ML20214W179

Affidavit of Sj Cereghino & Wv Cesarski Re Temp Margins in Environ Qualification by Thermal Lag Analysis of Asco Valve Models NP-8616,NP-8320 & NP-8321.Related Correspondence

ML20214W179
Person / Time
Site:	Vogtle
Issue date:	12/02/1986
From:	Cereghino S, Cesarski W BECHTEL GROUP, INC., GEORGIA POWER CO., WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20214W161	List: ML20214W159 ML20214W179 ML20214W238
References
OL, NUDOCS 8612100116
Download: ML20214W179 (48)

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of )

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GEORGIA POWER COMPANY, et al. ) Docket Nos. 50-424 (OL)

) 50-425 (OL)

(Vogtle Electric Generating Plant, )

Units 1 and 2) )

AFFIDAVIT OF STEPHEN J. CEREGHINO AND WILLIAM V. CESARSKI -

Stephen J..Cereghino and William V. Cesarski, being duly sworn according to law, depose and say as follows:

1. (SJC) My name is Stephen J. Cereghino. I am employed by Bechtel Western Power Corporation in the position of Nuclear Engineering Group Supervisor. My business address is Bechtel Western Power Corporation, 12440 East Imperial Highway, Norwalk, California 90650. A summary of my professional qualifications is attached as Exhibit "A" to this affidavit.

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,3 (WVC) My name is William V. Cesarski. I am employed by Westinghouse Electric Corporation in the position of Senior Engineer in the Equipment Qualification Department of the Genera-tion Techno1cgy Systems Division. My business address is West-A.

S inghouse Electric Corporation, R&D Center - Building 701, 1310 Beulah Road, Pittsburgh, Pennsylvania 15255. A summary of my professional qualifications is attached as Exhibit "B to this iaffidavit.

3. (SJC,WVC) The purpose of this affidavit is to respond to those portions of the Affidavit of Howard M. Deutsch (November

. 26, 1986) that could be construed as pertaining to the tempera-ture margins in the environmental qualification by thermal lag analysis of ASCO valve models numbered NP-8316, NP-8320, and NP-8321. In this respect, we will address pages 6 through 8 of Dr. Deutsch's affidavit. We have personal knowledge of the in-formation presented herein and believe it to be true and correct.

4. (SJC,WVC) At page 6 of his affidavit, Dr. Deutsch re-fers to an NRC report entitled " Impacts of Budget Cuts on NRC's Ability to Assure Safety" and dated April 30, 1986. A copy of this report is attached as Exhibit C hereto.

5. (SJC,WVC) Dr. Deutsch first refers to page 19 of the report. There, the report indicates that in many instances there are engineering judgments inherent in current environmental qual-ification methodologies, and that further research would have reduced such uncertainties. It concludes that research would have provided "the basis for revisions to the rules and regula-tory guides."

6. (SJC,WVC) We have reviewed this discussion in the report and have found nothing specific that might question the environmental qualification by thermal lag analysis of ASCO sole-noid valves at Plant Vogtle. The environmental qualification of ASCO solenoid valves at VEGP meets or exceeds the requirements of 10 C.F.R. 6 50.49, NRC Regulatory Guide 1.89 (which incorporates and endorses IEEE Standard 323-1974), and NUREG-0588 Rev. 1.

These standards recommend a 15 F temperature margin to compensate for uncertainties. The thermal lag analyses for ASCO valves at VEGP have even greater margin. While future research might prompt the Nuclear Regulatory Commission to refine its regula-tions, the standards and methodologies enumerated above are those which the Commission currently requires be met.

7. (SJC) At page 7 of his affidavit, Dr. Deutsch refers to page 26 of the report. At this page, the report discusses further research which could improve computer codes in the model-ing of postulated core-melt accidents. Dr. Deutsch uses this discussion to " question the applicability of the Bechtel FLUD code in calculating the Vogtle specific MSLB/LOCA conditions."

With respect to the ASCO solenoid valves qualified by thermal lag analysis for Plant Vogtle, Bechtel's ELUD code was used to e

determine environmental conditions (nodal temperatures, pres-sures, and velocities) resulting from postulated breaks in piping in areas outside containment. The thermodynamics for calculating these parameters in various compartments for the mass and energy releases from a pipe break are well known and can be reliably modeled. The FLUD code is not used to model core-melt accidents or loss of coolant accidents, and the discussion in the NRC report to which Dr. Deutsch refers is inapplicable.

8. (WVC) On pages 7-8 of his affidavit, Dr. Deutsch ar-gues that because of the statement in the NRC report discussed abovo, the averment that the 346 F [Isomedix] qualification tem-perature should be accurate to one degree is inadequate. The 346*F qualification temperature derived under the Isomedix quali-fication program was not based on the FLUD code or any other code. Rather the 346*F qualification temperature reflects the env,ironmental temperature, measured by calibrated thermocouples, to which valves were exposed for a sustained period (three hours) in a test chamber. The type of thermocouples used by Isomedix can measure temperature accurately to within one degree F.

9. (SJC,WVC) For these reasons, we conclude that the Af-fidavit of Howard M. Deutsch contains no information to question the temperature margins in the environmental qualification by thermal lag analysis of ASCO solenoid valve models NP-8316, NP-8320, and NP-8321. As demonstrated in our prior affidavit, in

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each instance the temperature margin exceeds the 15 margin rec-ommended by IEEE-323-1974.

hd N epfen. Ceregh .o Wu William V. Cesarski Subscribed and sworn to before me this a day of December, 1986.

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, Notar ublic My Com:nission ES:5I.o2311930 1

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i EXHIBIT "A" STEPHEN J. CEREGHINO l

EDUCATION: B.S., United States Naval Academy Naval Nuclear Power School Naval Nuclear Power Training Unit MBA, Business Administration, Whittier College

SUMMARY

7 Years: Bechtel engineering responsibilities in licensing and systans integration on the Vogtle project.

6 Years: Various training, operational and maintenance responsi-bilities associated with the naval nuclear propulsion program. .

EXPERIENCE: Mr. Cereghino is Project Vogtle's Nuclear Group Supervisor. .

In this capacity, he provides technical guidance and assistance in the licensing and design of Plant Vogtle. As licensing engineer, he coordinates the inter-discipline activities of project personnel and coordinates with the client, NSS$ and NRC personnel to ensure consistent application of licensing commitments. Mr. Cereghino supervises the administration of the NSSS contract, including such activities as: NSSS vendor data review, evaluation of NSSS proposals, and coordination of A/E-NSSS interface activities. In the systems integration area, Mr. Cereghine is responsible for the analytical evalua-tion of potential plant hazards, such as: radiation,- pressure, temperature, flooding, internal missiles and seismic inter-actions.

Prior to joining Bechtel, Mr. Cereghino was an officer in the United States Navy. His shipboard engineering assignments were as Reactor Controls Officer and Main Propulsion .

Assistaat. He routinely supervised the operation of the reactor plaat during all modes of operation, and directed the chemistry control and radiation protection programs for ships company. Mr. Cereghino's last assignment with the Navy was as a Division Director at the Naval Nuclear Power School; as such, he coordinated the instruction of Reactor Principles to enlisted plant operators. Before leaving the Navy, Mr. Cereghine successfully qualified to assume the responsibilities of Chief Engineer of a naval nuclear pro-pulsion plaat.

PROFESSIONAL AFFILIATIONS:

Professional Registration: Mechanical Engineering, State of California l

EXHIBIT "B" Summary of Professional Qualifications and Experience William V. Cesarski ,

I Senior Engineer Plant Engineering Division Westinghouse Electric Corporation My name is William V. Cesarski. My business address is Westinghouse Electric Corporation, R & D Center-Building 701, 1310 Beulah Road, Pittsburgh, Pennsylvania 15235. I an employed by ' Westinghouse Electric Corporation (" Westinghouse")

as a Senior Engineer in the Equipment Technology Department of the Plant Engineering Division.

I graduated from the United States Military Academy in 1964 with a Bachelor in Engineering Science degree. I was awarded an Atomic Energy Commission Graduate Fellowship while at West Point and used the AEC fellowship to obtain a Master of Science Degree in Nuclear Engineering from Massachusetts Institute of Technology in 1966. In 1972 I also received a Master of Science Degree in Industrial Management from New York University. After spending eight years in the U. S. Army, I joined Westinghouse in 1972 as an engineer in the Plant Apparatus Division. While working at WPAD, I obtained experience in nuclear valve and refueling equipment design, testing and procurement for the Naval Nuclear Program.

i In 1981 I joined the Westinghouse Nuclear Equipment Division l and have had lead engineer responsibility for the IEEE qualifi-cation testing of numerous NSSS valve and motor components.

