ML19350E737
| ML19350E737 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 06/16/1981 |
| From: | TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML19350E729 | List: |
| References | |
| NUDOCS 8106230483 | |
| Download: ML19350E737 (53) | |
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, TENNESEE VALLEY AUTHORITY SEQUOYAH NUCLEAR PLANT ,
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- RESEARCH PROGRAM ON HYDROGEN COMBUSTION AND CONTROL '
, QUARTERLY PROGRESS REPORT #3
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JUNE 16, 1981 ^ . i
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-TABLE OF CONTENTS I. Introduction.
1II. Task Descriptioa,. Schedule, and Status A. Industry. Degraded Core (IDCOR) Program /TVA B.~ Electric Power Research Institute (EPRI)/TVA/ Duke /AEP C. Westinghouse /TVA/ Duke /AEP C.1 CLASIX Modifications D.--TVA/ Duke /AEP D.1 Halon (Atlantic Research Corporation) . D.2 Electromagnetic Interference Study D.3 Catalytto Combustor D.4 Fog-ing , E. . TVA E.1 Browns Ferry'Probabilistic Risk Assessment (Pickard,
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Lowe, and Garrick) I E.2 Sequoyah Full-Scale Safety and Availability Analysis (Kaman Sciences Corporation)
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.E.3 Consequence Analysis .
E.4 Severe Accident Sequence Analysis (SASA)/(ORNL) ,- E.5 Ice Condenser Containment Code III. Appendices
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A. Program Details " A.1' EPRI Program A.2 Fenwal Report
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IA.3 Singleton Testing. 1
- B. Equipment Survivability
- 13. Permanent Hydrogen Mitigation System (PHMS) Decision Methodology
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~ 'I. Introduction This report is the. third of a series of quarterly research summaries presented to the Nuclear Regulatory Commission (NRC)
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by the Tennessee Valley Authority (TVA) to satisfy the following condition of the Sequoyah Nuclear Plant unit 1 operating license: During the interim period of operation, TVA shall continue a research program on hydrogen control measures and the effects of hydrogen burns on safety functions and shall submit to the NRC quarterly reports on that research program. TVA is pleased to document the various facets of its current t degraded core research program in this repoat and is confident that all possible efforts have been exerted to ensure the timeliness, effectiveness, and cocpleteness of the program. Increased attention was devoted to accidents beyond the design basis in early 1980 as TVA, with the aid of Westinghouse and three architect-engineering firms, produced a report tnat has since been. submitted to the NRC on September 2,1980, as Volume I of the Sequoyah Nuclear Plant Degraded Core Program Report. TVA has remained in the forefront of industry efforts in many areas of degraded core research and development. This leadership was demonstrated by the decision to voluntarily implement the interim distribute 6 ignition system at Sequoyah to extend the plant's capability for hydrogen control. TVA has continued to voluntarily conduct its own degraded core programs
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. o and , to cooperatively participate with _other utility groups in these research efforts.
These efforts are the subject of the present series of reports. The format of this report is designed to present in Section II
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an updated summary of the changes in scope, schedule, and status of each task as described in the first report with further technical details in appropriate appendices. In the second quarterly report, we included in Appendix B our analysis on equipment survivability. TVA submitted to the NRC on June 1,1981, an updated report on equipment survivability. This was to satisfy a Sequoyah operating license condition to f provide more detailed justification of TVA's pc sition that key equipment will survive a repeated hydrogen burn environment. Since this report was a recent submittal, we will reference it but not duplicate it in this quarterly report. Also in the second quarterly report, we described the decision methodology which we are using to ec1?ct a Permanent Hydrogen Miti6: tion System. TVA is presen'ly working on a separate report which describes our choice for the Permanent Hydrogen Hitigation System. We plan to submit that report before the end of June 1981. T
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II. Task Description, Schedule, and Status-
"ne majoe emphasis of TVA's current research program is to discover, collect, and evaluate enough information about degraded core events and potential mitigations for their risk reduction' to be ableL to. select, design, and install a permanent hydrogen mitigation system (PHMS) Isr Sequoyah Nuclear Plant.
This permanent system would satisfy the following condition of the unit 1 operating license: For operation of the racility beyond January 31, 1982, the Commission must confirm that an adequate hydrogen control system .for the plant is installed and, will perform its intended function in a manner that provides adequate safety margins. Figure 1 and Table 1 show a schedule of activities necessary to meet this unit 1 licensing condition. This section provides a summary of each individual or group effort in which TVA is actively involved that is related to hydrogen combustion and control, risk assessment, or overall degraded core studies. Here, current upda',es of scope, l
- schedule, and status of each effort are "ummarized with further details presented in the appendices.
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TABLE 1 TVA HYDROGEN MITIGATION PROGRAM SCHEDULE
. Submit quarte-ly report #2 to NRC' 3/16/81
. Complete preliminary evaluation of conceptual designs 4/1/81
. Submit ;esalution of IDIS equipment survivability to NRC 5/81 l
. Complete evaluation of conceptual designs 6/1/81
. Select PHMS and submit preliminary information to NRC Mid-81
. Submit quarterly report #3 to NRC' 6/16/81
. C0mplete EPRI igniter development program 9/81
. Complete EPRI combustion studies 9/81
. Complete EPRI H Control Studies 9/81
. Submit quarterly report #4 to NRC' 9/16/81
. Complete final system design 9/81
. Submit final safety analysis report to NRC 10/81
. Complete procurement of equipment 11/81
. Complete EPRI mixing studies 12/81
. Submit quarterly report #5 to NRC' 12/16/81
. Complete installation and testing 1/15/82
. Receive NHC approval and removal of OL condition (22)D(2) 1/31/82 eQuarterly reports will provide summary of progress made toward identified milestones.
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.' OVErTVIDi SCHEDULE SELECTION, DESIGN,AND INSTALIATION OF PERMANENT MITIGATION SYSTD4
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gjgy PRELIMINARY SYSTEM DESIGN
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'.4/81 mmTION/COWARISION OF ,
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6/81 -*-- 4 3 -- PRELIMINARY INFORMATION ON SELECTED PERMANENT SYSTEM SUBMITTED TO NRC; NRC FEEDBACK __ 7/81
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FINAL SYSTEM DESIGN 8/81 ..
