ML20236L963
| ML20236L963 | |
| Person / Time | |
|---|---|
| Issue date: | 10/19/1987 |
| From: | Thompson R NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| To: | AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20236L948 | List: |
| References | |
| CON-NRC-04-87-365, CON-NRC-04-87-366, CON-NRC-04-87-367, CON-NRC-04-87-368, CON-NRC-04-87-369, CON-NRC-04-87-370, CON-NRC-04-87-373, CON-NRC-04-87-374, CON-NRC-04-87-375, CON-NRC-04-87-376, FOIA-87-721 NUDOCS 8711110147 | |
| Download: ML20236L963 (37) | |
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N UNITEO STATES
. 8.
' NOCLEAR REGULATORY COMMISSION -
WAsHWGTON, D. C. 20006 "f
i 00T 19 ~ 1997
Dear' Offeror:
Subject:
-Sma11' Business Innovation Research (SBIR) Program r.y Phase I Awatts -1 Post Award Notice This is to ' advise you that awards 'have been ma' e under the subject _ Request for -
d
- Proposal for the U.S.: Nuclear Regulatory Commission.-
One Hundred and Eleven.Phas'e I proposals were ' received.. _ Ten firm fixed-price awards have been made' effective September 30, 1987. Specific information pertaining to each awa:M follows:
CONTRACT NUMBER 00NTRACTOR AND CONTRACT TITLE CONTRACT AMOUNT-LNRC-04-87-365.
- Center for Planning and Research,
$50,000.00 Inc.
" Scintillation Fiber Detector foro In-vivo Endoscopic. Internal Dosimetry" NRC-04-87-366 Mikayola International. Inc..
$49,660.00
" Inhalation Screening Dose Monitor for.' Workers at Nuclear Power Plants" NRC-04-87-367-Apo11ed Biomathematics, Inc.
$50,000.00
" Risk Based Analysis of Technical
'V 5: '
Specifications" NRC-04-87-368 Expert-EASE Systems, Inc.
$49,985.00 4
4'
' Automation, Creation and Renoda-1 "-
112ation of TRAC-BWR Input Decks" NRC-04-87-369 -
Fluid Physics, Ind.
$49,923.00
" Severe Accident Phenomena Material Circulation in Reactor Coolant System" NRC-04-87-370 EDA Consultants
$47,891.00 "A l'uziy Set'Aporoach for Seismic Safety Margins Assessment" NRC-04-E7-373 J.P. Langan and Associates / Sentry
$49,500.00 Equipment Corporation 3
" Analysis of State-of-the-Art Methods for Containment Integrated Leak Rate Testing" 0711110147 871109 PDR FOIA
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' CONTRACT NUMBER CONTRACTOR AND CONTRACT TITLE CONTRACT AMOUNT NRC-04-87-374:
CFD Research Corocration
$50,000.00 i
".A Comouter Code for Fire i
Protection 'and Risk Ana~1ysis i
of Nuclear Power Plants"
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- NRC-04-87-375, Kalsi Engineering, Inc.
$50,000.00 l
" Prediction of Check Valve l
Performance and Degradation
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in Nuclear Power Plant Systems" MRC-04-87-376 PDI Technology. Inc.
$49,708.00
" Aging Process in IInclear Reactors" Enclosed is the evaluation of your prososal which details the strengths and
. weaknesses as determined by the evaluation panel.
~
f The Nuclear Regulatory Commission (NRC) appreciates the time and ' effort expended by your organization 'in responding to this RFP and youriinterest in
- our programs. We encourage you to ressend to all notices for services which 3
your. organization is able to provide.
Sincerely, l hh Contract Negot a tion Branch No. 2 Division of Cor Office of Admin (1racts istration & Resources Managenent
)
Enclosure:
As stated 1
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APPENDIX B
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l U.S. NUCLEAR REGULATORY COMMISSION 1
SMALL BUSINESS INNOVATION RESEARCH PROGRAN (58IR-87 )
g PHASE I - FY 1967 PROJECT
SUMMARY
i i
FOR NRC USE ONLY Program Uffice I,1 N Proposal No.
