ML19290A130
| ML19290A130 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 10/12/1979 |
| From: | Norberg J Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 7910170348 | |
| Download: ML19290A130 (14) | |
Text
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NOTE T0:
Document Control Room 016 FRDM:
James A. Norberg, Engineering Methodology Standards Branch, 443-5921 Please place the attached document in the PDR using the following file and file points:
PDR File Related Documents (Select One)
(Enter if appropriate)
Proposed Rule (PR)
ACRS Minutes No.
Reg. Guide Proposed Rule (PR)
Draft Reg. Guide Draft Reg. Guide Petition (Pa4)
Reg. Guide Effective Rule (RM)
Petition (PRM)
Fffective Rule (RM)
Other 7d7 W4 Federal Register Notice EL Task No.
NUREG Report Contract No.
Subject:
SD Plan for Consideration of Degraded Core Cooling in Regulations (Memo from Robert B. Minogue to Lee V.
Gossick) m 79101793 2223 296
. l SD Plan for Review and Uooradino of USNRC Reculations and Reculatory Guides' for Consistent Acolication to the Desion and Analysis of Systems Imcortant to Safety Relative to Decraced Core Coolino Backcround The recent accident at Three Mile Island identified the need to reexamine the USNRC regulations and regulatory guides with respect to the occurrence of significant degraded core cooling during seve~re accidents. Review of these regulaticns and guides ind,1 cates that the consideration of effects of degraded core cooling is not applied u;riformly.
Currently, degraded core cooling is considered at three levels in the regulations. The first level is substantial degraded core cooling and associated fuel damage that is assumed to result in radioactivity releases as specified in TID 14844 (R.G.1.3 and 1.4).
This level of OCC is used primarily for siting considerations related to 10 CFR 100 and for design and qualification of engineered safeguards required as compensating features under part ICO.
The containment and equipment important to safety located within the containment is designed to accommodate TID source ;erms (see R.G.'s 1.52,1.8g).
The second level of DCC is that used for est< blishing design criteria of systems to control comoustible gas inside containment (see 10 CFR 50.44 and R.G. 1.7).
This amount of DCC is considerably less than that which would result in fuel damage causing TID 14844 source levels and at least five times greater (relative to hydrogen released from fuel clad-water reaction) than criteria specified for emergency core cooling design (see 10 CFR 50.46).
The third level of DCC is that bounded by the ECCS criteria,10 CFR 50.46 and Appendix K to 10 CFR 50.
TID 14844 specifies the folicwing:
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. "It is assumed that the reactor is a pressurized water type for which the maximum credible accident will release into the reactor
-building 100 percent of the noble gases, 50 percent of the halogens and 1 percent of the solids in the fission product inventory.
Such a release represents approximately 15 percent of the gross fission product activity."
Regulatory Guides 1.3 and 1.4 specify:
" Twenty-five percent of the equilibrium radioactive iodine inventory developed from maximum full power operation of the core should be assumed to be inmediately available for leakage from the primary reactor containment.
Ninety-one percent of this 25 percent is to be assumed to be in the form of elemental iodine, 5 percent of this 25 percent in the form of particulate iodine, and 4 percent of this 25 percent in the form of organic iodides.
"One hundred percent cf the equilibrium radioactive noble gas inventory developed frem maximum full power operation of the core should be assumed to be inmediately available for leakage from the reactor containment."
10 CFR 100 permits a trade off between safety features that are engineered into the facility and site factors, including population density and distribution.
Because of this trade off, Part 100 has the effect of a design regulation. Mcw-ever, only the containment and those safety features within'the containment are designed to fully accommodate TID type source terms. Other safety features that are engineered into the facility which are located outside the containment may not be designed to accommodate TID type sources.
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3-The TMI-2 accident vividly pointed out some of the non-unformities in design of systems important to safety relative to degraded core cooling.
