ML20235Q974
| ML20235Q974 | |
| Person / Time | |
|---|---|
| Issue date: | 10/01/1987 |
| From: | King T NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Beckjord E, Morris B, Speis T NRC |
| References | |
| PROJECT-672A NUDOCS 8710070755 | |
| Download: ML20235Q974 (9) | |
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UNITED STATES g
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NUCLEAR REGULATORY COMMISSION g
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00T 01 1987 L.
P MEMORANDUM FOR:
Distribution FROM:
Thomas L. King, Acting Chief Advanced Reactor & Generic Issues Branch Division of Regulatory Applications Office of Nuclear Regulatory Research
SUBJECT:
MEETING WITH DOE TO REVIEW STATUS AND KEY ISSUES /0 PEN ITEMS RESULTING FROM THE MHTGR REVIEW The subject meeting will be held on October 16, 1987 in Room P-ll8 and will begin at 9:00 a.m.
An agenda and draft connents are attached. The topics
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will include a status of the review, comments on reactor physics, performance l
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of the reactor cavity cooling systems, accident analysis, selected other items, and an overview of NRC criteria being developed for addressing key issues.
Reviewers and presenters for these topics, together with contractors from ORNL and BNL, should plan to meet in my office at 1:00 p.m on October 15, 1987.
If you have any questions please contact Pete Williams, the Project Manager (x29613).
Thomas L. King, Acting Chief Advanced Reactor & Generic Issues Branch Division of Regulatory Applications Office Nuclear Regulatory Research
Enclosure:
- 1. Agenda
- 2. Draft Comments j'l/
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AGENDA I:
DOE /NRC Meeting.on MHTGR PSID Review (Project 672)
October 16, 1987-Room P-118
Purpose:
To discuss review status and key issues /open items resulting from MHTGR-PSID review.
Agenda:
'9:00' Introduction and status of review T. King 9:30 Review of specific items for PSID Chapters:
- Chapter 4 (comments 4-41 thru 4-44 attached)
P. Williams /
D. Moses
- Chapter 5 (comments 5-37 thru 5-42 attached)
P. Williams /
J. O'Brien/
S. Ball 12:00 - 1:00 Lunch 1:00 Chapter 15 (corrnents 15-4 thru 15-7 attached)
P. Williams /
P. Kroeger 3:00 Other areas to be addressed at a future meeting P. Williams /
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T. King 3:30 Overview of NRC criteria being developed for T. King addressing key issues:
- adequacy of containment
- treatment of severe accidents
- emergency planning
- source term 4:00 Ajourn 1
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DOE /NRC Meeting, 10/16/87 Draft Agenda Comments 4-41 We have reviewed the responses to our comments on reactor physics concerns presented in PSID Vol. 5 (as categorized under "neutronics" on page R 4-iii) and in DOE-HTGR-87-085, "MHTGR Core Nuclear Uncertainties". We find the information presented acceptable in support of the MHTGR reactor conceputal design and for use in the transient and accident analyses at the conceptual design stage, but take the position that this cceeptability can not be extended to more advanced stages of design without substantial improvement of the data base. Data base improvements are needed in recognition of modern standards of accuracy in experimental techniques, the uniqueness of the inner reflector geometry, scarcity of experimental
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work with LEV fuel, and the growth of the plutonium fraction with burnup.
In DOE's Response 4-15, DOE " committed to validate MHTGR nuclear physics codes consistent with NRC regulations and industrial standards relevant to the MHTGR as the design development proceeds.
This long-term commitment will take considerable time to complete and will be developed during preliminary design."
In view of our finding that the reactor physics data base will need improvement and the commitment of Response 4-15, DOE should amend and augment its Response 4-20 with regard to inclusion of reactor physics as a new chapter in the Regulatory Technology Development Plan (RTDP).
In this response DOE should describe what reactor physics data could become available from cooperative programs with West Germany and Japan and whether or not DOE believes that such programs will be sufficient in themselves to provide the necessary improvements in the reactor physics data base. For the RTDP chapter, the reactor physics plan should be described with respect to background, objectives, approaches and acceptance criteria as currently available and should be expanded later as new information becomes available.