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9/81 - 41[ 4 -- PR,emEMENT
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FINAL INFORMATION ON PERMANENT SYSTEM SUBMITTED TO NRC; NRC 10/81 -- -- FEEDBACK
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11/81 INSTALLATION AND PREOP TESTING
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(PARTIALLY OtfrAGE DEPENDENT) LEGEND: 12/81- -- Q 5 CM - TVA/NRC COORDINATION MEETING
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QR - QUARTERLI RESEARCH REPORT DP - DECISION POINT FOR PERMANDTP SYSTEM 1/82 .. APP - APPROVAL OF PERMANENT SYSTEM
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e A. , Industry Degraded Core (IDCOR) Program /TVA (identified in
.the first quarterly report as AIF proposal)
A.1 .9 cope No update necessary. A.2 Schedule No update necessa"y. A.3 Status TVA has joined the IDCOR Program and is actively participating in all three levels (Policy Committee, Steering Committee, and Technical Advisory Committee) of the Program management. Technology for Energy Corporation (TEC) has been selected as overall Program Manager. The IDCOR Program does take into consideration the existence of the EPRI/TVA/ Duke /AEP Hydrogen Program (see Section II.B. and Appendix A.1). Note that the latter program is on a much expedited schedule relative to IDCOR. Attachment A-1 is the latest IDCOR Program Report. It provides a summary of the history and current status of this industry effort.
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z.- IDCOR Program Report uay1981 PROGR AM FORMATION IDCOR Formally Constituted The Industry Degraded Core Rulemaking (IDCOR) Program concept developed in December 1980 in late 1980,largely through the efforts of the Atomic Irdu< trial Forum's Policy Committee on Nuclear Regulation, was formally constituted on December 12, 1980 at a meeting of nuclear industry sponsors, subsequently designated the IDCOR Policy Group. IDCOR An Industry Respons. Tb t IDCOR Program was initiated in response to recent NRC licensing actions. To Recent NRC Actions spuific plant actions as well as a Proposed Rulemaking on degraded core issues, Suggesting Mitigation Designs which suggested that the NRC was rnoving in the direction of significant
,of Considerable Financial additional mitigation devices. In October of 1980, the NRC had issued in aad Operational rinpacts Advance Notice of Rulemaking on the subject of degraded core accidents.The wiih Unclear sarery Benefies stated purpose of the Rulemaking was to consider whether nuclear power plants should be designed to accommodate degraded or melted core accidents.The suggested mitigation fixes, such as filtered vented containments and core ladles, could have significant financial and operationalimpacts on operating nudear power plants - with uncertain safety benefits.
The highly technicalissues likely to be raised in the degraded core rulemaking are of a generic nature, applying more or less to all U.S. nuclear power plants and many foreign reactors. Program Objectivest Conei. . The goals of the IDCOR Program are to develop technical data necassary to
' Legical and well.noeumented determine whether changes in reSulations are needed, and if so, to aid in Tech nical Bases establishing regulations that are:
- Consistent with an overall nuclear safety goal;
- Based on thoughtful analysis which carefully considers the costs and benefits of the design or operationalimprovements which may result from ebe implementation of any new regulations;
- Expressed so as to minimize uncertainties with regard toits interpretation
- and implementation.
In support of these goals, the Program objectives are to develop concise, logical and well. documented technical bases for use in the degraded core rulemaking proceedings and to coordinate, where applicable, th: presentation oft he industry's technical bases. PROGRAM STARTUP , ACTIVITIES Under the directics of the IDCOR Steering Croup. the Program has gotten orf to a fast start.The Steering Group, responsible for developing and implementing programs consistent with the overall Program philoscphy, has been meeting regularly since January 12,1981. Major accomplishments to date have included the:
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The IDCOR Orgenlantion[[' .~'
- Completion of the IDCOR organization (Exhibit 1).The Stecting Group:
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A Program Manager. p , - Nominated Technology for Energy Corporation to be the IDCOR Technical Advisory GronP Program Manager; an Leg *IG' "P
- Selected the Technical Advisory Group (TAG) and chose Dr. Miles
*. Leverett of EPRI to be the TAG Chairman; I
The Industry Degraded Core Rulernaking Prograrn, Sponsored By the Nuclear Industry
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For Ipfoernation
Contact:
, Attcchm:nt A Joh R. Siegel Sperial Licensing Projects Manager /IDC')R Aininie Industrial Forum. Inc. 7101 Wisconsin Avenue Washingion. D C. 20014
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Telephone: (301) 654 9260 TWX 7108249602 ATOMIC f 0R DC Pro, grani Ryort
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- Determined a process and schedule for selecting the IDCOR Legal Group.
The selection process should culminate with a recommendation to the IDCOR Policy Group in June 1981. The Steering Group has also developed a working Charter for the Program. The IDCOR Progsam Plan
- Initiaticn of the Program's technical activities. HeHed by Dr. Anthon) ,
will be Available in Buhl, Technology for Energy Corporation (TEC) is currently: J:ne9e: - Conducting :n assessment of activities-research, analytic studies and programs-related to the technical scope of the IDCOR Program-
- Developing a detailed Piogram Plan; the plan should be completed by lune 1981;
- Cor..p ting plans to obtain the Program's technical contractors; TEC plai.s to ost a contractor information meeting in Knoxville on May 21, 1981.
The Assessment ofRelated
- TEC has been exhausuve in is. dysis of activities related to the technical Programs Included A scope ofIDCOR.TEC has conducted technical exchange meetings with:
Pr od uctive, DO E.5ponsored
- Industry - Commonwealth Edison Company on the Zion Indian Point Wor kshop Involving IDCOR study;TVA, Duke Power Company and American Electric Power aid the National Labs Company on the ice cordenser study; General Electric and EPRI, general overviews on degraded core technical issues; and Philadelphia Electric Company on the Limerick study.
- DOE and the National Laboratories (ANL, ORNL, BNL HEDL. LASL, Saadia. INEL,and Battelle Memoriallnstitute) -On April 14 15,1981, DOE sponsored a workshop which brought together IDCOR and the National Labs. The two-day session proved to be extremely productive, not only toward educating IDCOR regarding the various degraded core related activities at the laboratories, but as a management technique for enhancing communication and cross pollination among the laboratories.
roreign Prcg,amir Eager - Foreign entities - KFK. IKE. GRS, KWU, Battelle Institure Frankfurt, so Work with the IDCOR rnembers of the Reactor Safety Commission, and utilities in the Federal Pr og r a m Republic of Germany; and EDF and the CEA in France. All were enthusiastic about cooperating on a technical basis with the IDCOR Program. NRC ACTIONS RELATEDTO TIIE D EGRA DED CORE RULEMAKING The NRC Continues to The rationale for the IDCOR Program continues to appear extremely televant. Emphasize Mitigation of Recent activities suggest that the NRC will emphasize mitigation of Class 9 type Degraded Core Accidenta accidents in a degraded core rulemaking. The ACPS subcommittee on Class 9 accidents convened on March 10-11, 1981 to discuss the conclusions of an NR'., Staff report on source terms.
" Technical Bases for Estimating Fi.sion Product Behavior During LWR Accidents", and to discuss the report's regulatory impacts. The Staffs report appears to be at odds with earlier industry reports suggesting that iodine attenuation factors in regulatory analyses are overly conservative. Regarding
, potencial regulatory impacts, the Staff suggested that a wider spectrum of accidents (e.g., Class 9) might have to be considered in the future.
