from the degraded condition, 1
I
]
C.
accident scenarios that are accepted by the NRC staff,
]
]
D.
support for the design of the hydrogen control system
]
]
E.
demonstration of the containment structural integrity
]
]
F.
demonstration.of the survivability of the systems and
]
components necessary to establish and maintain safe
]
shutdown and to maintain containment integrity.
]
]
Tasks 1 and 7 in the Hydrogen Control Program Plan involve
]
defining and evaluting the consequences of various accident
]
scenarios which result in producing large quantities of
]
Tasks 8, 9, and 10 in the Hydrogen Control Program
]
Plan involve defining the consequence of varying accident
]
scenarios on the containment and drywell compartments.
Tasks 2
]
and 3 in the Hydrogen Control Program Plan defines HCOG's
]
approach in selecting and designing a hydrogen control system.
]
Task 4 in the Hydrogen Control Program Plan defines the approach
]
to establish the ultimate pressure capacity of the containment-
-]
structure.
Finally, Task 11 in the Hydrogen Control Program
]
I Plan defines the approach which will be used to assure that
]
equipment required to survive the accidents survives the_
')
environment which could be produced by hydrogen combustion.
]
6-3 2/28/85
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- s Paragraph (vii) (A) requires each licensee to submit a schedule
]2 for complying with the final hydrogen control rule.
Section 5
]2 of the Hydrogen Control Program Plan includes a general schedule
]2 for completing the HCOG's overall program of experiments and
]2 analysis.
This general schedule will be supplemented'with a
]2 more detailed milestone schedule in a subsequent revision of the
]2
'{
Hydrogen Control Program Plan.
]2
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