For example, the residual heat removal (RHR) system has not been operated, primarily because of the undesired possibility of leakage of highly radioactive primary coolant outside contlinment; sampling systems for the primary coolant could not readily acconnodate the high radioactivity levels; and the radiation shielding to permit early connection and operation of the ex-containment hydrogen recombiner system was apparently inadequate. Another example is the criterion that was used at TMI for containment isolation and the potential for transfer of highly radio-active water to the auxiliary building via the containment drainage system.
Some less obvious impacts of the degraded core cooling can be identified from the TMI accident.
For example, the presence of substantial quantities of steam and noncondensable gases trapped within the primary system affected the thermal-hydraulic performance of the primary system.
Such conditicns are not assessed by the usual LOCA type licensing calculations (App. K to 10 CFR S0).
As the TMI accident is further analyzed additional insights and understanding will be forthcoming to assist in the review and upgrading of the regulations relative to degraded core cooling considerations.
Action Plan SD has established a dedicated group within the Division of Engineering Standards that is performing a comprehensive review and evaluation of the Category I regulations and regulatory guides to assess the application of degraded core 2223 299
. cooling in a more uniform and across-the-board manner. Areas that need improve-ment are being identified a'.d reconr.endations made on how to implement such improvements.
Classificat.cn including qualification of plant systems and components is an araa thr.t will receive particular attention. The TMI-2 accident and post-accident experience have shown that scme systems and components may be more important to safety than previously thought.
Plant systems and components impacted by DCC conditions will be identified and their classification including qualification requirements will be evaluated relative to their importance to safety.
Recommendations will be made for changes in classification of such systems and ccmponents as may ce appropriate.
It is not anticipated that all impacted regulations or regulatory guides will require consideration of the same degree of degraded core cooling. For example, the ECCS criteria of 10 CFR 50.46 are not expected to be impacted; however, Appendix K to 10 CFR 50 might be impacted relative to small break analysis.
The development of the criteria to be used to define the extent of degraded ccre cooling is outside the scope of this SD effort.
Mcwever, the effort will address the regulatory form of the criteria and will provide insight on the potential impact of the criteria en the regulations and regulatory guides and their implementation on nuclear power plants.
It is recognized that consideration of the impact of degraded core cooling extends to a bread spectrum of the regulations, guides, and regulation practices; however, due to limited manpower resources, the initial effort will concentrate en Civision 1 Regulatory Guides and Regulat a ns. Close cognizance of the NRR Lessons Learned Task Force efforts will be maintained and their results, as well as results frem other NRC TM!-2 related efforts sill provide 2223 300
. additional guidance, particularly regarding priorities, for implementing any needed changes to the regulations and regulatory guides.
It is also anticipated that outside assistance may be required to help implement the DCC task effort and, if needed, such assistance will be defined as the work progresses.
The DCC dedicated group is comprised of the following SD staff members:
Name Cedicated Time, Hichest priority up to J. A. Norberg, EMSB 60%
E. G. Arndt, SCSB 85%
E. J. Brown, SCSB 85%
G. A. 'deidenhamer, SCSB 85%
A. H. Hintze, RSSB 85%
T. Scarbrough, RSSB 60%
M. S. Weinstein, FPSSB 85%
All time spent working on this effort will be charged to 79040 JYO T J91330000.
An outline of the major subtasks of the plan along with a target completion date is given belcw:
I.
Identify specific regulaticns, regulatory guides and standard review plan sections that may be impacted by consideration of degraded core cooling (DCC). 5/17/79 C II.
A.
Examine each identified item in I and:
1.
Su.marize tha existing requirements, implied assumptions, etc
~
for consideration of DCC.
5/20/79 C 2223 301 2.
Determine the potential degree of impact to consider substantial DCC (i.e., up to TID Source).
7/20/79 C 3.
Make an estimate of the value/ impact of considering the DCC.
7/20/79 C 4.
Establish a recommended priority for action based on safety considerations (i.e., potentir. for release of radioactivity to environment including hazards to plant personnel).
7/20/79 C B.
Review LWR plant systems to identify those impacted by DCC and to assess their classification / qualification relative to safety. 9/15/79 C C.
Identify research and technical assistance needs - coordinate with RES and NRR as appropriate.
