ML19323E566
| ML19323E566 | |
| Person / Time | |
|---|---|
| Site: | Indian Point, Zion, Crane  |
| Issue date: | 03/25/1980 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | Dougherty D NRC OFFICE OF ADMINISTRATION (ADM) |
| References | |
| RS-NRR-101, NUDOCS 8005230783 | |
| Download: ML19323E566 (7) | |
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-- - v M M 2 i 1580 fEDRANI13! FOR: Dennis J. Dougherty, Chief, Technical Assistance Centracts Branch FFCI:
Derrsood F. Ross, Acting Director, Division of Project Management SU3JFLT:
RS-NRR-80-101
" ADVANCED REACIOR ACCIDEhT DELINEATION AND ASSESSET" (SECY-80-67)
I am forwarding to you a revised Statement of Work for the subject technical assistance (TA) contract. The revised statement now deals exclusively with TA work associated with severe accident mitigation for LWRs, with the focus on the Zion / Indian Point plants (See draft 3 of BII-2 Action Plan, Su: tion II.B.6). Eis revision has been coordinated with Mr. William Menc er of your staff. Three points should be mde concerning this revisica.
First, it meets the two objections raised by the Comraission in their letter to William Dircks of bhrch 10, 1980, namely, by directing the work tholely to LKRs (as opposed to the initial direction to fast breeder reactors as well as to water reactors) and placing appropriate emphasis on the licensing nature of the work (as opposed to what appeared to the Comission to be a "research" orientation in the initial wording).
Second, even though the end product in tems of types of reactors is somewhat different, the scope of the work is similar to the initial scope because the subject matter is generic and applicable to a variety of reactor types. Le origir.a1 Statencat of Work had three Tasks (Task IA, Task IB, and Task II). Task IA (Heavy Water Reactors) has been dropped in the revision but was never intended to be a major portion of the total effort. Task IB (Light Water Reactors) is expanded somewhat and directed specifically to problaas associated with the present Zion / Indian Point action. In the revision this task (original Task IB) becomes Tasks II and III. The original Task II (Fast Breeder Reactors) is very similar to the revised Task I but directed to core melt problems associated with Zion / Indian Point as opposed to the original direction to Fast Breeder Reactors. Basically, the content of the effort remains the same; the specific direction to the Zion / Indian Point Action is what is different. This re-e=phasis is apparent in the "Bachground" statement.
pj Eird, the duration of this contract has been changed to 18 nonths l
TfiE 36 months) because the Zion / Indian Point Action program conducted i
by the Corrdssion is expected to continue for about 18 months. Because of
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D. J. Dougherty "'2: C the nature of the contract program we view the accomplishments to be alcut half those initially anticipated in the 26 =cnths contract.
Specific Zion / Indian Point milestones for this contract will have to, of course, be developed (the overall Z/IP plan should be completed shortly).
I wish to eghasize the urgency of implementing this contract.
We view this work to be an integral part of the Z/IP cffort within NRR. The Cc:=ission is comitted to completing the study of features to reduca the probability or mitigate the consecuences of severo accidents for the Zion Units 1 and 2 and the Indian Point Units 2 and 3 en a very tight schedule. Interim actions inposed on Z/IP are just that; the impicmentation of long-tem pcmanent actions are key for continued operation of these units. We anticipate a continuation of the fine cooperation your office has given us in expediting this matter, in particular, through the efforts of your staff member, Mr. William Menc:er.
If there is any way we can be of assistance, do not hesitate calling me or Dr. James Meyer of my staff on X-27976, oriciesT aftced 4 n 4 1;*a.,, -
Denwood F. Ross, Acting Director cc: William Menc:er Division of Project Fhnagement
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. s a result of the TMI-2 accident, the NRR staff has recognized the n?ed to re-examine the enrgency preparedness plans and capabilities of all nuclear pcuer plants.. Although programs are under.;ay that will evaluate all nuclear plant sites, two sites have been singled out for additional evaluation at this time.
The sites are those for the Zion Station, Units 1 and 2, in northern Illinois and the Indian Point Station, Units 2 and 3, in New York. These sites are being evaluated because they represent the four operating reactors which are located in areas of unusually high population density and therefore are believed to present a disproportion-ately hign contribution to the total societal risk from reactor accidents.
This evaluation will determine if additional design and procedural preventive or mitigative measures are warranted in order to reduce the probability of occurrence or to reduce the consequences of an accident more severe than the current design bases at these sites.