I have conducted numerous qualification test programs.and authored numerous Westinghouse qualification test reports on components such as valve motor operators, valve limit switches, solenoid valves, valve position indication devices, pump motors and pump assemblies. I am presently a Senior Engineer

and act as a lead engineer in the Equipment Qualification l Technology Department of the Plant Engineering Division responsible ,for electro-mechanical equipment qualification.

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EXHIBIT "C" A

APR 8 01986 MEMORANDUM FOR: Samuel J. Chilk Secretary .

FROM:

Victor Stello, Jr.

Executive Director for Operations .

SUBJECT:

REPORT REQUESTED BY COMMISSIONERS ASSELSTINE AND 20, 1985, enclosed is a report as In response to your memorandum of November requested by Comissioners Asselstine and Bernthal on research Thisprojects report which for budgetary reasons were deferred or could not be accomplished.

also addresses the impact of the deferred and cancelled research projects.

In prioritizing research activities, the staff has attempted to provide for a stable, properly balanced research program focused There on supporting is clear regulatory activities to ensure safety at operating facilities.

recognition in this prioritization that industry must assume I do the not burden of funding research where industry is the primary benefactor on the health and safety of the public.

research capabilities and the resultant inabilities to provide necessary information to answer safety questions cannot be underestimated for the future.

Original signed by Victor Stello . -

Victor Stello, Jr.

Executive Director for Operations

Enclosure:

Impact of Budget Cuts on NRC's Ability to Assure Safety cc: Chairman Palladino Commissioner Roberts Commissioner Asselstine Commissioner Bernthal Commissioner Zech I

OGC OPE ek Memo revised in OEDO 4/28/86 4 86 Encl. revised in RES 4/11/86

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3/14/86 :3/14/86 :3/16/86 :3/18/86 :4/ /86 :4/fr/86 DATE :3/13/86 0FFICIAL RECORD COPY

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IMPACTSOFBUDGETCUTSONNRC'SABILITYTOASSURESAFETY(OVERVIEW)

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NRC has reduced its research program significantly in the last five years partly because of completion of some major projects like LOFT but also to confom to the declining budget. The effect of some of the specific actions is outlined in the attached sheets. These are just a few of the larger programs that have been reduced, eliminated or delayed (probably indefinitely if the Grann-Rudman required budget levels are realized). Many smaller programs have been also eliminated or delayed.

As the NRC faces safety issues now before it, the effect of erosion of research capabilities and resultant inabilities to meet the need for new safety information to be used to help resolve these and other regulatory problems can already begin to be seen. The reduction in safety research information resources (scientific and engineering expertise, verified computer codes with knowledgeable support personnel, and experimental facilities) is already resulting in requests from the regulatory staff having to be denied or delayed for intolerable periods of time. For example, the regulatory staff has asked for a series of operational transient tests in a facility to simulate accidents in reactors built by Babcock and Wilcox to follow on behind current tests of small break 16ss of coolant accidents. They also asked for extensive tests of feedwater and. steam line breaks in Westinghouse and Combustion Engineering reactors. In the former case the tests have been delayed at least two years until funding is found. In the latter the tests were reduced from 14 to 5.

Currently NRC is in the final two years of a six year program of evaluating its regulatory approach to severe accidents in nuclear power plants. The NRC staff i is in the process of summing up the investigation in a series of reports which involve reassessment of the radioactive source tem from severe accidents, the -

risk and consequences of such accidents, the implementation of what has been learned into the regulatory process and the evaluation of rules and regulatcry instruments such as the siting and the emergency planning rules for the need for revision. In evaluating the various technical issues that must be resolved to provide a basis for these actions, it is becoming apparent that some of the uncertainties may be so large, even with knowledge gained from the four years )

of intense focused research to date, that NRC may not be able to provide a satisfactory reduction of these uncertainties with existing resources. This is at least aartly due to program reductions in this area over the past years.

Some of t1ese issues may be the chemical and physical forms of radioactive fodine and cesium and their mutual interaction, the degree of direct containment heating from the expulsion of molten reactor core materials, hydrogen generation and loads and containment performance in resisting these and other loads placed upon it from a severe accident. Current reduced research programs are addressing these and other severe accident issues, however some issues may not be resolved in the remaining two years of research because the programs have been cut. The program is now at the point where the most serious issues have been identified, but the ebility to solve them has been reduced.

The NRC research program addresses the ability to understand and predict the behavior of power plants as a result of transients and accidents. This information is used to help reduce the potential for accidents. The focus of this research is the understanding and modeling of thermal hydraulics. All i

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IMPACTS (CONT'D) integral experimental facilities in the United States will have been shu Although studies have been made of ways of providing by the end of 1986.

continuing experimental capability, plans for such capability have bee frustrated by budget reductions.

other countries in integral testing This as aallows means of reducing access funding require to results from and sharing safety research information.

foreign experimental facilities; however, it does not prov'ide the ability t perform experiments the NRC deems necessary, particularly on a U.S. industry people have taken the attitude that the plants are safe scale.

enough and industry has shut down or plans to shut down its integral test This inability to conduct experiments to examine the safety facilities.

implications of important plant transients, which typically Reductions occur at the ra in the safety

~ of one or so per year, may present real problems.

research budget are expected to have intermediate and long term implicat that will be detrimental to public health and safety their elimination.

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4 Fuel Melt Source Tem Tests in the Power Burst Facility (Reduction in FY 1984 -

54M, in PY 1985 - 35M In Fy 1956 - 312M and in Fr 1987 - $6M)

1. Safety Issue: Large uncertainties exist in analyzing in-vessel severe accident benavior at core melt conditions, including clad oxidation heating, hydrogen generation, core meltThis progression, and fission in turn leads to large product and aerosol generation and transport.

uncertainties in containment heating and penetration : calculations which in turn are used in health effects consequence deteminations.

2. Current Treatment by NRC: These processes are currently modeled in the I Source Term code Package (SCTP) by NRC codes such as MARCH and CORSOR which are not verified against experimental data. These codes contain Furthemore, almost many arbitrary assumptions and large uncertainties.

no data exist at the higher temperatures near fuel melting (3,100K) nor on the core-melt progression process itself; so the analysis is of necessity quite uncertain.

3. Role of Deferred Research: The high-temperature second phase of the Severe Fuel Damage and source Tem tests in the PBF test reactor would have provided unique integral in-vessel severe-accident behavior up to fuel-melt temperatures (3,100K) including data on core-melt progression and fission-product and aerosol release and transport. The earlier PBF tests provided most of the current infomation on integral in-vessel severe-accident behavior at intermediate severe accident temperatures.

One of the planned second phase tests was for a PWR accident transient, and the other would have been the only large integral BWR severe-accident test. The net effect of not doing the second phase tests is that no large integral data will be available on in-vessel fission-product behavior and early core melt progression to near fuel-melt temperatures to benchmark and validate the codes currently used in the source-tem code package (STCP) and the new advanced codes such based on actual physical and chemical processes.

Onlyas verySCDAP and MELPRO limited data on fission product release under actual in-core accident conditions including radiation and high pressure will be available to validate the currently-used CORSOR fission-product release code and the new advanced VICTORIA and FASTGRASS codes. ,

4. Work by Industry: Industry does not have available the test faci.6 ties to perfom large integral in-pile tests such as these.

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o Delayed or Cancelled Source Tem Tests in the Annular Core Research Reactor

[ Reduction in FY 1986 - 3100K, in FY 1987 - 31,500K)

1. Safety Issue: Large uncertainties exist in the in-vessel fission-product release rates currently used in Source Tem analysis with the Source Tem Code Package (STCP). These rates are taken from early old, relatively coarse, out-of-pile data at atmospheric pressure. The relatively volatile fission products not released from the fuel in-vesselmand then removed by plate out in the reactor vessel are imediately released ex-vessel by melt-concrete interactions. In cases of early containment failure these fission products can become the largest part of the release to the containment.