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& s Attachment A-1 On March 24,1981, the ACRS met to di, cuss the status of degraded core rulemaking activities with the NRC Staff Steering Group on Degraded Cooling. At thic ri.ne, the Staffindicated that: e It was anempting to formulate a research program and that this program would focus on mitigation;
- The program would involve a two phase effort - a preliminary phase to .
screen alternative design features (4/81-1/82) and a speci6c design feature analysis phase (7/f -4/83); e les preliminary schedule suggests a draft rule to the Commission by March 1982 and a final rule byjune 1983. The Staff also indicated that it may be placing additional emphasis on operator actions and that the engin. cred safety feature rulemaking - one that the Staff has viewed as focusing on preventive design features - could be conducted in parallel with the degraded core rulemaking. F211owing Their Forwarding In early April,the NRC Stecting Group on Degraded Cooling forwarded an af a Mitigation. Focused Action Plan on the degraded core rulemaking to the NRC's Executive Director for Action Plan to the Executive operations.The plan incorporated most of the Staff's discussion at the Marsh 24 Direceor's OfTice in April, ACRS meeting and included a mitigation (c.cus. the NRC's Steering Group The NRC's Steering Group was then disbanded. Responsibility for degraded c Degraded Cooling was core issues will pass to the recently reorganized offices of standards and regulatory Disbaaded research, possibly 'o the division of risk analysis. !:is not yet clear the exterit to which the transfer of responsibility on degraded core issues will impact NRC thinking regarding the focus and scheduling of degraded core rulemaking activities. PROGR AM SPONSORSillP , Program Participation is At this time progam sponsorship is estimated at 90% or 59 'nillion over the Estirnated at 90% or two. year effort. On January 30,1981, the AIF mailed invoices totaling $2.5
$9 Million million, and covering the first six months, to 75 potential sponsors. As oftarly May, about $1.9 milhon had been received and an additional 5250,000 had been committed.
The AIF is providing statTfunctions for the IDCOR Trogram - planning, analysis and liaison. john Siegel, AIF's Manager of Spesial Licensing Projects,is performing these functions. AIF . vill also serve as IDCOR's finance and contracting agent. These services, donated by AIF, h:.ve been estimated at
$300,000 for the duration of the two-year Program PROGR AM MILESTONES IDCOR Meets with NRC On April 21,1981. john Selby, the IDCOR Policy Group Chairman, and Cordell Reed, the IDCOR Steering Group Chairman, m:t initially with Joseph Hendrie, Chairman of the NRC; and later with Harold Denton, of the Office of Reactor Regulation, and Robert Minogue, Denny ? oss, and Tom Murley, of the Office of Research.The main purpose of this initial formal contact with NRC was to emphasize the formation, rationale, and methodology of the industry effort and industry's wish to keep abreast of related NRC developments.
NRC is Fully Veried on the Based on this meeting (and as reported in INSIDE NRC, May 4,1981),it appears IDCOR Program and is ' that the NRC: Seeking to Enhance Technical
- Clearly recognites the visibility and importance of the IDCOR Prog ~ ram:
Eacha nge witb IDCOR
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- Wants to cooperate technically (in fact NRC-sponsored nationallabs have requested IDCOR review of their program plans in degraded core technical areas);
e is incorporating a risk reduction methodology (probabilistic risk assessment type), which suggests that preventive and mitigative design features will be assessed on the same basis; -
- Will conduct its engineered safety feature and degraded core rulemakings in parallel.
JUNE 17 POLICY UROUP MEETING A meetieg of the IDCOR Policy Group has been tentatively scheduled forJune 17 1981. in the AIF's Washington office. Agenda items willinclude introductions of the IDCOR management team, a program briefing, focusing on tic Program Plan, by Dr. . anthony Buhl of Technology for Energy Corporation, a briefing by Dr. Miles Leverett on the Technical Advisory Group's composition and charter, review / approval of the IDCOR Charter, a financial report, and discussion / actions on Steering Group recommendations regarding foicign participation in IDCOR and technical relationships with NRC. IDCOR MANAGEMENT STRUCTURE IDCOR PotJCY GROUP
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. Oversil op.esten Program Reves /approvaloe l
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TE CH NIC AL IDC04 - LEGAL ADVISORY GROUP PROGR AM MANAGER GROUP
. ConswHeieneeevow or . Managemeat of day to. der . Conswhai.on sad scpport techn. cal progtem deveiopment on legsi esswas and .soue ensey..e . op.ratens Techncel program devek,pment
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The Industry Degraded Core Rulemaking Program. Sponsored By the Nuclear Industry
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- . ]'y w B. L Electric Power Research Institute (EPRI)/TVA/ Duke /AEP B.1 Scope The scope of the four EPRI research programs is unchanged. However, to facilitate the review of the program, we have in this revision consolidated the information we have about each project, its scope, description, and current status. We have included this latest information, including new sketches showing proposed test configurations for some tests, in the revised Appendix A.1.
The EPRI research program is currently at a stage where test matrices are being finalized and tests are commancing. B.2 Schedule EPRI has recently issued a revised schedule which includes the latest progress and changes for all four of the EPRI tests. The revised schedule is shown-in Figure II.B-1. The most significant change involves shifting the Hanford tests from the May through September test window to the July through November timeframe. The Hanford tests were delayed due to unexpected difficulties in the contract negotiations between EPRI and the U.S. Department of Energy (DOE). A contract has now been signed and work is progressing. B.3 Status The revised Appendix A.1 includes the latest status for each of the four projects.
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HYDROGEN COMBUSTION AND MANAGEMENT PROGRAM
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SCALE 1982 7/1 9/1 11/1 -1/1 (1033 FT ) 11/1 1/1 3/1 5/1 0.001.& ' IGNITER DEVELOPMENT (WHITESHELL/. ) 0.6 HYDROGEN COMTROL STUDIES (ACUREX) 1 0.2 HYDROGEN COMBUSTION STUDIES (AECL, WHITESHELL) i 30 _ _ _ _ HYDROGEN MIXING STUDIES _ ____ (HEDL-W) , t . . i
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' C. ' Westinghouse /TVA/ Duke /AEP TVA, Duke, and AEP are cooperating with Westinghouse in the modification of the Westinghouse / Offshore Power Systems CLASIX computer code.
C.1 CLASIX Modifications
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C.1.1 Scope No update necessary. C.1.2 Schedule '
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No update necessary. C.I.3 Status
,- The modifications discussed in our first and second quarterly reports have.been incorporated
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into the CLASIX code. A set of runs for the , Sequoyah plant have been completed using the ; modified version of CLASIX. Results from these ; runs were provided in our June 2,1981, submittal that addressed conditions on the Sequoyah license,
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D. TVA/Duko/AEP D.1 Halon ( Atlantic Research Corporation) D.1.1 Scope No update necessary. D.1.2 Schedule No update necessary. D.1.3 Status The final ARC Halon report was received by TVA/ Duke /AEP on May 43, 1981. The final report and our evaluation will be included in our June 1981 submittal to the NRC which describes the selection of the Permanent Hydrogen Mitigation System. TVA's revie/ of ARC's conceptual design of a Halon mitigation system indicates that the system could suppress h / drogen combustion in a postaccident environment. However, TVA conducted a study at their Singleton Materials Laboratory to investigate the corrosive effects of Halon decomposition productz on vital equipment inside the containment during the postaccident recovery period. These studies, ,. using the concentrations deterOined by Dr. S. ' A. Turner and presented in the fins' 'eport, showed that there was a virtual certaint; that stress corrosion cracking would occur at any point where sensitized steel (i.e. welds) l 1
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W. D.2 Electromagnetic Interference _(EMI) Study.