Iopic No.
10 BE COMPLETED BY PROPOSER Name dDn AGOre55 of Proposer Applied Biomathematics. Inc.
100 North Country Road. Bldg. B Secauket. New York 11733
]
{
N; ce anu i s tie of Principal Investigator a
l Tatyana Cinzburg. Project Manager Tatie of eroject RLsk-Based Analysia of Technical Specifications ijcnnoiogy. Researcn inrust, anc/or Potencia,i Conrnercial Appi. cation (Limit to 8 worcs)
Methodolorv. PC software for evaluation of technical specifications, lecnnican Aostract tLimit to 200 worcs j Tha FRANTIC computer code is the most advanced model for addressing time-dependent risk effects 1
essociated with technical specifications.
It is used by the industry for the evaluation of I
upgrades and/or improvements in existing technical specifications and by the NRC staff in responding to the licensee submittals.
The existing methodology has two maj or deficiencies: it does not provide the bounding evaluation ' cf. risk in case of unknown characteristics; it is available only as a research mainframe computer based code without convenient user-interface.
This project is directed towards three major goals:
(1) Add new models and techniques to j
the FRANTIC code in order to improve its capability to realistically describe time-dependent and test-frequency-dependent risk factors. Allow for bounding the risk evaluation in cases when tes schedule is unknown.
(2) Perform case studies to verify the methodology and provide examplea of realistic evaluations.
(3) Make the upgraded FRANTIC methodology available in a user-friendly PC form with convenient user interfaces.
Anticipated hesults/lsnotication of tne Approach (both Phase I and II/ Potential Conrnercial Applicat1on of Research Significaritly improved and availabic in the PC-based user-friendly environment FRANTIC
)
methodology will be provided to NRC and transferred to industry.
This will create for both government and industry a
convenient tool to monitor compliance with or evaluate modifications of technical specifications.
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APPENDIX B l.
U.S. NUCLEAR REGULATORY COMMISSION U
SHALL BUSINESS INNOVATION RESEARCH PROGRAM (SBIR-87)
PilASE 1 - FY 1987 PROJECT
SUMMARY
FOR NRC USE ONLY
~
Program Office llM Proposal No.
Topic No.
10 BE COMPLElED BY PROPOSER Name and Address of Proposer
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CFD Research Corporations 400-1-Imcerne Dr.
.28/3 8*5 g m, p,. u e.*, J w A 2**'
liu n tsville, AL MB&2 $5%s' Name and Title of Principal Investigator Dr. Ashok K. Singhal; President & Technical Director Title of Project A Computer Code for Fire Protection and Risk Analysis of Nuclear Plants Technology, Resgarch Thrust and/or Potential Comercial Application (l.imit to 8 words FIRE SAFETWISK ANALYSIS; FIELD MODEL; COMPUTER SIMULATION.
Tectinical Abstract (l.imit to 200 words)
The proposed research is to advance the state-of-the-art of deterministic risk analysis and fire protection methodology of nuclear power plants.
A computer code, based on the use of a.
field (discretization) modeling technique. will be developed that has sufficient flexibility for accurate / realistic representations of geometry. ventilation and other conditions present in cable rooms and control rooms, and other enclosures in nuclear power plants.
The computer code will be capable of three-dimensional.
transient, turbulent flow and heat transfer calculations with chemical reaction and radiation.
The code will have a modular structure, specifically designed for this class of problems.
Significant emphasis will be placed on rode portability and ease-of-use.
The code wi11 have 1imited verification and documenta-tion in ph.ase I. and it wilI he enhanced in Phase i!.