2. Current Treatment by NRC:

Current source-tem analysis with the Source-Tem Code Package (STCP) is based on data from out of reactor tests. Fission-product release rates inferred from the large and complex Power Burst facility reactor tests at 1,000 psi and in a radiation environment are ten-to-one-hundred times lower than out-of reactor rates. The uncertainties reflected in this difference produces uncertainties of up to a factor of ten in source tem analysis.

3. Role of Cancelled Research: A program of precise, improved fission-product release-rate experiments has been started to resolve this large and significant difference and to provide a much improved data base over the full range of in-vessel severe accident conditions. This improved 4

data base would be used to benchmark the new VICTORIA fission-product behavior module of the MELPROG core-melt-progression analysis code for use over the full range of severe-accident conditions. The program included eight in-pile separate-effects fission-product release-rate experiments in ACRR to cover the range of in-core severe-accident conditions, including the effects of temperature, pressure, fuel liquifaction, debris geometry, Improved and steam-hydrogen ratio, all in a realistic radiation field.

out of reactor experiments were also included in the now cancelled program. With the program cancellation, only two of the planned eight experiments can be performed these will give only a very limited data ,

comparison on the in-pile and out-of-pile difference and a pressure-dependence check, all in a reducing hydrogen environment, with no oxidizing steam-environment data.

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4. Work by Industry: There is no prospect of having this in-pile fission-product-release work done by industry. EPRI has supported some i

somewhat related work in the STEP aerosol experiments in the TREAT  !

reactor, but these do not provide the needed fission-product release-rate data.

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Reduced Capability for Severe Accident Sequence Analysis (Reduction in FY 1986

- 3300K, reduction in FY 1957 - 51,500KJ

1. Safety Issues: NRC Unresolved Safety Issues such as ATWS, the Hydrogen Rule, Station Blackout and Risk Rebaselining for the Source Term Reassessment have been supported by Severe Accident Sequence Analysis (SASA) by applying NRC safety analysis codes and experimental results from severe core degradation experiments to provide realis.ti.c results for -

specific plants.

Current Treatment by NRC: Approximations and assumptions are made that 2.

are conservative but result in accidents havingexample A current apparently much is the evaluation larger of societal impact than is realistic.

the decontamination factor for the secondary containment for BWR Mark I plants. The conservative view is to assume no credit at all for secondary containment.

Timely response to such licensing issues will

3. Role of Deferred Research:

be severely 11mited. The cut of more than 50 percent contemplated for FY 1987 would . require choosing among such current actions as secondary containment studies or the intentional venting of MARK I containments.

This would leave no resources to respond to substantive problems (e.g.,

hydrogen detonation) which may not be resolved on the current schedule.

None, this work is internal

4. Prospects for Work Being Done by Industry:

NRC support.

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1985 - $800, in FY 1986 -

Curtailment of Hydrogen Research (Reduction in FY 5800K and in FY 1987 - 5700K)

Safety Issue: One of the safety issues is whether or not, in the absence 1.

of a hydrogen control system, given hydrogen-steam mixtures in large dry containments, would such mixtures autoignite and burn slowly destroying containment seals or, alternatively, burn rapidly and damage containment equipment or' structures. The information would allow NRC to evaluate the need for changes in equipment qualification and hydrogen regulations.

Another issue not resolved involves the quantitative loads to containment walls and electrical and mechanical safety equipment from local hydrogen detonations and the probability of wall and equipment damage.

2. Current Treatment by NRC: The NRC hydrogen transport and combustion codes, HECTR and HMS-BURN, have been assessed against limited experimental data for both hydrogen transport and autoignition. Conservative assumptions are made by the NRC staff to account for these uncertainties in the codes. It is now assumed that in the absence of reliable autofgnition data, that ignition will occur from some random source at a later time when the hydrogen concentration is higher. Because detailed structural response data is not available, conservative loads are assumed to be transferred to containment walls and equipment. The likelihood of a high temperature hydrogen-steam mixture to autofgnite and form intensely-burning diffusion flames has not been extensively investigated. '

However, data is being generated on the consequences of diffusion flames from deliberate ignition systems (igniters). Data obtained from the BWR 6/ Mark III Hydrogen Control Owner's Group (HCOG) research program will be used to determine the thermal environment resulting from diffusion flames.

In addition the data can be used to assess hydrogen transport and mixing for the Mark III containments.

The hydrogen research program would have been

3. Role of Deferred'Research:

extended to provide a sufficient data base for the understanding of the diffusion flame and autoignition issue which the NRC regulatory staff These issues are the foundation upon which considers to be unresolved.

resolution of the damage to seals, walls, and equipment issues can be accomplished.

4. Prospects for Work Being Done by Industry: None. NRC and EPRI conducted some hydrogen experiments in a large sphere at the DOE Nevada Test Site, 1 however, these experiments did not provide information to bring about '

j closure to the autoignition and consequences of local detonation issues and industry plans no further tests.

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I Elimination of Advanced Nuclear Plant Analyzer Development (Reduction in FY 1986 - 31290, and in FY"1987 - 3600K)

1. Safety Issue: The NPA program, which began in 1983, addresses the NRC need to be able to perform rapid analysis of transients in operating power plants. The second main purpose of the NPA program is to reduce the costs to the NRC of performing analyses by improving the productivity of analysts costs.

and by increasing the calculational speed to . reduce and the time from when a calculation is begun to when the analysis is completed is relatively long. The preparation of input is time consuming The interpretation of and requires considerable experience and expertise.These factors combine to limit output requires similar resources. The advanced NPA of the code to a relatively limited set of experts.

allows more safety analyses to be done at less cost.

2. Current Treatment by NRC: Calculations of transients are currently performed using the NRC thermal hydraulic codes with the existing NPA onThe mainframe computers located at the DOE national labs. Use to perform a calculation and analyze the results is relatively long.

of a DOE lab mainframe computer also involves time sharing with on-site users who do not understand NRC urgencies and it involves telephone line problems.

Role of Delayed Research: Development of an advanced NPA would involve 3.

the use of dedicated m1nicomputers to permit simulations of operator The actions and variations in plant equ1rment responses to be performed.

NPA work included software improvements to take advantage of developments The in advanced computers such as vectorization and parallel processing.

NPA also makes the interpretation of code calculations more readily assimilated by visually displaying results.

The NPA development was

4. Prospects for Work Being Done by Industry:

associated with the NRC thermal hydraulic computer codes, therefore, there is no prospect of industry support.

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Follow-on Testing in the Multiloop Integral Systems Test (MIST) Facility (Reduction in fY 1987 - 53.5M)

1. Safety Issue: Operational transients in B&W plants (Rancho Seco, Davis-Bessie, Crystal River) have indicated that B&W reactor systemsNRC are significantly more sensitive to system upsets than other PWR's.

safety analysis codes now have a limited ability to.p'redict the outcome of B&W plant transients and accidents.

2. Current Treatment by NRC: System transients and accidents in B&W reactors are now analyzed using tne RELAP and TRAC codes which are notConsequently, configured for the once through steam generators used in B&W reactors.

appropriate correction factors must be used. However, because of system

- volume and piping arrangement differences from the C.E. and Westinghouse designs for which TRAC and RELAP were configured, safety analysis results are characterized with a greater degree of uncertainty than desired.

3. Role of Deferred Research: Test data from an experimental program in accident recovery techniques, feed and bleed cooling, steam generator tube rupture, feedline/steamline breaks, and loss of main feedwater accidents in 3&W reactors was planned in MIST for verification of the TRAC and RELAP codes to be able to analyze B&W plant upsets. This program has been A delayed for at least a year until appropriate funding can be secured.

test program in MIST in FY 1986 will provide small break loss of coolant acefdent data. The above program would be follow-on to the 1986 program if funding can be identified.

4. Prospects for Wo rk Being Done by Industry: Industry (B&W, B&W Owners and EPRI) participated in the current test program in MIST but have indicated that they see no regulatory or safety need to participate in the follow-on program. NRC staff (RES and NRR) are currently discussing options to gain industry support, especially in light of the current safety evaluation study being perfonned by the B&W owners group.

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Reduced Steamline/Feedwater-line Break Tests in Semiscale (Red

- 51,500K, in F_Y 1986 - 5500K)

Safety Issue: Significant uncertainties exist in the safety analysis

1. codes ability to calculate reactor response to feedwater-line/steamline breaks.