-D.2.1 Scope
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No update necessary. 4 D.2.3 Status Tests were conducted at Watts Bar Nuclear Plant on March 16 and 17,_1981, - to determine the
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l susceptibility of'several types of analog transmitters-to EMI. A Flaregas Model ETX-105 Igniter was used as-the source of the electromagnetic emissions. The transmitters under test were:
- 1. Foxboro Model E11GN, Pressure Transmitter
- 2. Foxboro Model E11DM, Differential Pressure Transmitter 3 Barton_ Hodel 396, Differential Pressure Transmitter
- 4. Barton Model 763, Strain Gauge Pressure Transmitter
- 5. Barton Model 764, Differential Pressure Transmitter The igniter was located at a distance of four
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Test results indicated only one or the transmitters, Barton Model 764, was susceptible to the radiated emissions from the spark igniter. However, the installation of this transmitter during the test was not representative of the normal installation. The normal loop was not available for testing; therefore, a epecial installation was required. This consisted of routing a temporary, twisted pair shielded cable from the transmitter under test to an instrument rack approximately 30 feet away. The cover of the transmitter terminal box was not installed in order to accommodate the temporary cable. In this configuration, the transmitter was very susceptible to the radicted omissions from the spark igniter. The cover of the terminal box was then installed loosely, but the interference from the tgniter was still above acceptable limits. After the cable connecting the transmitter to the instrument
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box installed correctly,- no radiated '
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transmitter indicates the effects the physical installation of the loop has upon the
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Background noise marked the igniter's power line emissions, and it was not possible to determine the susceptibility of the transmitters to this conducted ' interference. After conducting this study, TVA believes that further_ research is still needed but not warraated. Based on the proven acceptability and effectiveness of the thermal igniters, TVA has eliminated the spark igniter as a hydrogen control method and terminated its studies into EMI.
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D.3- catalytic combeator D.3.1 . Scope
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TVA/ Duke /AEP on May 15, 1981. comments are being compiled for resolution by the Acurex Corporation. The report will be submitted as
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evaluation of the catalytic combustor concept will be included in our June 1981 submittal on the selection of the Permanent Hydrogen 1 Mitigation System. u
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. The Acurex test facility is unde'rgoing final-preparations for use. in-mitigation research.
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f E. TVA LIn addition to the preceding, TVA is independently pursuing other areas or degraded core studies which are outlined.in-
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;E.1 . Browns Ferry Nuclear Plant'Probabilistic Risk Assessment (Pickard, Lowe, and Garrick)
E.1.1 Scope No update necessary. E.1.2 Schedule The study began 10/80. Revised dates for major milestones
. Data analysis, event trees, and fault trees - 8/81
. Explant consequence model assessment -
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. Seismic analysis reports - 7/81
. Explant consequence analysis - 9/81
. Final report
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; Tasks completed or' presently underway:
. Event sequence diagrams
. Event trees
. System quantitative analysis
.. Maintenance-and test data
. Seismic analysis
. External events analysis .
. TVA code conversion Near future tasks: .
. System quantitative analysis
. Containment analysis E.2 Sequoyah Nuclear Plant Full-Scale Safety and
~ Availability Analysis (Kaman Sciences Corporation)
E.2.1 Scope No update necessary. E.2.2 Schedule Phase I (preliminary availability assessment) 1/81 Phase II-A (preliminary safety assessment) 5/81 Phase II (final assessment) 6/82 E.2.3 Comprehensive system models have been developed for three systems ( Auxiliary Feedwater, Main Feedwater, Safety Injection) Preliminary models have been developed for all other. systems which are included in the _ _ _ _ -,
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Preliminary availability and safety assessments have been made. J t 4 l l
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- E.3 Consequence Analysis.
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E.3.1 : Scope -
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No. update necessary. E.3.2 - Schedule Obtain-CRAC2 computer program from Sandia 7/81 _
' Evaluate MARCH / CORRAL 2 and.KESS as to similitudes and differences in the prediction' of LWR Class 9 accident s consequences 12/81 Comparison of postulated accident
- sequences for SQN using MARCH / CORRAL 2
, and KESS 4/82 E.3 3' Status
, Completed training in use of MARCH, CORRAL 2 e and dose prediction codes 3/81 High probability accident sequences identified 4/81
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E.4' Severe Accident Sequence Analysis (SASA)/(ORNL) E.4.1 Scope No update necessary. E.4.2 Scheduls I. Browns Ferry Station Blackout Analysis A. Complete Analysis of Blackout Sequence 4/81 B. Issue Draft Report 5/81 C. Complete Peer Review 8/81 D. Issue Final Report (NUREG) 10/81 II. Analyze Other BWR Sequences Identified by NRC/INEL A. Provide Accident Signatures for 8 BWR Sequences 10/81 B. Identify Sequences for Follow-on Detailed Analysis 11/81 C. Perform analysis of failure to isolate a small
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break LOCA outside contaiment 1/82 E.4.3 Status 5/81 RELAPS model completed for BFN-1 5/81 Draft report for BFN-1 station blackout completed 7/81 Expected completion of predicted station blackout fission product release and
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E.5 - Ice Condenser- Containment Code
.E.5.1 Scope TVA,_ Duke, and AEP have recently completed negotiations with. Westinghouse /0PS to procure the propri9tary CLASIX computer code. At this.
. time, . therefore, we have not decided to use ~ the ice condenser containment code we are developing on licensing-related work. We are, how1ver, continuing the code's development for use in other areas.
E.5.2 Schedule
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The code development program consists of three
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E.5 3 Status Part I - Obtain a working code that will analyze conventional containnent transients. * < l Numerous corrections and modifications to
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CONTEMPT-4 were necessary P.o make the code [ operational. Some of the major corrections. l L required included changes in the input routines, L flow correlations, evaporation-condensation model, ice condenser flow model, and the fan ! model. Because of the extensive modifications, the code has been renamed MONSTER. Rurs have
- been made comparing results from MONSTER with results from COMPARE, CONTEMPT-PS, and LOTIC.
The cases chosen for this initial comparison !-
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l I were a 17-node analysis of a break in a pressurizer enclosure, an analysis of a large LOCA, and an analysis of a large LOCA at Sequoy 5 using two million pounds of ice. A review of the results shows good agreement with the results predicted by the other codes. .
