ML20136F998

Forwards Statement of Work Re Tentative Work Assignment Aj Re Analysis of Steam Explosion Uncertainties

ML20136F998
Person / Time
Site:	Indian Point
Issue date:	04/16/1985
From:	Carrington M Office of Nuclear Reactor Regulation
To:	Pandey S CALSPAN CORP.
Shared Package
ML20136F845	List: IA-85-448, Forwards Addl Info Re Steam Explosions.Plan to Resolve Steam Explosion Issue in 3 Yrs or Less Has Been Prepared.Plan Will Be Carried Out If Budget Adequate ML20136F841 ML20136F992 ML20136F998 ML20136G101 ML20136G107 ML20136G113 ML20137D259 ML20137D295
References
CON-FIN-B-6891, CON-NRC-03-81-130, CON-NRC-3-81-130, FOIA-85-448 NUDOCS 8505230755
Download: ML20136F998 (7)

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APR 16 1985 Dr. S. Pandey Franklin Research Center ARVIN/CALSPAN 20th and Race Streets Philadelphia, Pennsylvania 19103

SUBJECT:

CONTRACT NRC-03-81-130 TENTATIVE ASSIGNMENT "AJ", ANALYSIS .

OF STEAM EXPLOSION UNCERTAINTIES, INDIAN POINT 2 l In accordance with estab'lished procedures for the contract, I am fcrwarding for your review a work statement for the subject tentative assignment.

Please provide your estimate of the resources required to perform the work, and tell us whether you can meet the planned schedule. Because of the urgency for beginning the project, your reply is requested within five '

working days after your receipt of this letter.

The cos't of this work will be accounted for und.er FIN H-6891, Reactors. In addition to myself, the following NRC personnel wOperating ill be associated with this assignment, with responsibilities as indicated.

Lead Engineer Cardis Allen, 492-7347 Performance Monitor Thomas King, 492-7347 Sincerely, f

Mayo Carrington Technical Assistance Program

• Management Group

Enclosure:

As stated i

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( TENTATIVE WORK ASSIGNMENT "AJ" I

ANALYSIS OF STEAM EXPLOSION UNCERTAINTIES INDIAN POINT 2 NUCLEAR STATION STATEMENT OF WORK A. Background In the unlikely event of a major core melt accident in a light water reactor, the molten core penetrates the lower core plant, diffuser plate, and lower support plate, and into the residual water in the lower plenum of the reactor pressure vessel. The potential for instantaneous thermal mixing of the hot molten fuel and the water, and a subsequent steam explosion, therefore exists. Many analytical and experimental studies have been carried out in the past to simulate the expected behavior of the core under degraded conditions and fuel-coolant interaction. Nevertheless, use of this body of information to predict the likelihood of steam explosiens and the potential for subsequent containment failures will result in a number of uncertainties ranging from understanding of the

-mechanisms involved to the non-prototypical nature of the experiments. A detailed, systematic and traceable probabilistic approach will not only estimate the uncertainties involved in steam explosions phenomenology, but will also display the na'ture and extent to which the present lack of knowledge contributes to them.

Under a recent NRC sponsored study regarding the " Incorporation of Phenomenological Uncertainties in Safety Studies - Application to LMFBR Core Disruptive Accidents," by B. Najafi, T. G. Theofanous, E. T. Rumble, B. Atefi, a computer-based probabilistic approach was developed to assess the uncertainties in LMFBR core disruptive accident energetics.

B. Objective The objective of this assignment is to assist the NRC in the assessment of the likelihood of containment failure as a result of steam explosions at the Indian Point 2 Nuclear Station.

. ( C. Work Requirements General Description of Required Work The required work will evaluate the frequency distribution of containment failure. at Indian Point as a result of steam explosions. The work will also identify and rank the major sources of uncertainties contributing to the uncer-tainty in this mode of containment failure.

The phenomenological path of. accident progression will be delineated and a detailed model of the accident will be developed keeping in mind the dependen-cies which exist among the various contributing phenomena.

The computer program developed under " Incorporation of Phenomenological Uncer-tainties in Safety Studies - Application to LMFBR Core Disruptive Accidents" study will be expanded by adding the capabilities r'equired by the steam explo-sion models such as non-linear dependencies among the contributing phenomena.

The program will then combine the distribution of uncertainties in those phenomena to obtain frequenncy distributions of the containment failure as a result of steam explosions.

The computer program developed will also be capable of ranking the contribut-ing phentynena based on their impact on the uncertainty of the containment failure.

Results of this activity will include a documented description of the following elements:

1. Development of a computer-based approach for probabilistic analysis of steam explosions and their inherent uncertainties.

2. Estimation of the frequency distribution of containment failure at Indian Pnint as a result of steam explosions.

3. Discussion and ranking of the associated uncertainties and the key sensitivities impacting the uncertainty in the estimated occurrence frequencies.