4-42 In studies of ccnduction cooldown with and without RCCS availability, performance results in terms of peak fuel and vessel temperatures were seen to be highly sensitive to long term values of decay heat.
In Response 4-18 it was stated that more recent and better qualified data were being evaluated in comparison to the " original" PSID data.
Describe progress being made in this area and indicate the degree and effects of uncertainties in more recent data.
In this description-indicate the approaches being considered including collection and analysis of existing. material and possible experimental validation.
4-43 In Response 4-25 it was stated that if flux mapping detectors should fail, plant operation could continue and that no ISI was planned for these detectors. These detectors monitor the core for long term burnup effects and assure that undesirable fuel temperatures do not occur in the lower core regions. Therefore, our position is to classify these detectors as "impcrtant to safety" with the intent that they be built to quality standards, receive periodic testing and cali-bration and have appropriate Technical Specification requirements to monitor flux levels and to assure their availability and performance.
At a later design stage, the details of the safety standards that need to be met could be developed.
In a like manner we require the same treatment for the startup monitors discussed in Response 4-26.
The need for these safety standards is based on operator safety during
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refueling.
4-44 Discuss the data base supporting the values used for thermal conductivities in the core and reflector graphites. This discussion should include the thermal annealing effects on the conductivities of irradiated graphites.
5-37 In our independent calculations of RCCS performance under pressurized conduction cooldown we encountered two major concerns. The first is that DOE does not model downward by-pass flows in the region external to the core but considers that all convection cells are confined to the
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ccre region itself. The second is that the upper plenum thermal protection. structure may contain sufficient insulation to cause more p
heat than intended to be shifted to the vessel side wall from the vessel head, resulting in a potentially unacceptable temperature peaking at wall locations. Discuss these concerns and their effects on your conclusions regarding vessel integrity and RCCS performance.
Also, discuss if vessel integrity could be better guaranteed and uncertainties reduced if the system is depressurized before the vessel reaches elevated temperatures.
5-38 Provide sensitivity calculations to illustrate parametrically the RCCS performance with respect to peak and average fuel temperature and peak and average vessel temperature considering the following uncertainties:
(1) graphite conductivity, including radiation annealing effects, (2) surface contact and gap resistances between adjacent graphite blocks and between graphite blocks and the inner surface of the core barrel, (3) effects of convection flows in the core and reflector, (4) emissivities on the core barrel inner and outer surface, (5) effects of helium convection, seismic keys and helium ducts on the heat transfer across the core barrel, (6) the emissivity on the inner and outer surfaces of the reactor vessel, (7) effects of convection j
flcws exterior to the reactor vessel, (8) emissivity of the RCCS
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panel surface, (9) geometrical and asymmetrical effects, (10) influence l
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of the upper plenum thermal protection structure and other reactor internals that could cause shifting of thermal gradients or otherwise cause temperature or stress peaks, and (11) any other modeling practices or assumptions that could affect the NRC assessment of RCCS performance.
In performing this study, parameters that can be shown by simple analysis or a confirmed data base to have a trivial effect can be omitted. The overall goal of this study is to demonstrate that the RCCS can meet its performance requirements, including potentials for hot spots and unacceptable thermal stress locations with acceptable margins and to identify any programs that should be in the Regulatory Technology Development Plan to reduce uncertainties.
i 5-39 A very low value.for the' seismic failure probability of the RCCS is given in the PRA (Vol. 2). From our previous experience with structural support systems and the subsurface. location of most of the RCCScomponents,webelievethatsuchalowfailureprobabilitycould be achieved, reviewed and accepted. The design rules would use NRC approved structrual codes with input determined from Standard Review Plan Sections 2.5, 3.7.1, 3.7.2, 3.8.1 and 3.8.2.
Inservice inspection would have to be.very thorough and perhaps some disassembly of components would be required. The current status of the RCCS design activities regarding seismic integrity should be discussed in lignt of the above.