. Comparisons are planned in which small LOCA's and main steam line breaks will be analyzed.
Part II -Expand and modify the code to allow - evaluation of class 9 events. Modi; cations have been incorporated into the code to permit the analysis of class 9 events. i A burn model has been incorporated into the code along with code consideration of individual gases including hydrogen, carbon monoxide, and halon in addition to the air constituents. Fenwal test data and CLASIX results for an S D2 event are being used in comparison with MONSTER for degraded core events. Results from this work are not yet available. Part III - Continuing development of a best estimate code - ongoing beginning. In the near term, efforts are underway to
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APPENDIX A.1
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EPRI PROGRAM . a l
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,e s APPENDIX A.1 EPRI PROGRAM - TECHNICAL DETAILS
' NOTE: The following updated appendix was submitted to the NRC on May 15, 1981, and -is reproduced here for convenience.
1.0 Introduction In order.to design 295 install a permanent hydrogen mitigation
- system at Sequoyah Nuclear Plant, several questions concerning hydrogen management should be addressed.
TVA, American Electric Power (AEP), and Duke Power have entered into a research effort with the Electric Power Research Institute (EPRI) to study hydrogen mitigation and '
- control under degraded core accident conditions. The following is a description and status of that program to date.
1.1 Objectives and Technical Issues. This program is intended to meet the following limited objectives:
- 1. Determination of whether and when hydrogen can burn in postulated ice condenser accident environments resulting from degraded core scenarios;
- 2. Demonstration that if a hydrogen burn does occur, its effects will not exceed the realistic survival capabilities of equipment and containment; and 3 Demonstration that reasonable control methods can
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provide adequate safety margins assuring the integrity of the containment and of key safety-related equipment.
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Determination of the' effects of hydrogen deflagrations on
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. Determination of the effects of hydrogen deflagrations on containment and equipment requires investigation of several
_ questions, in particular:
- a. What are *.he lower flammability limits under degraded core accident conditions and how effective are thermal ignition sources;
- b. What is the character of deflagrations in various geometries and how can the effects be mitigated;
- c. What is the nature of hydrogen mixing and distribution in large compartmentalized volumes; and
- d. What is the potential for the acceleration of deflagrations, or fbr flame propagation between compartments in turbulent mixtures.
1.2 Program Elements. We feel the following projects wil" ' provide the information needed to satisfy the program objectives. The projects are related, and consist of:
- 1. Development and preliminary testing of thermal igniters for a deliberate ignition system ( AECL Whiteshell);
- 2. Experiments and analyses on basic hydrogen combustion phenomena including the effects of steam, turbulence, and flame propagation between compartments ( AECL Whiteshell);
3 Experiments on hydrogen control methods including water spray and fog (Acurex/ Factory Mutual);
- 4. Measurement and analyses of hydrogen mixing and distribution in a large compartmentalized volume (HEDL- W ).
Figure 1 shows tbze scales for the projects, including the test phase (solid line).
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2.0 Description of the Program Elements The four projects listed above will be accomplished at four facilities-(AECL Whiteshell, Acurex, Factory Mutual, and Hanford ~ Engineering and Development Laboratory). The following is a description of each of the four projects and the facilities they will use._to meet their program objectives. 2.1 ' Igniter Development (AECL Whiteshell) . The interim distributed ignition system which TVA installed in September of 1980 in Sequoyah Nuclear Plant employs a diesel engine glow plug. The ability of _ this glow plug to ignite hydrogen at low concer.trations under varicus environmental conditions of steam, pressure, henidity, water spray, and airflow across the igniter surface was demonstrated during the Fenwal test. The current EPRI program is designed to compare four thermal igniter types, including the glow plugs, for effectiveness under an
, expanded range of environmental conditions and to compile data for comparison with the larger body of existing hydrogen combustion data.
This program will be accomplished at AECL Whiteshell in a 0.6 ft3 vessel which is approximately 12 inches high by 14 inchen in diameter. Four types of thermal igniters will be subjected to identical test conditions. The igniters will be evaluated for their ability to ignite hydrogen in the range of 5 to 12 volume percent under various steam
. concentrations in the range of 0 to approximately 60 percent. The test will be conducted both with and without fan-induced turbulence. The vessel will be instrumented
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with fast response thermocouples (10 microseconds) and pressure transducers (1 microsecond) tied into a computerized data acquisition system. - Current Status The vessel has been instrumented and is currently undergoing shakedown tests. The testing program should begin within the next two weeks. 2.2 Hydrogen Combustion (AECL Whiteshell)
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The objectives of trmse tests are:
- 1. To provide fundamental understanding of n e mbustion; 2
- 2. To establish comparison between glow plug data and spark ignition data.
- 3. To study the effects of turbulence created by obstacles; and
- 4. To study the flame propagation between different concentrations of hydrogen.
In order to accomplish these objectives, the tests will be conducted in a 8-foot diameter, 220-cubic-foot sphere, and when necessary a connected 12-inch (id) pipe. The sphere has a design pressure of 1450 lbs/in2 . Both the sphere and the pipe are trace heated and insulated.
>s vessel will be instrumented with fast response thermocouples (10 microseconds) and pressure transducers (1 microsecond) as well as ion probes from which flame speeds can be determined. This vessel, like the smaller vessel discussed in 2.1 above, will employ a computerized data acquisition system. The vessel will also be equipped with
- two fans, the details of which are noted on Table 3, and
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provisions. for gas sampling both before and after each testing using gas chromatography. The tests have been divided into four parts to st udy: .
- 1. The lower flammability limits in steam,
- 2. Laminar spherical deflagrations, 3 The effects of turbulence on deflagrations, and 4 Propagation of deflagrations between connected volumes Current Status The 220-cubic-foot sphere has been instrumented and is currently undergoing shakedown tests. Both these combustion tests and the igniter development tests are being conducted-in parallel at Whiteshell. This particular testing program is expected to begin collecting test data in approximately one week.