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\ C. Work Requirements (continued)

Task Descriptions The contractor shall perform the following specific tasks in his accomplishment of the work of this assignment.

Task 1: Development of ' Indian Point Model a) Review and examine preliminary model for needed enhance-ment and completeness including use of supporting phenom-enological analysis and calculations.

b) Refine model as a result of Task 3a.

Task 2: Development of the Inout Parameter Distributions a )' Review the preliminary set of input distributions.

b) Review and improve the input distributions based on the results of Task 3a. Develop detailed supporting documen-tation.

Task 3: Quantification a) Preliminary execution to look for important parameters and elements of the model. Feedback to Tasks Ib and 2b.

b) Final quantification for the frequency distributions of the containment failure given core melt at Indian Point.

Identify major contributing factors.

Task 4: Uncertainty - Importance Rank input physical parameters in terms of their contribution to the uncertainty of the final result.

Task 5: Assessment of Model A peer. revew shall be performed for two of the technical tasks described above. The review will focus on assessment of the probabilistic methodology (Task 1), and quantification procedures (Task 3) used in the analysis.

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D. Meetings and Travel, It is expected that the following meetings and travel will be needed:

1. Two one-day meetings at NRC offices, Bethesda, Md.

2. One two-day meeting on the west coast.

E. Plants Affected and Schedule

1. This ass'.ignment is concerned with the Indian Point 2 Nuclear Station,

.0ccket No. 50-247.

2. This ass'ignment is to be completed within six months after work is authorized. A schedule for accomplishment of the specific tasks will be developed by the contractor within two weeks after start, and approved by the NRC-Lead engineer and the NRC Contract Project Manager.

F. Reporting Requirements

1. 'An interim technical report shall be provided at about the three month point of the performance period. Work and results to that time shall be described.

2. A fin ~al technical evaluation report shall be provided, describing the details of the work, findings and conclusions of the effort accomplished in accordance with Paragraph C. A draft of the final report shall be provided at least two weeks prior to the due date (six months after work authorizatiori) for the final TER. The final report shall incorp-orate NRC comments.

3. Cost and progress reporting for this assignment will be provided in the Monthly Progress Report in accordance with the requirements of the basic contract.

Enclosure 6 REASSESSMENT OF TMI-2 CORE DAMAGE TMI Program Office

. George Kalman A closed circuit television inspection of the THI-2 lower reactor vessel head revealed that approximately 10-20 tons of debris had accumulated in the void between the lower flow distributor and the lower vessel head. The physical size and quantity of the debris which included indications of liquified flow on the lower head refuted previous core damage assessments which maintained that the bottom of the core remained structurally intact. The visual evidence of more severe-damage was, corroborated by the analysis of samples which had been retrieved from the upper core void. These core debris samples contained evidence of previously molten UO , indicating that temperatures in localized ares of the core exceeded 5100*F 2 Prompted by. these data, GPU requested a reassessment of the accident scenario and core damage. Independent reassessments were undertaken by representatives from Rockwell, EPRI, and EG&G. Although specific details varied somewhat, all three assessment ' conclusions were similar. A crust of previously molten material formed at the water level approximately 3 - 4 feet above the bottom of the core (debris probing experiments support this hypothesis). Core cooling capacity below the crust was reduced and three hours, 43 minutes after the start of the transient, molten core material broke through the stainless steel' supporting structure and flowed to the bottom of the vessel. At this time, previously unexplained spikes in primary system pressure, neutron flux, temperature fluctuations, and failure of numerous incore sensors, all support the new core relocation hypothesis. The molten core attained a coolable geometry in the lower reactor vessel and resolidified. . A void or voids were left under the upper debris crust in the area vacated by the molten core material (see sketch).

Analyses of TMI debris samples and results of Power Burst Facility (PBF) testing indicate that the chemical makeup of resolidified core material can vary significantly. Liquified uranium resulting from UO,/Zry interactions can occur at temperatures as low as 2060*F. Eutectic reactions at 3450*F produce a metallic resolidified material, while reactions at 4700*F produce ceramic i

solids. A combined product of the two reactions can result in a material with both metallic and ceramic properties. A resolidified mass of PBF debris was very hard and nonbrittle. During the TMI accident, conditions existed which could have formed.the hard, nonbrittle material in the upper crust and in the resolidified debris in the lower vessel.

t Existing defueling concepts, with some tools already in final stages of i design, are geared to handle loose rubble, partially intact fuel assemblies, and loosely adhering resolidified material. The core damage reassessments i

have raised concerns over the adequacy of existing defueling plans. Jt is recognized that additional data are needed to develop more definitive

, defueling criteria. Access to the core and the lower reactor vessel will be simplified this sumer after the plenum assembly is removed. Additional video j inspectiu.n. core bore sampling, and lower head debris sample retrieval will be attempted iollowing plenum removal. ,

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