Application of ResearchAnticipated Results/ Implication of the Approach (both Phase I The resultano computer code will be of significant benefit to:
11 NRC in improvino tho Requiatory Requirements such as in t
Appendiv R t o f:FR 50 and section 9.5.1 of Standard Review Plan;
- 2) Utilation in improvino dnsion and operations safety, as wnll as in comnlying with and in seeking exemptions from the NRC requ2rements: and
- 7. ) National laboratories and private contractors in providing support to NRC and utilities, i
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i U.5, NUCLEAR REGULATOR) COMM15510N SMALL BUSJNESS INNOVAT]ON RESEARCH PROGRAM (SBIR-87)
PHASE 1 - FY 1987 PROJEC1
SUMMARY
FOR NRC USE ONLY Program Office TlH Proposal No.
Topic No.
10 DE COMPLETED BY PROPOSER Name and Address of Proposer EDA Consultants 10723 Thorneliff Drive Humble, Texas 77396-i Name and Title of Principal Investigator Chelliah Sundararajan,, President Title of Project A Fur.zy Set Approach for Seismic. Safety Margins Assessment.
Research Thrust and/or Potential Conrnercial Annlication (Limit to 8 words)
Technology, Safety Margins Seismic Technical Abstract (Limit to 200 words)
The contractor shall develop a fuzzy set methodology for uncertainty analysis in probabilistic seismic safety assessment of nuclear power plants. Fuzzy set logic will be used f6r collecting, quantifying and propagating subjective, engineering judgement used in the uncertainty
' analysis. Ir. Phase I, feasibility cf the proposed approach will be
. demonstrated by applying it to selected parts of a probabilistic j
seismic safety assessment. The fuzzy set approach provides a consistent, I
rational approach incorporating engineering judgement in seismic safety assessment studies.
l A,.;.,u.7,cted ResuRs/ Implication of the Approach (both Phase I and II/ Potential Commercial j
A,.,A, J ori of Research J
Fuzzy set approach has been an effective tool for incorporating l
subjective' opinions in decision making in many branches of engineering I
and medicino. The attractiveness of fuzzy set logic lies in its ability to solicit subjective opinions in the form of " qualitative statements" end to cohvert them to formal mathematical statements by use of " fuzzy expressions".
If the proposed Phase I and Phase 11 research establishes the method to be effective for incorporating engineering judgement in j
the uncerte.inty analysis of probabilistic seismic safety as sessment i
studies it will be used by regulatory agencies and electric power utilities.
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PHASE I - FY 1987 PROJECT
SUMMARY
FOR NRC USE ONt'Y
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Program Office TTM Proposal No.
Topic No.
TO BE COMPLETED BY PROPOSER
~~
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Name and Address of Proposer Expert-EASE Systems, Inc.
1301 Shoreway Road Belmont, CA 94002 t
Name and Title of Principal Investigator Dr. Rodney R. Gay Title of Project
.. r.
Automatic Creation and Renodalization of TRAC-BWR Input Decks N,:6 Technology, Research Thrust and/or Potential Ccmercial Application (Limit to 13 words)
If Technical Abstract (Limit to 200 words)
The objective of the porposed research is to build a user-friendly, personal computer based workstation for performing TRAC-BWR analysis. The software will allow a user to read-in data from existing TRAC-BWR input decks (ASCII), draw nodalization diagram, modify input, automatically renodalize the VESSEL, CHAN and PIPE components, generate input deck, submit job, bring back plot data, display results on model diagtmn, and j
produce report quality plots.
The workstation can s!so serve as a database management system for storing and reviewing input / output from a variety of TRAC-BWR models, allowieg
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comparisons of results from ditferent TRAC-BWR executions. A methodology for convertiag existing RETRAN code input decks into TRAC-BWR models will be ittorport.te ' into the workstation.
Anticipated Results/ hnplication of the Approach (both F.tase I and II/ Potential Comercial l
Application of Research t..
The proposed work shows significant potential for man-hour savings in the performance of USNRC funded reactor safety analysis. In addition, the proposed workstation is the ideal mechanism for technology transfer' of the TRAC-BWR analysis capability to electric utility companies and
,i their consultants.
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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION SMALL BUSINESS INNOVATION RESEARCH PRO. GRAM (SBIR-87 )
PHASE 1 - FY 1987 PROJECT
SUMMARY
FOR NRC USE ONLY Program Office TTH Proposal No.