The following is a description of each of these parts of the program. A test' matrix which summarizes these tests is
presented ac Tables 1 through 4. 2.2.1 Lower Flammability Limit and Extent of Reaction. At concentrations below 10-percent H2, the combustion reaction may not be complete. These experiments are designed to determine whether size and shape of the vessel affects the extent of reaction by comparing the results with those obtained previously at Whiteshell in a small cylindrical vessel. In these experiments, uniform hydrogen / air / steam mixtures will be spark-idnited at the bottom of the
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a spherical vessel. Pp analyzing the mixture-before
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and after the experiment, using gas chromatography, the extent of reaction- will be determined. Transient pressure and temperature measurements will be made and ionization probes _ will be used to detect flame speeds. . The flammability limits and extent of reaction depend on the location of tho ignition source, and some experiments will be performed with central and top ignition which aae known to be less effective in producing a reaction. 2.2.2 Laminar Sherical Deflagrations. At hydrogen concentrations greater than 10 percent, complete combustion is expected. Burning rates are important for determining the effects of high flame i speeds and for estimating the time during combustion that heat cransfer can occur. A correlttion for laminar burning velocity as a function of hydrogen concentrat3 in and temperature has already been developed at Whiteshell based upon their bench scale experiments. This correlation allows a prediction of the pressure transient resulting from a laminar deflagration, and it is the intent or these experiments to validate those calculations. Central ignition of uniform hydrogen / air / steam mixtures in the 8-foot sphere will be performed for hydrogen concentrations in the range of 10 to 42 percent. Maximum burning velocities and the fastest deflagration transients are expected at 42-percent i
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hydrogen. High speed pressure and temperature transducers .will record ' the transients and ionization probes will detect flame arrival. 2.2 3 Effect of Turbulence and Structures on Spherical o_ Deflagrations. Since turbulence can significantly accelerate the combustion rato, these experiments are proposed' to examine this effect. ' Turbulence in containments may be caused by convection currents from the vent?lation fans and by obstacles (such as pipes, grids, etc. ). The proposed experiments would
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utilize: . (1) a fan to produce convection flows prior to controlled ignition of uniform hydrogen / air / mixtures; and (2) obstacles to determine their effect. The obstacles to be used t'or these tests would consist of 1/4-inch perforated plates placed horizontally at two different elevations within the sphere. The perforations in the plate are one inch in diameter and are spaced equally over the surface of the plate leaving approximately 50 percent of the area of the plate open to flow. The elevations of the two plates will be approximately 2.6 feet and 5 3 feet from the bottom of the vessel. The experiments will be performed in a manner similar to those described previously. 2.2.4 The Sphere and Connected Pipe. Tne Whiteshell high pressure sphere and pipe will be combined to study propagation of deflagrations between connected volumes. Unequal concentrations of hydrogen in different volumes will be ignited in one of the
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pressure wave 'till be measured.. The -pipe is 20 feet long by 1 root in diameter. 23 Hydrogen Control Studies (Acurex/ Factory Mutual) The purpose of these tests _ is to study the effects of sprays and fogs on hydrogen deflagrations. This will be accomplished by studying the interrelationship between water droplet size, water concentration, and hydrogen concentrations in small scale tests at Factory Mutual. Based on the data obtained faom Factory Mutual, large scale tests will be conducted by Acurex to study the pressure supprossant effects of sprays and fogs on hydrogen deflagrations. 2.3 1 Factory Mutual Studies A schematic of the experimental apparatus to be used is'shown in Figure 2. The test vessel is a , Plexiglass tube approximately 3-1/2 feet long and 6 inches id. The ignition source will be a spark
. igniter. Ionization probes will verify the presence of a flame during an ignition attempt. Thermocouples will measure the test vessel's atmosphere temperature.
The current test matrix will provide data for droplet sizes in the range of 10-400 microns and water concentrations varying from 0-5 volume percent. The hydrogen concentration will be varied from 4-12 volume percent. These tests will provide information on conditions that produce inerting. From this data, spray droplet sizes with potential to reduce the
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maximum pressure from controlled burning will be
. selected for further study at the Acurex test facility.
Current Status Factory Mutual is currently performing droplet
- characterization studies and facility checkout tests.
Testing will begin June 18, 1981, and is expected to be complete on or around July 8, 1981. Factory Mutual expects to complete their final report by August 6, 1981. 2.3.2 Acurex Studies Figure 3 provides a flow diagram of the racility to be used. Vessel instrumentation is shown in Figure
- 4. The vessel is approximately 17 feet high with a 7-foot id. The volume is 630 cubic feet.
Instrumentation includes pressure transducers, Type K thermocouples, and ionization probes. Ignition will be by spark and glow plugs. This will allow correlations to be made between ths Factory Mutual studies and postulated conditions inside containment. Although the exact test conditions will not be determined until the Factory Mutual studies are well underway, the range of parameter values will be within those specified in Section 2.3 1. These tests will provide information on the pressure suppressant effects of fogs / sprays oa hydrogen deflagration. , w
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. l O' D j From this information, the relative merits of incorporating fogs / sprays into a hydrogen mitigation system can be evaluated. Current Status The Acurex test vessel has been instrumented. The vessel, hewever, has not been delivered to the test site. Acurex expects to move the vessel and begin shakedown tests by June 1,1981. Testing will begin by approximately June 15 and be completed on or around July 20, 1981. Acurex expects to complete their final report by September 14, 1981. 2.4 Hydrogen Mixing and Distribution Studies (Hanford). The purpose of these tests is to quantify the degree of hydrogen mixing in a simulated ice condenser geometry under accident i conditions. Compartments will be built in the 30,000-cubic-foot containment system test facility to simulate a simplified ice condenser containment geometry (see Figures 5 and 6). The tests will then be performed with scaled hydrogen and steam flows based on average release rates obtained from the MARCH computer code. Hydrogen ca1centrations and temperatures at various spatial locations will be measured as a function of time. Other measurenents to be taken include water / vapor concentrations, and convective gas flow patterns in a few key locations. Testing will be restricted by safety regulations to use of only 4-volume percent hydrogen since the vessel is located in a building which houses other experiments. Identical tests will be run with
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both helium and hydrogen, and a correlation between the two will be obtained to allow the majority of tests to be run
'with helium.
Status Hanford has not as yet finalized their test matrix. Currently, they have been concentrating on selecting appropriate scaling factors and test geometry. The test , vessel will be available starting July 7, -1981, and testing is expected to begin by August 7, - 1981. 1
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TABLE 1 Test Matrix for Whiteshell Experiments
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- 1. Extent of Reaction of Lean Mixtures Exp. # SH, -5H,0 Ignition -
~ u 1-3 5.0 0 bottom 4,5 6.5 0 bottom 6 8.0 0 bottom 7-9 5.0 15 bottom 10,11 6.5 15 bottom 12 8.0 15 bottom 13,14 6.5 30 bottom 15 8.0 30 botton 16 6.5 0 center 17 6.5 15 center
' 18 8.5 0 top 19 B.5 30 top Note: Steam concentrations are in the process of being revised upward to address current NRC concerns.
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- TABLE 2 2.'1 Laminar Spherical Deflagrations-
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Exp. ( -g _ [Hg _ Ignition 1 10- 0 . center
, 2 20 ' 0 center
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3 30 0 center 14 42 0 center - 5 10 10 center a 6' 20- 10 center
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7 30- 10 center
- 8. 42 10 center.