In the event of such a severe accident, releases of radioactivity to the public may be conveyed through the air, through the ground water supply, or by both paths. Postulated ra'dioactivity releases might occur rapidly due to a rupture of containment, or such postulated releases may occur at a later time due to a slow development of containment overpressure or as a result of a core melt-through which could eventually ledd to liquid pathway releases.
A rapid release of a given amount of radioactivity could result in high public consequences if there was insufficient time to implement protective measures such as evacuation.
Several rapid atmospheric release paths being evaluated include: steam explosions in the reactor vessel or the containment building which rupture the containment; hydrogen explosions; open vents in the containment at the time of an event; and Event V considerations.
Releases at various other time framas are also being evaluated and include: loss of power events; loss of heat removal events; 'and slow overpressurization.
In addition, slow liquid pathway releases will be evaluated.
Recent studies (see for example, NUREG-0440 " Liquid Pathway G'eneric Study")
indicate that the probability of a steam explosion rupturing containment during such a postulated accident is relatively small in comparison to other release mechanisms and therefore, in this evaluation, priority will be given to the study of the other release mechanisms.
The potential hydrogen explosion release path, open containment release path, and Evtnt V check valve failure path will be examined further, and, if found necessary, design and procedural preventive and mitigative measures will be required in order to reduce the probability of occurrence of these release paths.
The sustained loss of all AC power leading to core melt may inititate release paths of various time scales.
The probability of occurrance nf these release paths will be further evaluated, and preventive measures such as more reliable decay heat removal systems, or mitigative measures-such as a filtered vented cont;inmant system to prevent the resultant rupturing of the containment building will be required as necessary.
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o i Ulayed air path;.ay raleases due to a sicw cvarpressurization of ccntainrant nay be ganarated as a result of loss of contain 2nt uat ri oval, hyd.ogan burning, or due to a buildup of gases from a r31 ten fu21 - ccntainm2nt basenat interaction. These sic.er release paths will be enmined further, and mitigative or preventive design and procedural c asures will be required as necessary.
The sicw liquid pathway release results from fuel melting through the reactor vessel and through the containmant basemat or the containment walls. This release path will be examined further, and mitigative or preventive design and procedural measures such as core retention devices will be required as necessary.
Recognizing the length of time that may be required to implement some or all of the severe accident mitigation features (probably one to two years), the staff has evaluated a number of interim operational actions that should be implemented at these high population density sites for this period of time. Additionally, the staff is undertaking a concerted effort to accelerate current outstanding generic and plant specific licensing actions at these plants.
The general objective then is to define design or procedural measures that significantly reduce the likelihood and/or mitigate the con-sequences of an accident more severe than the current design bases at the Indian Point and Zion nuclear plant sites. I'.easures are to be identified that significantly reduce the probability of an event or.the source tem nagnitude of such an accident, or that result in significant additional time to respond to an accident at these sites.
Risk analysis may be helpful in establishing general concepts of appropriate action, but will not be used quantitatively to rule out positive plant improvements.
The general approach will be to pursue actively those design features that contribute favorably toward the prevention as well as the mitigation of the consequences of a severe accident. Llhere reliance is placed on the response of the external population, the time required should be commensurate withthe evacuation times estimated by the Federal Emergency Management Agency (FUM),
as available.
0_bjective, The objective of this project is to perform analyses and assessments in three specific areas directly related to the above program in order to aid in the licensing decision making which is to take place during the next 18 months.
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Th< contractor shali provide all %.w ico personnel, equipment, and fac lities to provide technical yt m N to the NRC in the area of licensing associated with the Zico and inolan Point activity.
Specific tasks to be included in this program and associated effort are:
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f.bK I CGE :di AC1lEif.i'GiYSIS The centr.w :or shall aerfom state-of-the-art analvses and assessments for Light h'ater Reactors (UGs), particularly for the indian Point Nuclear Units
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2 und 3 and for the Zion Nuclear Units 1 and 2 in the core-relt related areas indicated in the folicwing four subtasks:
Subtask:
a.
.biten Pool and Debris Bed Heat Transfer This includes analyses related to the themal-hydraulics of core debris penetration (molten or solid) into either containr.ent materials (steel, concrete) or core retention system materials (e.g., magnesim oxide, graphite).
Special emphasis in the area of molten pool heat transfer shall be devoted to phenomcna associated with freezing /re-melting, penetration of a hot solid mass of cote debris, concentration / density gradients, and gas generation effects from concrete interaction. The coolability of debris beds for the aforementioned water reactors will also be assessed, including a detailed study of the r.cenarios and conditions necessary to form a coolable debris bed.