Topic No.
TO BE COMPLETED BY PROPOSER Name.and Address of Proposer Fluid Physics Ind 4265 Manchester Ave Encinitas, CA 92024 Name and Title of Principal Investigator Richard Traci, President l
Title of Project
" Severe Accident Phenomena: Natural Circulation in Reactor Coolant System" Technology, Research Thrust and/or Potential Comercial Application (Limit to 8 words)
Reactor Safety; Thermal-Hydraulics Technical Abstract (Limit to 200 words)
A research program is proposed for the development of a validated methodology for predict-ing the ex-vessel buoyancy-driven flows in the hot leg and steam generator of a Westinghouse plant, and in the candy cane hot leg of a B&W plant.
The proposed Phase I work plan consists of (a) reviewing available Westinghouse experimental data and simulating the experimental configuration, (b) validating the model using the test data and updating the model, if required (c) performing sensitivity calculations assessing the effects of flow and geometric parameters on the established flow, and (d) providing best estimates for buoyancy-driven flows within B&W hot legs.
7 Anticipated Results/ Implication of the Approach (both Phase I and II/ Potential Comercial Application cf Research The' proposed program will provide NRC with a validated methodology and associated parametric / sensitivity calculations for buoyancy-driven flows in the hot legs / steam generators of PWR plants.
The effort will assist NRC in the design and interpretation of any future experimental work, and in the assessment of the effects of buoyancy-driven flows on severe accidents in PWRs.
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SMALL BUSINESS INNOVA 110N RESEARCil PROGRAM (5B1(1-87) i.
PilASE.1 - FY 19fi7 '
PROJECT
SUMMARY
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FOR NRC USL ONLY
<I Program Office 1IM Proposal No.
,lopic No.
/NRC/SBIR-SBIR 87-76 9.2 10 BE COMPLETED BY PHOPOSER Name and Address of Proposer Kalsi' Engineering Inc.
745 Park Two Drive l
Sugar Land TX' 77478.
.)
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Name and Title of Principcl Investigator j
M. S. Kalsi', President litle of Project
- Prediction 'of Check Valve Performance & Degradation in Nucicar Power Plant Systems
.]
- Technology, Research Thrust and/or Potential Connercial Application (Limit to 8 words) j Check valve degradation, performance prediction Technical Abstract (Limit to.200 words)
F
- Thia. project is aimed ' t developing techniques for predicting check valve performance. and its a
internal components degradation in practical piping systems used in nuclear power plants.,The main focus of 3 the ef fort is to systematically evaluate the effect of upstream disturbances.
lx which are within 10 diameters upstream. By using a combined. theoretical and experimental approach, generalized nondimensional parameters will be developed which will facilitate their application to practical piping systems. Disc stability and its mocidn under unstable
. conditions will be a major ' factor of evaluation which will be used to establish wear and fatigue degradation trends. The successful completion of this projnct will help eliminate
]
- - most of the f ailures 'which have become a major cause for concern. Maintenance and inspection j{
frequency, to avoid unacceptable degradation,- will also be quantitatively. predicted.as a result of this project.
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- Anticipated Results/Irnplication of the Approach -(both Phase I and II/ Potential Commercial i
Application of Research
' - Successful completion of Phases I and II will result in an objective, technical prediction
' me, id.. which. has the potential of drama tically reducing the check valve failure. The prediction methods - will also be available to assist in identifying the problem valves in
' current application and the predicted remaining useful life based on internal component wear.
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i APPENDIX B
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U.5.-NUCLEAR' REGULATORY COMMISSION-SMALL BUSINESS INNOVATION'RESEARCH PROGRAM (SBIR-87 )
PHASE' I - FY 1987
]
.c PROJECT
SUMMARY
.m FOR NRC USE ONLY Program 0ffice TTM Proposal No.
Topic No.
~
TO BE COMPLETED BY PROPOSER Name and Address of Proposer J;Pcitangan & Associates / Sentry Equipment Corp.