9 10 20 center-10 20 20 center 11 30 20 center 12- 42 20 center
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13 10 30 center 14 20 30 center 15 30 30 center
- 16 42 30 center 17 14 0 bottom 18 20 0 bottom Note: Steam concentrations are in the process of being revised upward
. to address current NRC concerns. , e l l
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- TABLE 3 3 Effect of Fens and Obstacles Ex. # $H, jH,0 Fan Speed Grating Ignition 1 6 0 50% No bottom 2 6' O 100% No bottom 3 7 0 1001 No- bottom 4 6 0 0 Yes bottom 5 7 0 0 Yes bottom 6 7 0 1005 Yes bottom 7 14 0 100% No center 8 20 0 1005 No center 9 14 0 0 Yes center 10 20 0 0 Yes center
, Fan Particulars Grating Particulars blade tip diameter: 16 in. type: 1/4" perforated plata air deflectors hole size: 1: dia. max. speed : 1800 rpm blocked area: 505 max. flow rate : 1500 cfm spacing: 2 plates placed horizontally at 2.6 and 5.3 (sphere vol = 220 f t 3 feet from the bottom of the No. of fans t 2 sphere continuously variable speed
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- TABLE 4
- 4. Sphere and Protruding Pipe s
Exp. # %Hg TH 0 Ignition 1 8 0 sphere-ecnter 2 20 0 sphere-center 3 8 0 pipe-end 4 20 0 pipe-end 5 8p/6s O pipe-center 6 10p/6s o pipe-center p
- pipe s = sphere
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HYDROGEN COMBUSTION AND MANAGEMENT PROGRAM * .. FIGURE 1 SCALE 1982 11/1 1/1 - (1033FT ) ' 11/1 1/1 3/1 5/1- 7/1 9/1
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0.001 IGNITER DEVELOPMENT (WHITESHELL/ )
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0.6 HYDROGEN CONTROL STUDIES (ACUREX)- 0.2 HYDROGEN COMBUSTION STUDIES
' (AECI., WHITESHELL)
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30 HYDROG5N MIXING STUDIES (HEDL-W)
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Droin FIGURE 2 EXPERD! ENTAL ARRANGE! CIT FOR k'ATER FOG INERTING AND QUENCllDiG TESTS "-- - - -
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Notes: I. Jet Location 1. Figure not to scale II. Jet Location 2. Configuration has same diam. IXI. Diffuse Location to lower comp. ht. ratio as plant crane wall diam, to camp. ht.. Figure 6 CSTF View for H Mixing Tests 2 . ., . l j . , APPENDIX A.2 FENWAL REPORT . [ In the second quarterly report we included the Fenwal Phase II final report.. TVA has been requested by the staff to revise the pressure , - data recorded in .the final report's tables from mm of Hg to pounds per square inch (psi). The attached tables have been included to comply with that request. Note that-this is an editorial change only. The
- data collected during the Phase II testing and reported in ti.ese tables has not- been changed.
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I I l _ . - . _ _ _ _ , _ . _ _. - _ _ _ .. . ~ _ _ .- , , . . . Page 95 - . , TABLE NO. 1 SERIES 2 - PART 1 ,' Test No. .2-1-1 2-1-2 2-1-3 2-1-4 2-1-5 2-1-6 2-1-7 2-1-8 2-1-9 2-1-10 Date 10/10 10/14 10/14 10/15 10/15 10/15 10/16 10/17 10/17 10/27 - %H 9 8 7 6 5 8. 6 10 6 6' 2 s - TV F 136 138 140 142 144 138- _142 212 212 212 V ft/seg 0 0 0 0 0 5 5 0 0/5 0/5-Baro lb/in a 14.8 14.7 14.7 14.8 14.8 14.8 14.8 14.8 14.8 14.7 ~ R.H. 36 76 35 80 43 34 60 74 55- 52 . T amb )F 56 41 55 44 51 49 67 65 65 50 Air lb/in a 16.2 16.3 16.4 16.5 16.6 16.4 16.5 10 3 11.2 11.2 H lb/in a 1.9 1.7 1.4 1.3 1.0 1.7 1.3 2.1 1 3' 13 2 HO lb/in a 2.6 2.7 2.8 30 3'. 2 2.7 30 8.3 - 8.3 83 2
- F 210 141 140 142 144 230 190- 250 - 210/225 212/225 T)
T F 175 130 135 142 144 183 152 242 200/220 210/247 2 T F 142 140 140 142 144 N.O. N.O. 240 N.O. 227/289'- 3 T F 960 165 N.O. 142 144 685 335 700 245 205/208 TIgn see 15.8 15.9 15.4 17 17 15 17 17 17/1.0*- 19/6.1*- Tp sec 2 6.6 5.4 5.3 11 3 4 9 9.6 .13/10 '1 9/4.8-AP lb/in 38 31 1.5 1.0 0.2 36 15 30 .75/2.7 0.2/3 2 H (P) % 9.2 8.8 9.0 8.0 6.4 9.6 6.8 17.9 11.5 6.1 N (P) % 60.G 09.9- 69.3 68.6 74.5 72.2 72 9 66.4 71.7 73 7 , 0 (P) % 21.9 21.9 21.8 21.8 22.6 21.6 19.0 16.9 17.9 19.3 H (A) % 0 33 4.5 6.2 5.1 0 3.6 0 92 6.1 - N (A) $ 78.5 75.8 74.7 71 9 75.0 82.9 75.6 85.4. 74.3 73 9 0 (A) 5 18.9 20.3 20.6 21.8 22.5 19.6 18.0 12.6' 17.8 18.3-N.O. - Not Obtained * - Timed from Fan Start ' E71126.03 _ * . n ., . l < ~Page 16 TABLE NO. 2 SERIES 2 - PART 2 -Test No. 2-2-1 2-2-2 2-2-3 Date 10/28' 10/29 10/30 TV F 80 80 160 Baro ~ lb/in a 14.7 14.7 14.9 R.H. 5 95 65 57 T amb F 34 34 37 H '- SCFM 4 4 4 2 H2 0" lb/ min 0 0 23 T 3 F 215 226 265 T F 120 130 190 2 .T F 193 198 240 3 . .T 4 F 318 330 370 Tign'" sec 65 100 84 ' Tp see 12 12 4 2 A P,,x lbs/in 6.1 7.8 10.1 H2 (A)- % -- 23.6 23.9 N2 (A) 5 -- 72.2 71.0 02(A) % -- 4.8 7.3 . ' Hydrogen Flow Rate anSteam Flow Rate '#' Approximate Time From Hydrogen Flow Start to First Ignition , , - , - - - . , - - , - . , .--,a., . _ _ _ _ . _ -_-___ _ ._____ _ - _____ ____ _ _____ ____ ______ __ _ _ _ _ _ _ _ _ _ -__ " . rage i#- ,. ,. TABLE NO. 3 SERIES 2 PART 3 Test No. 2-3-1 2-3-2 2-3-3 2-3-4 2-3-5 ~Date 10/23 10/31 10/31 10/31 11/3 %H 105 10% 6% N.A. 10% 2 TV ( F) 39 80 80 80 80 Baro lb/in a 14.9 14.7 14.6 14.6 14.9 " R.h. 5 45 50 34 50 50 T amb ( F). 39 47 48 50 40 H lb/in a 1.7 1.6 0.9 0 -- 1.7 2 H' (SCFM) 0 0 0 4 0 2 2 H2 0" (gal / min) 2 2 2 2 2 T F 125 135 80 135 120 3 2 T F 110 130 120 100 120 2 T F 40 N.O. 133 155 145 3 T F 665 650 407 505 360 4 Tign'" see 14.8 11.4 22.0 90 14.9 Tp .sec 50 .65 1.50 6.0 1.1 AP lb/in 60 50 32 31 42.5