These uncertainties affect NRC's ability to evaluate such effects as: 1) thennal shock to the reactor vessel under pressure in the event of steamline break; and 2) coolant system overpressurization during feedwater line breaks. An example of the problem faced by NP.C is that the CE

" conservative" primary to secondary heat transfer assumed for feedwater line breaks is about the same as experimental data from semiscale tests.

This indicates that the CE System 80 primary coolant system overpressurization calculated by the vendor for this " conservative" case may in fact be realistic or even non-conservative and could represent a serious safety problem.

2. Current Treatment by NRC: To assume a conservative approach in dealing with feedwater-line/steamline breaks (f.e. to assume that more overpressurization occurs than is likely to be the case).

A wide variety of break sizes, break

3. Role of Deferred Research:

locations, reactor coolant pump operations (on/off), and recoveryThe number techniques 14 to 5.

were to have been tested.These additional tests would have For instance, a requirement to leave degree of uncertainties in the code.the primary coolant pump ru without good backup da'ta.

The industry claims its

4. Prospects for Work Being Done by Industry: Industry (C.E. and W) current procedures are adequate for safety. facilities to do this kinl  ;

years.

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Reduced Steam Generator Testing (Reduction in FY 1986 - $600K, in FY 1987 -

3600K)

1. Safety Issue: For such safety issues as the reestablishment of natural circulation after a small break loss of coolant accident (S8LOCA),

steamline or feedwater-line breaks, and pressurized thermal shock (PTS) the contribution of steam generator phenomena to the . recovery of the reactor system to a safe state is poorly understood.

2. Current Treatment by NRC: Current steam generator models in the NRC's thermal hydrau11c safety analysis codes TRAC and RELAP are thought to be conservative; however the heat transfer assumed may be conservative (lower than realistic) for reestablishment of natural circulation, but would be non-conservative (giving lower temperature gradients) for pressurized thermal shock events.

3. Role of Deferred Research: Tests would have been conducted in the MB-2 "

steam generator test facility to examine steam generator behavior under natural circulation conditions, to determine heat gain / loss contribution in an overcooling transient (PTS), and to conduct tests under This data would be used to verify feed-line/steamline break conditions. These codes could or modify the steam generator models in TRAC or RELAP.

then perforin more realistic safety analyses allowing for more assured regulatory decisions.

Prospects for Work Being Done by Industry: Industry feels that the 4.

current ability to analyze these events is adequate to prove safety and. it is prepared to live with the conservative (or non-conservative) computer models. The MB-2 facility has been shutdown for over a year and may be partly cannableized.

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Heavy Section Steel Technology Program. Pressurized Therwal Shock Research (Reduction of 19M in FY 1987)

Safety Issue: The safety issue is whether the stainless steel cladding of a vessel helps to inhibit crack extension under PTS conditions, or if the l l

1rradiation embrittlement has caused enough change to promote crack extension, thus making crack extension worse leading to deep cracks or penetration. ,

Current Treatment by NRC: Currently, the NRC assumes that the cladding is tough enough and will remain sufficiently so despite irradiation that it will We are conducting zot promote through-wall crack er. tension and cracking.

" separate effects" tests of small irradiated specimens to learn of the cladding embrittlement and of crack extension under the clad.

Role of Deferred and Cancelled Research: The cancelled test would have presented a realistic assessment of the effect of cladding, complete with the true stresses including pressure-driving forces. Without the integral vessel test of PTSE-3, we will never really be sure if or how much the vessel clad will inhibit crack extension or even promote crack extention. If the inhibition effect had been exceptionally high, we might have been able to safely reduce our margins already in place on PTS.

Prospects of Work Being Done by Industry: It is extremely unlikely that industry will undertake this work, primarily because of the cost, even though they would benefit greatly from the positive result. Industry has shown little inclination to do much materials work on PTS thus far, and no change in .

attitude is seen.

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Heavy Section Steel Technology Program Integration Center for HSST (Reduction of 30.5N 1n FY 1987)

Safety Issue: The issue is that we will continue to suffer by not having a large, ready capability to integrate research results and translate them into application to regulatory problems. ,

Current Treatment by NRC: Currently, the technical integration is performed by the REs staff which is quite limited in numbers.

Role of Deferred and Cancelled Research:

The integration center would have greatly expanded the capability to analyze activities and results, to bring them into clear focus, and to have the personnel available to sunnarize and set forth the application of regulatory issues. The RES staff controlling the center could then define large issues and provide appropriate critique of the center's work. ,

Prospects of Work Being Done by Industry: This kind of activity is totally outside the appropriate area for industry.

12

Structural Intsority of Water Reactor Pressure Boundary Components (Reduction of 3500K in FY 987)

Safety Issue: The safety issue is whether the current S-N (stress versus number of cycles) fatigue design rules, which are based on smooth polished specimens tested in air at room temperature, accurately represent as-fabricated anterials in the coolant environment at operating temperatures.

Current Treatment by NRC:

The NRC currently must accept the S-N curves in the A5Mt Code,Section III.

By performing this work, we will have Role of Deferred and Cancelled Research:

a better idea of the effect of the reactor environnent on fatigue properties of pressure vessel steel, an issue not treated by the ASME applications for license extension beyond the present 40 year plant life.

There is industry interest in this Prospects of Work Beinn Done by Industry:

"irea, based on the inv tation of ASME-III to RES staff responsible for thiswork to York. Whether they would pick up such work if cancelled by the NRC is not clear; based on past performance, however, that is unlikely.

13

Leak Detection-Program (Reduction of $370K in FY 1986) l Safety Issue: Accuracy and reliability of current comercial leak detection systems wnien are relied upon for leak-before-break.

Current Treatment by NRC: It is presently assumed that in-place leak detection systems are capaDie of detecting the large volumes of water that would leak from a sub-critical crack.

This work was intended to antify Role of Deferred and Cancelled Research:the accuracyor couldof current in-place not) detect, in a reliable way, the smallest volume of water that related to a crack that was of concern to safety; and to improve the equipment and techniques of leak detection to achieve the reliability, location accuracy and leak sizing capability needed to back up the rule changes on reliance on leak-before-break in piping systems. The work supports recommendations for improved leak detection systems made by the Piping Review Comittee.

Prospects of Work Being Done by Industry: Only TVA and the Philadelphia D ectric Company have shown some interest in improving acoustic emission leak detection which is the basis of this program. TVA is trying to join the PNL program on acoustic emission continuous monitoring for broader scopeIt is unlikely th application of the technology.

work.

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14

Heavy Section Steel Technology Program de-Plate Crack Arrest Test and K,_ Shift and Intermediate Test Vessel Effort, W ""

Shape Change with Irradiation (Reduction of 3400K fn FY 1986)

Safety Issue: The safety issues are: (1) an understanding of fatigue crack growCU to failure in a nozzle corner, (2) understanding crack arrest to assure accurata predictions under PTS conditions, and (3) to preclude non-conservative predictions of embrittlement in vessels based on Charpy-V surveillance test.

Current Treatment by NRC: Currently, the NRC relies on the ASME-XI crack growth rate curve for nozzle corners, but this is not applicable because of the triaxial stresses in the nozzle (The curve, however, is developed from typical Crack arrest is currently measured by small specimens of biaxial stress load).

relatively thin sections; testing and analysis methods have not been validated.

Presently, it is assumed that irradiation embrittlement follows the shape of the ASME Code reference g K , curve, so that radiation induced shifts can easily be translated from Charpy curve to fracture toughnesss curve shift.

• The proposed ITV-10 test of the Role of Deferred and Cancelled Research:

nozzle corner crack would provide conclusive evidence of how to calculate and evaluate repairs.

such crack growth, and thus allow some more freed of arrest, because the thick sections accurately represent material behavior in The code reference fracture toughness curve must be operating reactors. validated using irradiated drop-weight specimens to validate the ir NDT, and thick compact tension specimens must be tested to compare with Charpy-Y correct.

specimens to assure that the assumed shift in embr could be made from the Charpy data.

The industry probably sees these Prospects of Work Being Done by Industry: items as necessary fo validity and safety inherent in regulations.

industry would undertake to do these items.