'856 East Armour Road /P.O. Box ~ 127 Oconomowoc, N! 53066 Name and Title of Principal Investigator James P. Langan, Engineering Consultant Title of Project L Analysis of State-of-the-Art Methods for. Containment Integrated Leak Rate Testing:
Technology ~Research. Thrust anc'/or Potential Comercial Application (Limit to -8 words)
ILRT Accura te Automated Testing Techniques Technical Abstract (Limit ;to 200 words).
)
The J.P. Langan.& Associates, Inc/ Sentry Equipment 'Lorp. team will develop two modeling
'j methods to enhance ILRT testing confidence.
We intend to establish-'a mathematical relationship between containment size and the frequency duration between discrete: data points for ideal automated datascollection and evaluation.
We will support the proposed Appendix J by analyzing 20 Type A and verification tests under p.the new ILRT test criteria.
We will evaluate date enhancement techniques used extensively to eliminate random data error from transducer measurements to increase test confidence.
Anticipated Results/ Implication of the Approach (both Phase I and II/ Potential Commercial
{
i Application'of Research
'It' is-anticipated 'that a relationship between containment size' and the ideal frequency
- duration between discrete collected data can be established.
It is also believed that data enhancement techniques will produce more accurate test results by the elimination
.,of' random error, Page 2 of 20
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U.(. NUCLEAR REGULATORY COMISSION SMALLBUSINESSINNOVATIONRESEARCHPR0di..A.
(SBIR 87) j PHASE l - FY.1967 j
PROJECT
SUMMARY
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~
FOR NRC USE ONLY Program Office TTM Proposal No.
Topic No.
?
10 BE COMPLETED BY PROPOSER i Nane and Address of Proposer M t V:A Yot-A I MmRhl ATie^l Al t?2, via ii St.
Sade iIe5 Nkr York, hl. '/.
Iceos Name and Title of Princi al Investigator 1
$4eebert Ba <ds%
co asot r M EM&t aEW
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TTtTe of Project
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Technical Abstract (Limit to 200 words)
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(5,e Py e.3) k Anticipated Results/ Implication of the Approach (both Phase I and-II/ Potential ' Commercial Application of Research a*fsc,4 p6DG(
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Trt:t 226606 WALL UR (T.CA) i MIKAYOLA INTERNATIONAL l
Advanced Technology e, Advisory Services l
Trade and Industry 82 Wall Street, Suite 1105 l
l Tehnical Abstract 1
An " Inhalation Screening Dose Collar-Stanchion Manitor" with embedded miniature thin-wall Geiger-Mueller radiation detectors
)
evaluates the thyroid gland for nanocurie activity inventory.
l The primary technical objective is to establish that the " Collar-'
l Stanchion Monitor" will be capable of detecting within 15 seconds and with 2 sigma confidence an activity inventory of 0.7 nanocuries of l
I-131 dose equivalent in an nuclear power plant administrative area.
l The " Collar-Stanchion Monitor" is made of plexiglas arri lined with lead.
]
The " Collar-Stanchion Monitor" is invested with a neck phantom which allows the nesting of calibration sources typically representative of thyroid I-131 &
I-133 uptake in nanocuries by radiation wrkers unprotected by breathing apparatus when exposed for cne hour in various occupancy areas throughout typical U.S. BWR and PWR plants.
A prototype Collar-Stanchion Manitor will be produced. Upon completion and calibration of the prototype, a currelational analysis
.will be engaged at a National Laboratory to compare the sensitivity, process time, counting time, MD%
confidence levels and l
reliability of the " Collar-Stanchion Monitor" with chair-type thyroid l
counters, exit whole bcx1y counting chambers, and portal nonitors viewing the calibration sources.
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APPENDIX B
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U.S. NUCLEAR REGULATORY COMMISSION SMALL BUSINESS INNOVATION RESEARCH PROGRAM (SBIR-87 )
PHASE I - FY 1987 I
PROJECT
SUMMARY
l FOR NRC USE ONLY l
Program Office TIM Proposal No.
Topic No.