Open III.
Prepare a report su::narizing the results of I and II above.
10/1/79 IV.
A.
Initiate standards tasks for high priority items. Such tasks will be initiated as socn as the items are clearly identified and their priority agreed upon by SD/NRR.
Open Several regulations and regulatory guides are readily identified as prime candidates for modification to account for degraded core cooling. Among these are 10 CFR 50.44 and R.G. 1.7 dealing with ccmbustible gas control in the containment, R. G. 1.97 dealing with instrumentation to follow the course of an accident, and R.G.1.139 dealing with residual heat removal.
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Consideration will be civen to blanket regulation / regulatory guide to generically address DCC. A new rule and/or regulatory guide could be implemented that would eliminate the need for IV, A, above, or could be an interim measure pending ccmpletion of revisiens to affected regulations and guides.
Open V.
Preliminary List of NRC Regulations and Guides that May be Impacted by Consideration of Degraded Core Cooling.
Each regulation and Division 1 regulatory guide in this preliminary list will be examined in accordance with Section II above.
If cther exisiting regulaticns or Division 1 regulatory guides are identified as requiring c:nsideration relative to degraded core cooling, they will also be examined. Additionally,1. new regulaticns or Division 1 regulatorf guides are identified, they will be apprcpriately factored into the task ef#crt.
Tne preliminar/ list of regulations and Division 1 regulatory guides has been separated into *wo categories; A - Probably Impacted, and 3 - Possible Imoacted.
The list has been broken dcwn and tentative assignments made to the memoers of the group, as shcwn belcw, for their action relative to Section II.
E. G. Arndt. SCSB Category A - Probably Impacted
- R. G.1.57, Design Limits and Lcading Cecoinations for :!stal Primary Reactor Centainment Systems Cemconents 2223 303
. Category B - Possible Impacted 10 CFR 50.55a 10 CFR 50, Appendix J R.G.1.19, Noncestructive dxamination of Primary Containment Liner Welds R.G.1.90, Inservice Inspection of Prestressed Concrete Containment Structures with Grouted Tendons R.G.1.107, Qualification for Cement Grouting for Prestressing Tendons in Centainment Structures R.G.1.136, Material for Cencrete Containments R.G.1.142, Safety-Related Concrete Structures for Nuclear Power Plants (Other Than Reactor 'lessels and Centain-ments)
- 5. J. Brcwn. SCSB Category A - Probably Impacted 10 CFR 50, Appendix K R.G.1.25, Quality Group Classifications and Standards for Water,
Steam, and Radioactive-Waste-Centaining Components of Nuclear Power Plants Category 3 - Possible Imcacted m 3 304 10 CFR 50, Aopendix A
. R.G.1.11, Instrument Lines Per.etrating Primary Reactor Containment R.G.1.57, Installation of Overpressure Protection Devices A. Hintze, RSSB Category A - Probably Impacted R.G.1.40, Qualification Tests of Continuous-Outy Motors Installed Inside the Containment of Water-Cooled Nuclear Power Plants R.G. l.73, Qualification Tests of Electric Valve Operators Installed Inside the Containment of Nuclear Pcwer Plants R.G. 1.75, Physical Independence of Electric Systems R.G.1.89, Qualification of Class 1E Equipment for Nuclear Pcwer Plants R.G.1.97, Instrumentation for Lignt-Wate"-Ccoled Nuclear Pcwer Plants to Assess Plant Conditions During and Folicwing an Accident R.G.1.131, Qualification Tests of Electric Cables, Field Solices, and Connections for Light-Water-Cooled Nuclear Pcwer Plants 2223 305 Category 3 - Possible Imcacted R.G.1.3, Assumptions Used for Evaluating the P0:ential Radiciegical Consequences of a Loss-of-Coolan Accident for Boiling Watar Reactors