0 15 i

Facilities Undergoino Decomissioning (Reduction of 1150K ' n PY 1986)

Safety Issue: Data needed to evaluate licensee funding assurance and decomissioning plans and occupational exposure should be obtainad from actual reactor decomissionings to assure sufficient funds are available to complete decomissioning and that it be done safely according to plan.

Current Treatment by NRC: Current NRC decommissioning actions are taken on a case-by-case basis. The proposed decomissioning rules will require licensee to submit the above plans and NRC to evaluate them.

Role of Deferred and Cancelled Research: Additional data on occupational exposure, waste volume, manpower requirements and decommissioning techniques would have better established the data base used to evaluate decomissioning plans.

Prospects of Work Being Done by Industry: None. The data would be used

.primarily by NRC staff or NRC contractors.

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I 16

Performance Evaluation of Electrical Equipment Durino Severe Accidents (Reductions of 5170K in FY 1986 and 3700K in FY 1987, Safety Issue: To confirm the adequacy of data obtained from plant instrumentation and the availability of electrical equipment under severe ,

accident states which provide the basis for operational and emergency preparedness actions; and to provide more accurate data foV deterministic and probabilistic calculations as they pertain to severe accident states.

Current Treatment by NRC: The Connission has requested that the staff conduct research to determine if regulations should be written covering severe accidents which are beyond the design basis accidents. The information generated in this program will help formulate new regulations should they be required.

The purpcse of this research is to Role of Deferred and Cancelled Research:

determine tne performance of existing plant instrumentation and electrical components under severe accident environments.

The research will be utilized The research by the NRC in establishing a final policy for severe accidents.

data will help to determine the performance of instruments and electrical.

equipment under severe accident conditions and help to identify cost effective ways to reduce the consequences of a severe accident as called for in the NRC Severe Accident Policy Statement. Specific work to be accomplished under these two projects included testing of a main steam isolation valve control manifold assembly, completing the evaluation of the performance of plant effluent monitoring system, and evaluating other identified electrical equipment, all under severe accident conditions.

Industry has been invited in the Prospects of Work Being Done by Industry:They have generally taken the position th past to share in this work.

would spend their money on operational problems.

1 0

4 17

Fire Protection Research (Reductions of 3.3M in FY 1986 and $.7M fr FY 1987)

Safety Issues: The requirement of separation of the redundant trains of safe shutdown equipment prescribed by 10CFR50, App. R cannot be met in many control ,

rooms. The rule requires in such cases that an alternative or dedicated l shutdown capability be provided. Whether in the event of a credible control j room fire the operator will have time enough and the physical ability to {

transfer control of the reactor to the alternative or dedicated shutdown panel t is at issue. There is also the question of how much of the control equipment will survive and whether and when control room operation can be resumed.

PRAs are developing into an important tool for identifying safety-deficient areas of plants and for evaluating the effect of back-fits on the overall safety in terms of core-melt frequencies. Large uncertainties in the risk from fires, because of the sparse data base, have been noted. There is a need for experimental data on the fire fragility of safe shutdown components and for an analytical methodology for estimating the environment to which such components may be subjected in the event of a postulated fire.

Current Treatment by NRC: Implementation of Appendix R requirements provides for exceptions when the licensee can show that the proposed alternative is as effective as a particular provision of the rule. The staff often lacks specific technical data that supports either accepting or rejecting a licensee request for exemption. Staff licensing decisions are then based upon best engineering judgment and knowledge gained from commercial and industrial fire protection experience and pract1ces.

This research would have provided the Role of Deferred and Cancelled Research_:

staff with test data on credible nuclear plant fire sources (e.g., cable tray formations electrical cabinets, etc.), on potential full-scale fire environments in plants, and on the thresholds of failure of some of the more fragile components. The program would also have provided the staff with an analytical tool (a computer code) for extrapolating to various plant-specific rooms and equipment layouts.

No similar program is being

Prospects of Work Being done by Industry: As far as we can determine, they see

' sponsored by the nuclear power industry.

no possible benefit from this type research and will not undertake such work in the future.

l 18

Ecuf pment Qualification peductions of 1.5M In FY 1986 and $1.9M in FY 1987)

Safety Issue: Safety related electrical and mechanical equipment must be able to survive various accident environments (loss-of-coolant, hydrogen burn and i earthquakes) and be able to perfom their functions of shutting the reactor down, cooling the core and containment, monitoring and mitigating accident consequences.

Current Treatment by NRC: Safety related electrical equipment is required to l be qualified for a loss-of-coolant accident by the rule 10CFR50.49 and for a

' hydrogen burn by the rule 10CFR50.44. Safety related mechanical equipment is required to be qualified for a loss-of-coolant accident and for earthquakes by the Standard Review Plan (Sections 3.10 and 3.11, respectively). NRR is i currently reviewing the licensees equipment qualification submittals using standards and criteria which are Lased on engineering judgment and have in many instances not been thoroughly validated.

The equipment qualification research Role of Deferred and Cancelled Research:

cancelled was to confirm the adequacy of the methods and standards employed by industry to meet the provisions of the rules and resolve NRR/IE questions identified in inspections at operating plants. Open issues will not be totally resolved, such as (a) does the NRC source term research indicate the need for changes to the equipment qualification rule. (b) should industry (be required toc) me address synergisms in qualifying accident monitoring equipment, d

correlating seismic qualification and fragility are(e)ata, the BWR(d)utilities the effect of high efforts frequency vibrations on electrical equiprent.

adequate to demonstrate the survi. val of equipment in a hydrogen burn relying solely on analysis, and (f) will electrical penetrations maintain containment integrity under a severe accident envirorment? The rules and qualification efforts of industry are based in many instances on engineering judgment.

Failure to complete the research leaves the future: performance of equipment in a TMI-2 type of event significantly more uncertain and increases the risk to

the public.

The research on equipment qualification wculd have provided the basis for resolving many uncertainties in the methods and added assurance that equipment would be available and function during and following accidents. It would thus I

provide the basis for revisions to the rules and regulatory guides.

Prospects of Work being done by Industry: The utilities, vendors and EPRI have essentially no significant ongoing reseirch effort in equipment qualification to resolve gliestions and confirm the methods, other than (a) the Seismic Qualification Utilities Group (SQUG) which plans to provide a method for l

i assuring the ruggedness of specific categories of equipment in operating plants by the use of experience data, and (b) the BWR Mark III Utilities Hydrogen The Control Owners Group (HCOG) quarter schle experiments at Factory Mutual.

HCOG experiments have purposely omitted the inclusion of equipment and rely ,

solely on approximate analytical methods for predicting equipment thermal '

response. The complimenting NRC equipment themal resporsse research tests, to be used to provide data to evaluate the survival of equipment, will have been cancelled. ,

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' i Reliability Analysis of Nonlinear Behavior of Concrete Structures _

(Reduction of 3200K in FY 1986)

Safety Issue: Reliability analyses are being performed by NRR to evaluate the 5eismic Category I Structures subjected to multiple static and dynamic design loads. 'Since structures generally respond linearly to these design loads, the method of reliability analysis used, the Reliability Analysis of Structures (RAS) Code, is based on linear assumptions. The response of structures to severe accident loadings will be substantially nonlinear. Therefore, if reliability analyses of Seismic Category I Structures are to be perfomed for these loads, the method of analysis must be expanded to consider nonlinear -

effects.

Current Treatment by NRC: The Seismic Safety Margins Research Program (SSMRP) has provided a simplified, linear analysis that is available for the NRC staff use if needed to approximate linear structural response for seismic risk l analysis. The RAS computer code is being used by NRR, but is limited to linear The RAS enables NRR to evaluate the safety behavior of concrete structures.

margins of nuclear structures under various static and dynamic loads and to generate the structural fragility curves for PRA studies. Since many structures will respond nonlinearly to severe accident loadings, NRR is unable to confidently judge the adequacy of structural safety urgins or fragility curves for such loadings until the RAS code methods have been improved to treat nonlinear effects.

NRR needs a method to perfom Role of Deferred and Cancelled Research:

reliability and margins studies of structures responding nonlinearly to loads beyond design levels. Without it, NRR will not have the tool they feel is necessary to evaluate nonlinear effects on the reliability and fragilities of

! Seismic Category I concrete structures. This would leave NRR unable to adequately confim structural safety margins in these load ranges.