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Ignitor Orientation Normal Norml Norm l Norml Rotated-H2 (P) 5 N.O. 6.7 N.O. N.O. 7.8 N2 (P) % N.O. 73 1 N.O. N.O. 73.5 02(P) % N.O. 19.4 N.O. N.O. 19.3 H2 (A) % N.O. .8 N.O. N.O. O N2 (A) % N.O. 79.4 N.O. N.O. 82 3 0 (A) % N.O. 16.6 N.O. N.O. 17.5 2' Hydrogen Flow Rate ! " Water Spray Flow Rate N.O. - Not Obtained - -_ , .,- . _ _ _ - _ _ _ _ _ ._ . _ . - .. .. ._ __ _ .= - ... . Page 18 , TABLE NO. 4 SERIES 2 - PART 4 Date 11/12 11/13 11/14 11/17- 11/16 11/07 11/18 Test No. 2-4-1 2-4-2 2-4-3 2-4-4 2-4-5 2-4-6 2-4-7 $H 2 125 105 10% 125, 125 , 12%, 12%, TV ,F 129 129. 129 129, 146, 146, 129 Baro lb/in a 14.6 14.7 14.6 14.9 14.7 14.6 14.5 R.H. $ 558 428 30% 573- 608 55g 933 T amb F 40 37 55 29 39 65 26 2 16.0 Air lb/in a 16.0' 16.0 16.0 16.0 16.3 16.3 2 2.4 Hg 1b/Inga 2.4 2.4 2.4 2.4 2.1 2.1-M0 lb/in a 2.2 2.2 2.2 2.2 2.8 2.7 2.2 2 Tj F -- -- -- -- 380 510 -- T F 255 395 365 395 432 510 357 2 T F -- -- -- -- 202 195 -- 3 T F 710 760 760 755 790 760 735 4 140 --- 130 T F 140 -- -- -- 5 T F 150 155 - -- -- -- 140 6 T F 135 140 -- -- -- -- 133 7 T F 230 250 - -- -- -- 143 8 Tg F -- -- 240 250 - -- -- Tg F -- -- 170 138 -- -- -- T F -- -- 240 250 - -- -- 33 T F -- -- 228 183 -- -- -- 12
! _ _ _ _ _ _ _ _ _ _ _ _ _ - - _ _ _ _ _ _ _ _ - - _ _ - _ _ .. - - , , . .. .;. 4 b *- a, m ' Page 20 Legend For Table No. 1, No. 2, No. 3, and no. 4 % ll - Hydrogen Test Concentration (%)l 2 ' TV' - Vessel Test Temperature (*F) V- - Air Velocity At Glow Plug (ft/sec) Baro. - Barometric Pressure (lb/in a) , .R.H. ' - Relative Humidity (%) T'amb - Ambient Temptiature ( F) P air - Partial Pressure (lb/in a) of Air Loaded. PH - Partial Pressure (lb/in 2 a) of Hydrogen Loaded 2 - Partial Pressure (lb/in 2,) of Steam Loaded - P 11 0 . 2 T - - Glow Plug Box External Wall Maximum Temperature ( F) 3 T - Vessel Internal Wall mximum Temperature ( F) 2 T - G1 w Plug Box Internal Gas Maximum Temperature ( F) 3 T - Vessel Air Maximum Temperature ( F) 4 T 5 - Barton Transmitter 2& 4 T 6 - Barton Transmitter 4 ( F) T - Barton Transmitter- 5 ( F) 7 Tg - Barton Transmitter - Outside Surface Maximum Temperature ( F) .T g - Limit Switch - Outside Surface Maximum Temperature ( F) Tg - Limit Switch'- Internal Maximum Temperature ( F) Tg - Solenoid Valve - Outside Surface Maximum Temperature ( F) .T 12 - S len id Valve - Internal Nximum Temperature ( F) Volts _ - Voltage at Glow Plug (VAC) . INS. - Insulating Wrap Type , h,. , , . . - . . y _ . . , , - e.,.. ,w , e o . , A?PENDIX A.3 SINGLETON TESTING Several high temperature tests have been conducted at TVA's Singleton Laboratories on the incore thermocouple cables, the hot and cold leg RTD cables, and the power cables for the igniters located in the upper plenum of the ice condenser. The purpose of these tests were to demonstrate the survivability of the cables should they be subjected to the temperatures produced, in the containment, as a result of burning hydrogen. Figure A.3-1 is the temperature profile to which the thermocouple and RTD cables were exposed. Figure A.3-2 is the temperature prorile of the oven in which the igniter cables were tested. A description of the tests and the results were reported in our " Resolution of Equipment Survivability Issues for the Sequoyah Nuclear Plant" document dated June 1,1981, i a l t , , , w rw y p- w ---+- w - r- - r + - - --s -e'-
n . . THERMOCOUPLE AND RTD CABLE TEST PROFILE . .. , .. 30 SEC p [ 4_ 1400- _3 l- <C O' ' UJ Q_ s 300-h 170 SEC __ AMBIENT -> 60 k!N 4- - . ,. -> 60 MIN 4-630 MIN -> 4 TIME FlGURE A.3-1 . & (J . s O - . e . E L I F O R P T S M U 2-E M 3 T A I N I E R M U S E M E T A E T I T R U _ R U E N P M I M S E G I E T T 5 4 U N F N - I E M V 0 O 1 . E L . z B 4 A C R E T I N G I ] ,_ T N E I B M A - 0 0 - 7 _ _ _ _ t Y?0W3-E . _ _ _ ~ _ _ . S - . a a APPENDIX B EQUIPMENT SURVIVABILITY On June 2, 1981, TVA submitted a report cn equipment survivability to satisfy a Sequoyah Operating License condition to provide a more detailed justification of TVA's position that key equipment will survive a repeated hydrogen burn envircnment. We would, therefore, reference that submittal for our discussion on this subject. ' - , ., .., . APPENDIX C ' PERMAliENT HYDROGEN MITIGATION SYSTEM (PHMS) . DECISION METHODOLOGY From the Sequoych Nuclear Plant Unit 1 operatir,g license: (22-D-2) For operation of the facility beyond January 31, 1982, i the Coomission must. confirm that an adequate hydrogen control system for the plant is installed and will perform its intended function in a manner that provides adequate safety margins. To comply with this condition, TVA is taking steps to select, design, and install a Permanent Hydrogen Mitigation System (PHMS) for Sequoyah unit 1. TVA will .aubmit by June 30, 1981, a report to the Commission providing preliminary information on the Permanent System we have selected. ' 8 I d 4 4 4 d ,}}