1 5-40 During an event in which only the RCCS is used for decay heat removal, the reactor vessel could be exposed to temperatures that possibly approach or even exceed code allowables. Therefore, for reasons of accident progression monitoring, decisions regarding depressurization and to assess the reuse capability of the vessel, it is the staff's position that safety grade instrumentation suitable to determine vessel temperature history during a conduction cooldown event be provided.
15-4 We have reviewed material provided in Appendix G of the PRA study that addresses complete failure of the RCCS. For this severe event, DOE
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should provide a sensitivity study similar to that requested in Comment 5-38 for the RCCS. This study should be augmented to include such modeling factors as the resistance of the RCCS panel itself, convection currents in front and behind the panel, emissivities of both sides of the panel and the cavity surface, methods for modeling cavity and vessel geometry in two and three dimensions, time dependent consequences of failure to depressurize the vessel, and soil I
conditions external to the reactor cavity.
15-5 The Appendix G calculations for RCCS failure are made on the basis that there are no structural failures other than the RCCS or the cross duct. Other failures that coincide or follow these failures that j
might result in further elevation of fuel or vessel temperatures j
should be discussed qualitatively, and quantitatively if a significant t
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increase in off-site doses is predicted. These might include local or gross vessel failures from over temperature and changes in heat L
transport geometry within and exterior to the vessel as might be caused by earthquake, after shock, and edditional over heating.
Furthermore, concrete failure, particularly above the vessel, either from earthquake or over heating, should be considered from the stand-point of causing additional structural failures, combustible gas generation, or significant chances in the heat transport mechanisms in the reactor cavity.
Also, DOE should indicate whether it plans to commit to design the reactor cavity, vessel support, and any other critical structural items to the same seismic standards as the RCCS itself.
15-6 In Appendix G, consequences of the various severe events explored are presented in reference to the silicon carbide coating degradation temperature of 2000 C.
Explain why this parameter was chosen and describe the safety consequences, including time dependent off-site doses, if this limit is exceeded.
15-7 As a result of our review of Appendix G of the PRA, we have determined that it might be possible to develop a mechanistic basis for the site suitability source term rather than following the customary approach of postulating a non-mechanistic source term that can be prudently
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considered to bound all conceivable credible accident mechanisms.
In order for us to investigate this possibility and in accordance with Response G-19, we believe it would be useful to construct postulated accident sequences that explore bounding events to provide a highly conservative, as opposed to best estimate, time-history calculation of off-site doses that could result from the types of severe events presented in Appendix G.
For example, the cases of RCCS failure and cross duct failure with and without seismic initiation appear to be sufficiently conservative accident sequences at the present state of the MHTGR design for these purposes, provided that conservative modeling as explored in Comments 15-4 and 15-5 is used and that sufficiently conservative models are to be used for fuel failure and fission product transport.
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Development of a site suitability source term by consideration.of conservative off-site doses in terms of a time-history reflects the f,
-unique features of the fuel and the MHTGR design. We believe this approach might become an acceptable means for establishing a SSST for the MHTGR if performed conservatively. Evolution of radioactive accident releases as a function of time would thus becomes an important factor in decisionmaking in this regard and for determining containment and off-site emergency planning requirements.
P
E-OCT 01 1987 j
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' DISTRIBUTION 4'
'RES Cire Chron-1 ARGIB~R/F E. Beckjord e
T. Speis.
B. Morris-
- 2. Rosztoczy T. King J. N. Wilson C. Allen R. Landry P. Williams J. Flack M. Dey.
.j B.'Hardin-l
- 0. Gormley R. Baer N. Anderson F. Cherny S. Shaukat i
R. Johnson I
D. Thatcher J. Hulman J. Glynn L. Soffer l
J. Read D. Cleary l
A. Murphy G. Arndt i
R. Kirkwood j
E. Podolak
]
R. Erickson B. Mendleschn J
H. VanderMolen I
E..Chelliah L. Beltracchi F. Congel C. Hinson O. Lynch D. Matthews J. O' Brian R. Senseney M. Spangler F. Coffman S. Ball, ORNL P. Kroeger BNL G. VanTuyle, BNL 1
R. Ireland, Reg. IV M. El-Zeftawy, ACRS/H-1026
,PDR - Project 672 CProject-Filej672L(Centra,1 Files)
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