Prospects of Work Being Done by Industry: The results of this research will be used to validate structural safety margins calculated by licensees to exist l

' under a variety of loads and load combinations. It willIt provide the NRC an should, therefore, be independent means of reviewing structural behavior.

sponsored and developed by the NRC to retain independence from bias due to practices of any design fim. It is expected that eventually the results of i

' this research, as well as the probabilistic load combinations research that I preceded it, will be adopted by national structural design standards.

20

Nuclear Materials Transportation (Reduction of $800X in FY 1985 and FY 1986; no funds for FY 1987)

1. Safety Issue:

Questions have arisen in connection with the degree of protection being provided by NRC regulations applicable to the transportation of radioac-tive materials. These questions relate to the comparison of packaging and transportation performance requirements to accident forces which would be experienced in an actual accident. The requirements at issue deal  !

principally with the shipment of large quantities of radioactive I materials, in particular, spent fuel.

2. Current Treatment by NRC:

Licenses for casks and packages used to ship spent fuel and larger quantities of radioactive material are issued on the basis tofofdemonstrated deter-package compliance, either b ministic performance tests (y test or analysis, These with a seimpact, fire, pun corresponding limits on the post-test release of activity.

requirements are set for the in the Packaging and Transportation of Radioactive Material Regulation (10 CFR 71). This regulation and equivalent international transportation regulations have been in use by the Commission and IAEA member nations for almost 20 years. The potential environmental impacts of a spectrum of transportation accidents appropri-ate for each mode of transport (air, rail, and road) have been prepared in the Environmental Statement on the Transportation of Radioactive Materials This statement has concluded, and by Air and Other Modes (NUREG-0170). -

the Commission has concurred, that the risks to public health and safety resulting from the shipment of radioactive material under current regulctions does not warrant any changes in those regulations.

L

3. Role of Deferred Research:

Because of the anticipated increase ist future shipments of spent fuel and waste and in response to the continuing public concern regarding such shipments, NRC initiated a program to develop and document the degree of protection regulations.

being provided by spent fuel casks licensed under existing specific accident forces to which a shipping cask might be subjectedIfand show corresponding estimates of the release of radioactive m=terial.

it was detemined that the protection being provided by the! . casks in certain extreme severe accident environments could not be established with sufficient certainty, it was planned to carry out a series of experimental tests of large shipping casks (or components) to develop and validate models which could predict cask perfomance under those accident conditions. These. programs would have produced a validated, systematic analysis of transportation risk which would have reduced the potential for litigation in licensing of intersite transfers of fuel shipments to a monitored retrievable storage system or to a waste disposal facility.

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21

A Nuclear Materials Transportation (continued)

Because ci continued budgetary pressure, the' experimental portion of this program has been cancelled and reliance will have to be placed solely on calculational analyses.

4. Prospects of Work Being Done By Industry:

Because the results of this work will be used directly to fonsulate revisions to the Commission's regulations, there is no real incentive for industry to conduct such a program, nor would it be prudent for NRC to rely exclusively on information developed by potential licensees.

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Safeguards (Red'uction of $500K for both FY 1985 and 1986, with no funds for l FT 1951) l

1. Safety Issue: l The imposition of physical protection requirements set forth in 10 CFR 73 '

(Physical Protection of Plants and Materials) depends on NRC staff assessment of the likelihood of adversary actions There againstare such veryfacilities large and on the potential consequences of such events.

uncertainties in both these parameters, particularly with respect to consequence estimates. These uncertainties have the potential for leading to contested ifcensing actions is the areas of independent fuel storage and the transportation of spent fuel and high level waste, where the principal contention is the adequacy of current safeguards regulations.

2. Current Treatment by NRC:

Requirements for the physical protection of spent fuelThose at independent requirements storage sites are currently contained in 10 CFR 73.

were originally developed for a broad range of materials and facilities, and were not developed specifically to deal with problems inherent to independent storage of spent fuel and waste, e.g., an Independent Spent Fuel Storage Installation (ISFSI). Preliminary studies of the conse-quences of adversary actions, some of which relate to transportation systems, suggest that some of the current physical protection requirements '

applied to independent fuel or waste storage activities may not be appropriate and may require modification.

3. Role of Deferred Research:

The deferrad research would provide an improved technical basis for revising the regulatory safeguards requirements for ISFSI (and waste stor-age) facilities as well as for developing perfomance The deferred safeguards goals and establish-research ing criteria for acceptance performance.

would have assessed IFSFI sabotage consequences by examining results of latest spent' fuel sabotage studf as, comparing shipping cask designs to dry storage containers designs, evaluating possible radioactive release An fractions and form, as well as establishing maximum credible events.

experimental program would be carried ou; to provide any Other required deferred validation of the assumptions used in these assessments.

safeguards research was directed toward sup;iorting the update of severa obsolete physical protection guides to include the latest. state-of-the-art technology on physical protection devices.

4. P_rospects of Work Being Done by Industry:

i The results of research in this area will be used directly to fomulate safeguards regulations. For this reason and because of the classified nature of such work, RES sponsorship of such studies, in cooperation with DOE, is the only practical approach to obteining the required information.

23

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Reactor9 1986, 0)erations and Risk (Human Factors) (Reductions of $2.0M with no funds for FY 1987) 51.5M in

1. Safety Issue: -

The capabilities of reactor operators to respond correctly to severe  ;

transients or other abnomal conditions can be assessed only in a very Best estimates of the contribution of human '

general, qualitative manner. error to reportable events in operating reactors ~

percent of all events, thereby confirming the importance of human reliability in ensuring the safe operation of nuclear power plants.

2. Current Treatment by NRC:

To better ensure an adequate response of plant personnel to operating transients and to maintain a high level of perfomance in the maintenance and repair of equipment, NRC currently requires the licensing of some of the plant operating personnel and has established requirements forIn additio reviewing licensees' management functions.

2 mented the Systematic Appraisal of Licensee Perfoman NRC has also required the installation of and corrective actions taken.

an improved safety parameter display system to assist plant operators in recognizing and analyzing plant transients or accidents.

There are, however, no objective methods for assessing the effectiveness of current or proposed regulations applicable to human reliability or for measuring the performance of plant operating and maintenance personnel.

The clear need to replace the subjective assessment methods now in use with more obiective methods of performance has been detailed in the Human Factors Program Plan, NLEG-0985.

3. Role of Deferred Research:

Current implementation of the severe accident policy, the resolution of outstanding safety issues, and the analysis of operating events all It involve, to a substantial degree, the element of human reliability.

has become quite evident inAccordingly, the analysis of reactor t a research program the event or mitigating its consequences.

was developed which would have to provide a technical base for measuring and evaluating the effectiveness of operator actions in responding to This research included work in the areas of operator reactor accidents.

training and licensing, development of operating procedures, control room design and plant maintenance.

l Completion of this research would have permitted a more systematic a comprehensive regulatory approacn to be taken with respect to establ l

human reliability requirements and would have allo I

engineered safety features in resolving outstanding safety issues and '

other plant problems as they arise.

I 24

Reactors Operations and Risk (Cen't)

4. Prospects of Work Being Done By Industry:

Currently (, there is a large program on human factors being industry Institute for Nuclear Power Operations, Nuclear Utilities Management and Resources Committee, etc.). The deferred program of RES would have provided an independent NRC audit capability to verify the results and insights developed by the industry program. .

0 25

(Reductions of $1.1M in both FY 1985 and 1986,

' Severe Accident Risk Analysis with no funds _for FY 1987)

1. Safety Issue:

Implementation of the Commission's severe accident policy and the relat requirement to effectively interface with industry's seve the availability of independent NRC technical positions on the risks presented by the current generation of LWRs and on a number of specifi topics which significantly impact such estimates.

containment perfomance and fission product behavior in severe reactor accidents, the risk contributions of common cause failures and severe natural phenomena (e.g., saismic safety margins), and accidents.

The positions taken by NRC staff on these topics will have a profound influence on the future requirements for backfitting engineered safety features on operating plants, on the resolution of outstanding safety issues, and on future revisions to emergency planning regulations.

2. Current Treatment by NRC:

Licensing considerations of severe accidents in the current generations of LWRs are specified in 10 CFR 100 (Reactor Site Criteria) and These 10 CFR 50 (Domestic Licensing of Production and Utilization Facilities).

regulations require confomance These requirements with a deterministic, nonmechanistic, were initially intended to design-base accident.

provide an adequate margin of safety for the public Since in theTMI, event of a very severe accident, but one which did not involve core melt.

questions have been raised as~ to whether core melt incidents should be explicitly considered in Comission regulations and, further, whether the risks presented by currently licensed facilities are such as to warrantThe Com backfitting additional safety features on these plants.

position on this question has been set forth in its policy statements on severe accidents and backfitting.