J
~~~
TO BE COMPLETED BY PROPOSER
~
Name and Address of Proposer PDI Technology, Inc.
246 Viking Ave.
Brea, Ca.
92621 Name and Title of Principal Investigator D. G. Swanson, Vice-President i
Title of Project Aging Processes In Nuclear Reactors Technology, Research Thrust and/or Potential Coninercial Application (Limit to 8 words)
Nuclear Reactor Aging Processes, Reactor Safety Technical A)stract (Limit to 200 words)
As nuciear plants grow older, environmental conditions may affect the, performance of plant concrete and steel structures.
We propose to develop a strctogy which can be applied to concrete structures.
It will be directed towards the development of methods to analyze and understand aging effects, including the combined effects of time, temperature, radiation, humidity, stress, chemical environment and electrical and magnetic fields.
It will include consideration of current methods and technology for surveillance and maintenance of concrete structures.
In addition, methods for post-service examination and testing will be proposed.
Finally, aging assessment methods will be, developed for concrete structures.
The strategy will be applied to a single concrete component of a nuclear plant in Phase I.
In the second phase, each of the concrete aging processes will be examined in greater detail.
Plant components in different types of containments will be cxamined to determine where failures are most likely to occur and where they are most nianificant in hme n f' ~*r Anticipated Results/ Implication of the Approach (mfatmboth Fhase I and II/ Potential Connercial Application of Research Development of methods to analyze aging effects, post-service examination m3thods and aging assessment methods will solve an operational problem for the nuclear industry, reduce the cost of plant construction, and increase the life of expectancy of existing reactor systems.
The development of thece methods will enhance the safety of nuclear power plants, reduce the possibility of radiation exposure to the public and provide information on and solutions to a major potential problem in the nuclear industry.
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i APPEN0ZX 8 U.S.4 NUCLEAR REGULATORY' COMMISSION
~
SMALL BUSINESS INNOVATION RESEARCH PROGRAM (SBIR-87)
PHASE II-FY 1987 PROJECT
SUMMARY
FOR NRC USE ONLY I
Program Office TIM
. Proposal.No.
Topic No.
TO BE COMPLETED BY PROPOSER '
J Name and Address of Proposer.
Analysis and Measurement Services Corporation 9111 Cross. Park Drive,'
oxville, Tenn. 37923-4599 lName and Title of Principal Investigator l
H. M. Hashemian, President.
Title of Project Degradation of Nuclear ~ Plant Temperature Sensors
.R Technology, Research Thrust and/or Potential-Connercial Application (Limit to 8 words)
' Temperature Sensors Calibration,: Response Time, Aging, Periodic Testing Technical Abstract (Limit to 200 words) r A' systematic' Phase II effort is' proposed to evaluate the long term perfernance of 2
temperature sensors of the type used in nuclear plant safety systems. This vill involve-experimental drift: testing, development of verified calibration procedures, data reduction methods and tolerances, development 'of sensor design guidelines, development of in-situ -
testing methods, development of in-situ " sensor healing". methods, and evaluation of nev
-sensor-technology. This research is needed to enable the utilities to make bett.er measure--
ments,(uith reduced cost) and to enable the NRC to perform its rule-making function from a basis of fact, i
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Anticipated Results/ Implication of the Approach (both Phase I and II/ Potential Connercial Application of Research The results of'this study will enable the utilities co satisfy their requirement of insuring accurate, timely safety system sensor. performance and will provide the NRC with a data base and testing methodology to ensure compliance.
This vill result in safer = plants and lower costs (due to elimination of current confusion and vasted efforts).
Page 2
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APPENDIX B
' U.SI NUCLEAR REGULATORY COMMISSION g.
SMALL BUSINESS INNOVATION RESEARCH PROGRAM (SBIR-07 )
p PHASE I - FY 1987 5
PROJECT
SUMMARY
1 FOR NRC USE ONLY Progrim Office TTM Proposal No.
Topic No.
~
TO BE COMPLETED BY PROPOSER Name ind Address of Proposer Future Resources-Associates, Inc.
