ML20214W458
| ML20214W458 | |
| Person / Time | |
|---|---|
| Issue date: | 12/30/1983 |
| From: | Ross D NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Reed C COMMONWEALTH EDISON CO. |
| Shared Package | |
| ML20213E209 | List:
|
| References | |
| FOIA-87-113, FOIA-87-60 NUDOCS 8706160030 | |
| Download: ML20214W458 (8) | |
Text
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uNitta svaras NUCLEAR REGULATORY COMMissi!N SARF GROUP WASHINGTON, D. C. 20688 RJMattson 5
December 30, 1983 TSpeis
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RVo11mer RBernero 0Bassett GArlotto JLarkins Mr. Cordell Reed DFRoss Vice President Commonwealth Edison Company 1 First National Plaza Chicago, IL 60690
Dear Mr. Reed:
As a consequence of the NRC-IDCOR management meeting of December 19, 1983, I am confirming the following understandings:
The general instructions to the NRC staff with respect to 1.
interactions with IDCOR are governed by a memo from W. J.
Dircks dated December 29,1983 (Attachment A).
The general flow sheet of NRC-IDCOR interactions are derived 2.
at our meeting, and is as shown on Figure 1.
The issue papers from the NRC family (staff and/or contractors) 3.
will have the following properties Those issues directly related to the source term a.
characte.-ization will be written and defended by NRC contractors. The remainder will be done by NRC staff, as augmented by contractors.
For convenience Table 1 shows the issues, and assigns one of these two categories.
b.
All issue papers will be considered draft or i
interim until they are fully integrated.
Further, the source-term subset of issue papers will be considered as tentative until the APS peer review is in hand.
The purpose of the issue papers is to provide c.
focus for the NRC-IDCOR interactions, so that we j
can concentrate on areas needing resolution.
4.
We set meeting schedules as follows:
l Technical meeting #2 on fission product transport:
a.
i February 7-8,1984 l
8706160030 870610 h$LL 60 PDR B/34,
Figure 1 IDCOR-NRC INTERACTIONS-DISPOSITION REDRAFT I
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TO BE RESOLVED LATER ll
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TABLE 1 SEVERE ACCIDENT ISSUES SPONSOR 1.0 Severe Accident Phenomenology NRC 1.1 Progression of core melt in the reactor coolant system Reactor coolant system thermal and hydraulic behavior 1.1.1 1.1.2 Rate and magnitude of hydrocen production in the vessel and release from reactor coolant system 1.1.3 Fuel debris and in-vessel structure interaction 1.1.4 Fuel debris and vessel or vessel penetration interaction 1.1.5 Likelihood and magnitude of in-vessel steam explosions 1.1.6 Recovery potential prior to vessel failure 1.1.7 Primary system failure from overpressure 1.2 Loading of the containment 1.2.1 - Containment thermal and hydraulic behavior 1.2.2 Rate and magnitude of combustible gas production, ex-vessel Distribution of combustible cases and conditions leading 1.2.3 to and resulting from detonations 3
Conditions leading to and resultino from deflacrations, 1.2.4 diffusion flames and flame acceleration 1.2.5 Likelihood and magnitude of ex-vessel steam explosions or steam spikes 1.2.6 Debris coolability in ex-vessel locations 1.2 7 Debris relocation following vessel failure Fuel debris - containment shell, floor and internal 1.2.8 structure interactions 1.2.9 Rate and magnitude of non-condensible gas production
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NLC ex-vessel
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l SPONSOR _
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.NRC Response of the containment and other essential equipment 1
1.3 1.3.1 Characteristics and likelihood of containment leakaae
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from shock loadings Characteristics and likelihood of containment leakage 1.3.2 resulting from steam spikes and/or hydrocen burning Characteristics and likelihood of containment leakage 1.3.3 resulting from slow pressurization 1.3.4 Characteris, tics and likelihood of containment leakaae resulting from external events l
Characteristics and likelihood of containment leakaoe 1.3.5 resulting from thennal loading Characteristics and likelihood of containment leakage 1.3.6 resulting from internal missiles 1.3.7 Potential for basemat penetration 1.3.8 Reliability of early containment isolation NRC 1.3.9 Equipment and instrumentation survivability Contractor 1.4 Fission product release and transport Rate and magnitude of release of fission products from 1.4.1 fuel Deposition of fission products during in-vessel transport 1.4.2 Rate and maanitude of release of radionuclides from fuel 1.4.3 (ex-vessel) 1.4.4 Deposition of fission products in containment due to natural processes Effect of engineered safety features on fission product 1.4.5 retention contractor 1.4.6 Deposition of fission products in other plant buildinas NRC 1.5 Ex-co.ntainment transport and consequences 1.5.1 Environmental dispersion 1.5.2 Food chain transport O
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1 SPONSOR _
1.5.3 Dosimetry and health effects 1.5.4 Modeling of emergency response 1.5.5 Cost Analysis NRC 2.0 Safety Assessment 2.1 Characterization of plants and sequences 2.1.1 Plant categorization 2.1.2 Identif-cation of accident sequences 2.1.3 Quantification of sequence likelihood 2.1.4 Equipment performance and success criteria 2.1.5 Influence of operator action on accident sequence 2.2 Assessment of existing plants 2.2.1 Response of reference plants to selected severe accidents 2.2.2 Qualitative assessment of severe accident likelihood 2.2.3 Integrated probabilistic risk assessments for reference Contractor plants 1
NRC 2.2.4 Credibility of PRA techniques l
2.2.5 Applicability of conclusions concerning existing plants 2.2.6 Effects of uncertainties and sensitivities on the estimated severe accident consequences 2.2.7 Effects of uncertainties and sensitivities on estimates of severe accident likelihoods 2.2.8 Sabotage l
2.3 Assessment of plants with modifications 2.3.1 Designaid operation changes to prevent accidents including sabotage protection 2.3.2 Improvement in severe accident management capability NRC 1
2.3.3 Design changes to mitigate accidents Contractor i
l 2.3.4 Changes in emergency response capability NRC
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NRC
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3.0 Decision Methodology 3.1.1 Selecti~en of decision techniques Selection of attributes to be used in cost-benefit 3.1.2 analysis Selection of weights for cost-benefit analysis 3.1.3 3.1.4 Role of safety goals 3.1.5 Balance of decision approaches Balance of preventive and mitigative measures 3.1.6 m
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Mr. Cord 211 Reed
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Technical Meeting #3 on Integrated Assessment:
b.
.v.
April 10-13,1984 Technical Meeting #4 on Design, Operations, and Risk:
c.
May 30-31,1984 d.
NRC-IDCOR Management Meeting:
January 24, 1984 The first set of issue papers, about eight in number (on the general category of melt progression) will be sent to you by January 1,1984.
5.
These (we understand) will be discussed at your next steering group meeting, and thence at the IDCOR-NRC management meeting (see 4-d We expect to transmit the BCL version of some fission above).
product issue papers sometime in January, in advance of Technical Meeting #2 (item 4-a, above).
As noted in Figure 1, we will be sending issue papers (drafts and 6.
redrafts) to the ACRS.
We understand that your IDCOR Technical Summary will not be available 7.
until mid-March 1984.
The flow sheet for Source Term Work is shown on Figure 2.
8.
If you have coments on this record of meeting, please write.
Sincerely.-
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/1 N y D. F. Ross, Deputy Director i
Office of Nuclear Regulatory Research l
Enclosure:
As stated I.
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