The development of plans to implement the severe accident policy requires the use of a combination ofCurrent complex computer computer codes used for code inception to final impact ori the public.

these purposes are not integrated, are expensive to run, and do not adequately incorporate the latest research results on, for example, comon l cause failures or external events--nor do they provide an explicit estimate of the uncertainties involved.

I 3. Role of Deferred Research:

The RES risk methods development program has been directed toward two I

objectives: (1) to develop methods for better assessing accident frequency by incorporating common cause failures, improved human error i

models, and severe natural phenomena into reactor risk assessments; and (2) to develop fast-running, fittegrated computer codes capable ofThese l efficiently modeling accident phenomena and effects.

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L_- -. . _ - - - . - - . - . . . - - _- __ -- - - - _ _ _ _ _ . . - _

Severe Accident Risk Analysis (continued) provide staff with competent estimates of the probability of core damage, the characteristics and distribution of fission products in reactor containments, and the subsequent release RES hasandbeen dispersal able, of radioactive over the last contaminants into the environment.

several years, to maintain a coherent program in this area in the face of budget reductions. We have, however, been compelled.to stretch out these programs to the extent that we are not yet able to apply reactor plants. This situation has resulted in an added expenditure of over $1 million, to provide required source term and consequence estimates using older, more manpower-intensive codes, and the final results will have a large range of uncertainty than would be present had deferred research been performed.

4. Prospects of Work Being Done By Industry:

Because of the fact that the analyses being developed in this program will be used to implement the Commission's Severe Accident Policy and similar analyses are being carried out by the industry's IDCOR program, there is no possible alternative other .than carrying out an independent NRC analysis of accident risk.

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27

High-level Radioactive Waste Research (Reductions of $6.9M in FY 1986 and $3.8 in FY 1987)

There Safety Issue: A HLW geologic repository is a first of a kind facility.

is little previous experience in science and technology thatIndeals orderwith to the specific, long-term questions of geologic disposal of HLW.

effectively discharge its responsibilities as regulator of. DOE's HLW disposal under the Atomic Energy Act and the Nuclear Waste Policy

1) DOE Act, the compliance NRC with NRCmust have an independent technical base to review and assess:and EPA regula Current Treatment by NRC: NRC presently is undertaking pre-licensing consultations with DOE on the three media which DOE indicates it is likely to select for site characterization: basalt, tuff, salt. Research specific to each of these media has been supporting those consultations in the area of geology, geochemistry, rock mechanics, However, due to budget geohydrology, cuts in FY 1986 materials properties, an and 1987, our perfonnance assessment. ability to keep up with DOE on such a broad technic The absence of research results could become felt beginning in 1987 during site characterization and may ultimately hinder review of DOE's license application in 1991.

By taking full and judicious Role of Deferred and Cancelled Research:

advantage of the fact that research often takes longer a the middle of the fiscal year, RES was able to maintain the HLW researchCutting the program at an effective spending rate of $5.0M durino FY 19  ;

precipitous and premature termination of all basalt repo) l of HLW, and indefinite deferral of all work pertaining to the s repository.

repository components and systems, which will contribute to NRC's review DOE's demonstration of safety, will be significantly drawn out and delayedFinally,l until well into the first licensing review beginning in 1991.  !

$3.0M level for FY 1987, RES will not be able to perform any additional or new l l

research on generic issues not already identified which may arise during the ;

pre-licensing consultations with DOE or during the site characterization.

If DOE would perform research that allows Prospects of Work Being Done by DOE _: critical scrutiny of its own te application of the results of that research would lack the degree of independence that NRC needs and obtains in sponsoring this research.

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Low Level Radioactive Waste Research (LLW) (Reduction of $1.8M in FY 19 Safety Issue: ~ Final disposition and closure of existing non-operating Currently L;y shallow land burial disposal sites is yet to be accomplished.

operating LLW disposal facilities will require closure in the next few years.

Most States are considering engineered enhancements andand

'riividual States alternatives to State Compacts conventional shallow land burial (SLB).

are making siting, design, and other deci*:ons involving public health and safety concerning LLW disposal now and are relying on the NRC for technical guidance.

Current Treatment by NRC: Monitoring programs and criteria for facility closure decisions are not yet finalized. NRC recently completed a study to assess the applicability of the Licensing Requirements for Land ~ Disposal of No Radioactive Waste (10 CFR Part 61) to regulating alternative LLW disposal.

generic safety related siting, design, performance, or operational criteria However, some potential design gaps pertinent to were found to be needed.

licensing alternatives were identified which will require further work.

Performance assessment tools for LLW disposal primarily deal with SLB and associated radionuclide transport in soils. Scurce terms, performance of LLW waste forms and packaging, monitoring techniques, and performance assessment techniques should be upgraded to deal with engineered enhancements.

The research deferred by the reduction in FY 1986 Role of Deferred Research:

funding would have begun the development of design criteria for engineered enhancements and alternatives to SLB, completed development of design criteria for engineered enhancements and alternatives to SLB, completed development of statistically-based monitoring protocols for SLB of LLW, and begun development -

of test and measurement criteria to facilitate closure decisions. These efforts will be delayed until FY 1987, making guidance to the States and State Compacts significantly less timely and less useful.

Frospects of Work Being Done by Others: Industry is augmenting the technical These efforts base needed to comply with regulatory criteria for LLW disposal. .

are concentrated in the areas of waste form stability (AIF/NESP) and alternativetechnologyassessment(EPRI). States and State Compacts have >

neither the technical base nor, individually, the financial ability to-support research efforts and are concentrating on the development of disposal capacity.

DOE is doing limited LLW disposal research. DOE has concentrated on institutional matters and has provided limited technical support to States.

The current program places increasing emphasis on technical matters, such as alternative technology disposal. However, the Low-Level Radioactive Waste Policy Amendments Act of 1985 explicitly charges the NRC to provide technical information on LLW disposal alternatives to the States.

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Radiation Rist Management (Reduction of 1.570K in FY1987 for Termination of the Research Program)

1. Safety Issue: Major scientific decisions of the NRC are made using the best data base available; such decisions often control the management In many casesof radiation risks associated with NRC-licensed operations.

the adequacy of the data base is an important issue..and is often the deciding factor.

Current Treatment by the NRC: Where a deficient data-base can be 2.

significantly improved at reasonable cost, NRC research and technical assistance funds are used to support risk management decisions.

Currently available data for developing

3. Role of Deferred Research:

residual radioactivity criteria for a regulation on the This situatf on decommissioning prevails also for of lands and structures are inadequate.

certain implementation aspects of the proposed revision of Standards for Protection Against Radiation (10 CFR Part 20) and for theThe renagement deferral ofof releases of radionuclides into sanitary sewer systems.

support funds will curtail the solution of certain Part 20 problems, will terminate residual radioactivity research, and will prevent solution of

'he sanitary sewer disposal problem.

None.

4. Prospects of Work Being Done by Industry.

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30

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Occupational Radiation Protection (Reduction of 3710K in FY1987 for Termination of Research Program)

Safety Issue: The major reason for conducting a governmental licensing 1.

program for operations involving potential radiation exposure is to ensure The proper control of the radiation risks for those who could be exposed.

NRC licensing offices who authorize such operations are assured of adequate worker protection 'through a dual program involving the development / adoption of regulatory standards and periodic, on-siteThe s inspections against these standards.

necessary by guidance documents to assist inspectors and licensee compliance problems may arise.

standards includes basic recommendations from scientific advisory organization (e.g., ICRP, ICRU, NCRP) as well as consensus standards (HP Other informational needs often require NRC 4

ANS.IEEE, resources: ANSI, ASTM).(I) research results; (2) the number of workers b and the magnitude of these exposures; (3) the nature and extent of problems adversely affecting the agency's risk control effort; and (4) the cost of alternative solutions.

2. Current Treatment by the NRC: To the extent possible reliance is placed on advisory organizations, consensus standards, Federal guidance developed under the EPA mandate, and research funded by outsid through(4)above. An interagency-funding approach'is used if indicated.