2000 Center Street, Suite 418 j
Berkeley, California 94704 Name and Title of Principal Investigator Robert J. Budnitz, President j
"[x"thnYfngaHCLPF-BasedSeismicMarginReviewtoAnalyzePlantDamage States and the Role of Human Factors and Non-Seismic Failures" Technclogy, Research Thrust and/or Potential Corrrnercial Application (Limit to S words)
Improved methodology for seismic margin reviews of nuclear power' plants Technical Abstract (Limit to 200 words)
Techniques have recently been developed to analyze the ' seismic margin' of a nuclear-power plant using the HCLPF concept (high confidence of a low probability of failure).
)
6 This HCLPF-based approach can screen a plant's seismic ' margin' without undertaking a full-scope PRA.
However, insights from the method are limited, partly because those j
accident sequences that dominate'the HCLPF-based analysis are not themselves analyzed i
further.
In. Phase 1 of this project, the feasibility of carrying the HCLPF-based analysis l
one step.further was explored: additional insights in the areas of non-seismic-induced failures, human interactions, and the potential for severe offsite consequences were found to be available using an expanded methodological approach.
In the proposed j
Phase II, the exparded methodology will be applied in three test cases: by reanalyzing the recently completed margins reviews of Maine Yankee and Catawba, and by a trial appli-cation for the upcoming SWR margin review that NRC and EPRI are planning together.
If the expanded methodology is shown to be effective and inexpensive, it should be feasible to i
use it for routine application in seismic margin reviews.
Anticipated Results/ Implication of the Approach (both Phase I and II/ Potential Commercial j
Applic1 tion of Research if the proposed research is successful, the usefulness of NRC's new HCLPF-based methodo-logy for seismic margin reviews of nuclear power plants will be significantly enhanced, and at only slightly greater cost Decision-makers in both the NRC and the utility will j
have great use for the additional information produced (about the potential for severe offsite consequences, non-seismic-induced f ailures, and the rols of human interactions in the key seismic accident sequences.)
Several commercial firms that have the capability to perform seismic margin reviews would benefit from this improved methodology.
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- APPENDIX BL
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-U.S.' NUCLEAR RFGULATORY COMMISSION SMALL BUSINESS' INNOVATION RESEARCH' PROGRAM (SBIR-86)
PHASE 1 - FY 1967 l
PROJECT
SUMMARY
1
)
i FOR NRC USE ONLY Program Office TTM Proposal No.
Topic No.
TO BE COMPLETED BY PROPOSER
- )
^
Name and Address of Proposer
-Quantex Corporation 2 Research Court Rockville, Maryland 20850 i
. Nam and Title of Principal Investigator,,
Charles Y. Wrigley, Vice President, Technology 4
Title of Project Integrating Fiber 0ptir"Radiacion Dosimeter - Phase II M'
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'NerTp' tic"*[o'simeter"s", (agf[ggee8cp,ntygl f'8sTEltyp Application (Limit to Il words)-
s Pt c
~
Technical Abstract (Limit to 200 words)-
._ Having demonstrated the feasibility of employing electron trapping materials' with optical fibers for dosimetry in Phase I,'it is proposed that materisis and equipment.be further developed to monitor the various radiation L
species likelyJto be. present-in a reactor environment.
Specifically, the Quantex ET materials will be optimized for response to specific kinds of
. radiation, bonding and encapsulation techniques and materials will be worked out, optical ; coupling -optimized, and calibration constants for the various i
radiations quantified.
Also,.long term evaluation of the integrated dose i
retention and probe aging will be performed.
Prototy'pe instrumentation will be i
. designed, constructed and exercised.
)
l Anticipated-Results/linplication of the Approach (both Phase I and II/ Potential Commercial J
' Application of Research The. development of photonic materials for fiber optic dosimeters can
! j lead to new radiation measurement instrumentation.
Fiber microprobe dosimeters b
could have both commercial and governmental applications to determine radiation levels in power generating reactor environs, medical, isotope detection, micro-1 volume' remote sensing, etc.
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