Subtask:
b.
Thermophysical Properties This includes preparation of a data file on all relevant themophysical properties.
The properties which will be compiled are those which are especially relevant to perfoming consequence evaluations of core meltdo,sn events. This subtask shall complement but not be duplicative of the DOE Safety Analysis Coordinated Reactor Data (SACRD) program.
Subtask:
c.
Core Melt Computer Code Applications This includes implementation and application of computer codes (e.g., GRONS, CORCON, HESCHL) being developed, particularly as part of RES programs, for analyzing core melt front penetration by either molten material or a hot solid mass of material. These codes shall be applied to Zion and Indian Point for performing core meltdown consequence evaluations. }bdifications to existing computer codes shall also be included where this is deemed to be appropriate.
Subtask:
d.
Post-Accident Core Retention and Containment Systems This includes a review and independent evaluation of heat transfer analytical methods used by reactor designers in developing proposed core retention / containment system approaches. Special emphasis shall be devoted to perfoming sensitivity analyses for detemining the relative importance of various thema1/ hydraulic parameters, and examining the effectiveness of various proposed core retention / containment system approaches in accoraodating low probability accidents.
In the performance of subtasks a through d above, the contractor shall maintain l
close interaction with the BNL effort on' post-accident containment analyses, j
State-of-the-art technical information relevant to the above, in particular, data and analyses generated by centractors sponsored by NRC's Office of Nuclear Regulatory Research, shall be used by the contractor as appropriate.
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Tnis activity involves the evaluation of advanced decay-heat reaval syste. s for P'.Gs that have the. potential to improve the overall reliability n
De of decay heat recaval under operational and accident ccaditic systens (such as increasing the PORY relieving capacity, i prove:unts to existing Auxiliary Feedwater Systems (AFiO, and upgrading FiPI and recircula Tnis could
, as well as the additica of a completely diverse system.
pums) dedicated high pressure FJE system or a comparable feracater system,
- impacts, be a either of which could be " hardened" to protect against aircraftis now part of A system such as the last one toxic fires and sabotage.
~ The Geman System is located in a separate the Gewan Standard PhR plant.
This effort should be focused on the Indian building partially below grade. Point Nuclear Plants Units 2 and 3 and the ~ ion shall include the following subtasks:
Survey and Evaluation Subtask:
a.
Evaluate existing designs (e.g., the German bunkered system) and/or other conceptual designs and options that can improve Consider the reliability of the decay-heat removal in the Z/IP plants.
changes / add-ons, as well as totally diverse paths from the presently Establish design criteria and requirements and evaluate existing ones.the usefulness and/or the potential improvements to reliability of proposed ADMR systems.
Subtask:
b.
3pplication to Z/IP Based on the results of subtask a, determine appropriate Based on safety approaches for backfitting ADMR systems to Z/IP plants.
goals and design bases (to be determined by NRC) propose design criteria and requirements, and conceptual designs which are appropriate for the Z/IP plant / site.
Impact of ADMR SysterIs en Overall Safety Subtask:
c.
Using deterministic as well as quantitative-probabilistic (at 1 cast in a relative sense) assessments determine contribution t safety of the potential candidate ADMR systems backfitted to the Z/IP plants, including an estimation of the uncertainties involved in such assessments.
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tai!. III. R!sK RhKi10N FRITERI.; MR A ?]i.L a-E.ihD e ~J Q cif Wre have been several concepts suggested for use as engineered safety features ained at mitigating the censequences/ effects of degraded core /
core melt accidents. Of pri;r.ary interest is the use of a filtered-vented containnent system (FYCS) in conjunction with methods to cope with hydrogen generation / accumulation for maintaining containment integrity following i
these potential accidents. The ultinate decision to require these features depends upon their potential for risk reduction.
i In ti s task, potential acceptance criteria for employment of a FVCS in the
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Z/IP plants, including the effects of hydrogen generation / accumulation will be derived. These criteria will be fonnulated in tems of irproved safety potential.
From the criteria, the requirements for designing these engineered features will be obtained. The various proposed concepts and designs will be examined to determine whether or not such requirements can be net and/or what design changes could be made to meet them.
Included in this task is the consideration that will be given for both generic (such as steam explosions) and site specific (such as earthquakes) effects.
In the latter context, various design bases such as the OBE and SSE used for other safety and non-safety grade features will be explored for filtered-vented concepts, as well as their interaction.
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