. R.G.1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss-of-Coolant Accident for Pressurized Water Reactors R.G.1.2%, Periodic Testing of Protection System Actuation Functions R.G.1.32, Criteria for Safety-Related Electric Pcwer Systems for Nuclear Power Plants R.G.1.47, Bypassed and Inaperable Status Indication for Nuclear Power Plant Safety Systems R.G.1.58, Initial Test Progrcms for Water-Cooled Reactor Power Plants R.G.1.79, Preoperational Testing of Emergency Core Ccoling Systems for Pressurized Water Reac' rs R.G.1.81, Shared Emergency and Shutdown Electric Systems for Multi-Unit Nuclear Power Plants T. Scarbrauch, RSSB Category A - Probably Impacted R.G.1.139, Guidance for Residuil Heat Removal R.G.1.141, Containment Isolation Provisiens for Fluid Systems Category 8 - Possible Impactad R.G. 1.96, Design of Main Steam Isolation 7alve Leakage Control Systam for Soiling Water Roactor Nuclear Pcwer ?lants 2223 306
. G. A. Weidenhamer, SCSB Category A - Probably Impacted R.G.1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants Category 8 - Possible Impacted 10 CFR 50, Appendix G 10 CFR 50, Appendix H R.G. 1.45, Reactor Coolant Pressure Boundary Leakage Detection Systems R.G.1.82, Sumos for Emergency Core Cooling and Containmen:
Spray Systems R.G.1.83, Inservice Inspection of Pressurizec Water Reactor Steam Generator Tubes M. S. '4 einstein. FPSSB Category A - Probably Imoacted 10 CFR 50.44 R.G.1.5, Assumptions Used for Evaluating the Potential Radiological Consecuences of a Steam Line Break Accident fer Sailing Water Reactors
} }g R.G.1.7, Centrol of Comcustible Gas Concentratiens in Centainment Folicwing a Loss-of-Coolant Acc fent
. R.G.1.21, Measuring, Evaluating, and Reporting Radicactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents frem Light-Water-Cooled Nuclear Power Plants R.G.1.52, Design Testing, and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adscrption Units of Light-Water-Cocied Nuclear Pcwer Plants R.G.1.143, Design Guidance for Radioactive Waste Management Systems, Structures, and Ccmponents Installed in Light-Water-cooled Nuclear Pcwer Plants Category 3 - Possible Impacted R.G. 1.13, Spent Fuel Storage Facility Cesign Basis R.G.1.24, Assumptions Jsed for Evaluating the ?ctential Radiological Consequences of a Pressurized Water Reactor Radioactive Gas Storage Tank Failure R.G.1.25*, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and 3ressurized Water Reactors J. Marcere, (O. Notlev), EMSR Category 3 - Possible Imcacted R.G.1.120, Fire Protection Guidelines for :luclear Pcwer Plants
. VI.
Division 1 Regulatory Guides under development that may be impacted by Consideration of Degraded Core Ccoling In addition to the inplace regulations and Div. 1 regulatory guides, several biv.1 guides that are under development have been identified as possibly being impacted by consideratien of degraded core cooling.
These guides will also be examined in accordance with Section II above only en a icwer priority basis, i. e. after c:mpletion of Section V.
A preliminary list of these draft guides is shown below; however, personnel assignments have not yet been made for their assessment.
Category A - Probably Impacted Proposed Guide ( R.C.
l.SCl) - Rec:mmendations for Inservice 7esting,
of Val /es Required to Perform t Safety Functicn in L'ARs Proposed Guide - Criteria for Electric Instrumentation and Control Systems Portions of Protection Systems and P otective Action Systems (endorses IEEE 503-1977)
Proposed Guide (RS811-5) - Qualificaticn of Electric Mcdules fer 11PPs.
Proposed Guide (EM622-5) - Single Failure Criteria for Light Water Reacto'r Safety Related Fluid Systems (endorses At15 53.9)
Category 3 - Possible Imaacted Procesed Guide (SC704-5) Functional Specifications for Safety-Related Valve Assemolies in ltuclear Power 31 ants
14 -
Category 3 - Passible Impacted (cont)
Proposed Guide (EM710-5) - PWR and BWR Containment Spray System Design Critaria (endorses ANS 56.5) 9 e
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