Due to prior RES budget constraints the budget

3. Role of Deferred Research:

request for FY 86 was reduced from $1310K to $600K, and no f .

area was requested for FY 1987. This request

research projects considered to be of interest to DOE also.Had these was subsequently declined.

performed, the results would have been used in th of NRC worker-protection responsibilities.

Within the nuclear power

4. Prospects of Work Being Done By In'dustry:

industry extensive research f s befng funded in the ALARA engineering area because dose reductionHowever, through engineering technology has been found to little or no research in health physics reduce operating costs. technology is funded by the utilities or by associa EEI,EPRI,INPO). No assistance for the NRC is expected from this quarter.

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Radiation Risk Assessment (Reduction of 3630K in FY1987 for Termination of Research Program)

Safety Issue: Regulatory actions for controlling radiation risks are 1.

based to the extent possible on current scientific information regarding Gaps in existing knowledge often require the NRC adverse health effects.

to base its decisions, at least temporarily, on technical assumptions.

Subsequent requirements may therefore provide inadequate personnel protection, or they may be overly restrictive and costly.

Current Treatment by the NRC: Where informational gaps of this nature are 2.

specific to NRC-11 censed acfivities, so that other Federal agencies are not prepared to fund the necessary research, needed research projects are funded by the NRC, and the results are used in NRC decision-making processes regarding regulatory standards and practices.

Gaps of the type mentioned above include the

3. Role of Deferred Research:

behavior and effects of certain radionuclides following deposition in the body. Since the necessary research has been deferred, regulatory controls will continue to be based on assumptions.

None.

4. Prospects of Work Being Done by Industry:

e 32

.....m.,__ -. - , _ . . - - . _ _ . . - _ _ _ _ , - , - _ _ . _ . _ _ -.____..___-_y__ ,,__-..-.,_._-.___--_._.._..-_._..___.,_,_..__,_-__.._._m. -

a Support of Advisory Groups (Reduction of 34004 In PY1987 for Termination of the Program)

Safety Issue: Basic radiation protection standards used by the NRC are 1.

recommended by scientific advisory organizations such as the International Commission on Radiological Protection (ICRP), National C I

These organizations are supported by contributions from various user organizations such as the NRC. At issue is whether the NRC should J

continue its contributions.

2. Current Treatment by the NRC: Relatively small annual contributions have ,

been made to these organizations for several years.

NRC support has been eliminated for FY 1987.

3. Role of Deferred Research: NAS may have to The ICMP and NCRP are .11kely to continue their work.

abandon its plans to update radiation risk coefficients, as published by the NAS Committee on Biological Effects of Ionizing Radiation, in accordance with revised dosimetry data.

NAS is attempting to obtain

4. Prospects of Work Being Done by Industry:

funds elsewhere.

l l

l l

i

. l 33

Seismology Research (Reduction of 31.0M in FY 1987)

1. Safety Issue: There is a particular uncertainty associated with estab11shing the design earthquake for NPP's in the Central and Eastern

' U.S. This uncertainty results from a lack of knowledge about Thethe uncertainty seismogenic mechanism for the seismicity in this region.

causes overly conservative design requirements, which'can result in i

hazardous conditions, for example; e.g., rigid piping systems when more flexible ones are safer. Uncertainties can also result in unconservative design assumptions.

2. Current Treatment by NRC: Currently the NRC tries to make conservative 11 censing decisions based upon the best avafiable data. These decisions have usually been well received by the licensing boards and ACRS, and have resulted in the resolution of complex seismic and geologic issues raised by intervenors. This is based, at least in part, on the recognition by these groups that the NRC was engaged in an ongoing earth sciences research program to stay on the forefronts of the science, and thus continues to be well aware of new developments in the earth sciences that might affect licensing decisions.

The Central and Eastern U.S. Seismographic

3. Role of Deferred Research:

Networks will be terminated with FY 1987 funding, i.e., no FY 1988 funds, and, the intended replacement, a U.S. Geological Survey (USGS) managed National Seismographic Network will not be started.

Without the networks, the NRC will stagnate at the current level of knowledge, because there will be no way Thereto will obtain additional be no data to location, judge magnitude and focal mechanism data.

whether an earthquake is an isolated event or part of an increasing swam sequence. Such data will not be available to the Comission and staff for their use in making licensing decisions or to address issues raised by 1.icensing boards', the ACRS or intervenors both for new and existing plants.

4. Prospect of Work Beint Performed by Others: The prospects for this work being performed by in(ustry are very small; numerous attempts ThetoNational elicit i

just partial support from industry have not been sucessful.

Seismographic Network project was an effort to place the responsibilityTo for seismic monitoring and data collection in the hands of the USGS.

accomplish this, the NRC was to provide funds to purchase new equipment while the USGS was to provide support in the form of design, construction i

and operation personnel. Without NRC equipment funding, the USGS will not l

carry out this project.

i 34

Extreme Flood Probability Determinations (Program Terminated in FY 1986 With a JZ50K had been planned for FY 1987.)

Reduction of 5250K.

Safety Issue: The safety margins for the protection of nuclear facilities The 1.

against external flood events cannot now be adequately dete  :

cannot be incorporated into probabilistic risk assessments The ACRS.- inneeded to has particular, quantify the hazard due to external floods.

been requesting research to obtain information to resolve the safety margins issue.

2. Current Treatrent by NRC: At present, the design bases flood levels are The PMF is not, based on the concept of the Probable Maximum Flood (PMF). Instead, as its name would suggest, based on probabilistic methods.The assumption is that deterministic methods are used to assess the PMF.

the PMF criteria used for the design bases flood levels are conservativeThe and adequate, but there is not an adequate basis for this assumption.

NRC cannot now include the external flood hazard in probabilistic risk j assessments.

Role of Deferred Research: The deferred research was to: (1) provide the 3.

information needed to assess the flood protection safety margins at nucle:r facilities, and (2) to develop a methodology which would permit l including external flood events in probabilistic risk assessment (PRA) studies.

Essentially none.

4. Prospect of Work Being Performed by Industry:

O e

l 35 l

• Real-Time Atmospheric Dispersion. Plume Rise and Washout The out year Model Evalua (Program Terminated in FY 1984 wf th a reduction of 5900K.1985, 1986, and 1987 projections had been $350K for each of FY The currently available real-time atmospheric dispersion,

1. Safety Issue:

precipitation washout and plume rise models for dose projections in emergency response situations have not been adequately validated.

Accordingly, there is the potential for an inappropriate, unreliable or inaccurate model being used in an emergency response situation and would produce incorrect dose projections. Wash Risk assessments and other ifcensing been accurately quantified. evaluations in which the models are used The currently available models are necessarily

2. Current Treatment by NRC: )

used for emergency response dose projections by licensees and the NR pending further information on their performance under different /

~ meteorological and other terrain conditions and their accuracy and l reliability.

NRC staff risk assessments involving plume washout continue l to have large uncertainties in a scenario that contributes Conservative significantly assumptions are to calculated potential prompt fatalities. ,

used to compensate for uncertainties in licensing evaluations using th models.

The deferred research was to validate models

3. Role of Deferred Researchi used for emergency responseThisdose projections validation can be  ;

terrain situations in which they will be applied. accomp .

field tests.

Essentially none.

4. Prospect of Work Being Performed by Industry:

b 36

Uranium Recovery Research (Reduction of $2.0M in FY 1985 for Termination of Program)

Safety Issue: Uranium mill tilings at both operating and decommissioned mills remain a significant source of potential radiation exposure to the public.

Current Treatment by NRC: To deal with the major uncertainties in the accuracy and reliability of techniques and instrumentation for assessing the long tem stability of and level of radon emissions from mill tailings very conservative stability requirements and conservative measurement techniques have imposed by the NRC.

consuming.

Data on effluent and environmental monitoring methods, Role of Teminated RES:

equipment, and instrumentation for detemining stability of and measuring The emissions from uranium mill tailings was not collected and analyzed.

collection and analyses of these data would have allowed significant reduction in the level of conservatism now applied to stabilization requirements and measurement techniques for mill tailings.

Since the bulk of tailings exist i

Prospect of this Work Being Done by Tndustry:at decommissioned ra the Federal Government, it is unlikely +. hat industry will pe research.

unlikely that any single mill or combine n!11 undertake the work.

4

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