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Award Data

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The Award database is continually updated throughout the year. As a result, data for FY20 is not expected to be complete until September, 2021.

Download all SBIR.gov award data either with award abstracts (290MB) or without award abstracts (65MB). A data dictionary and additional information is located on the Data Resource Page. Files are refreshed monthly.

  1. ON-LINE PLANT ANALYZER UTILIZATION IN CONTROL ROOM FOR SURVEILLANCE OF PLANT SAFETY

    SBC: Dynatrek Inc.            Topic: N/A

    N/A

    SBIR Phase I 1984 Nuclear Regulatory Commission
  2. ENERGY INCORPORATED (EI) PROPOSES TO ESTABLISH A PROGRAM TO GATHER PLANT-SPECIFIC DATA FOR THE NUCLEAR PLANT DATA BANK (NPDB).

    SBC: Ei International Inc.            Topic: N/A

    ENERGY INCORPORATED (EI) PROPOSES TO ESTABLISH A PROGRAM TO GATHER PLANT-SPECIFIC DATA FOR THE NUCLEAR PLANT DATA BANK (NPDB). THE REQUIRED INFORMATION INCLUDES GEOMETRIC AND OPERATING DATA FOR THE PRIMARY SYSTEM SECONDARY SYSTEM, BALANCE OF PLANT, AND THE TRIP AND CONTROL SYSTEMS. ALTHOUGH, EVERY COMMERCIAL REACTOR IN THE U.S. HAS UNIQUE DESIGN FEATURES, MANY SHARE A COMMON BASIS BY WHICH THEY CA ...

    SBIR Phase I 1984 Nuclear Regulatory Commission
  3. SEISMIC PRA

    SBC: Future Resources Associate Inc            Topic: N/A

    TECHNIQUES HAVE RECENTLY BEEN DEVELOPED FOR PROBABILISTIC ANALYSIS OF THE RISK FROM EARTHQUAKE-INITIATED ACCIDENTS AT NUCLEAR POWER REACTORS. THESE TECHNIQUES ARE NOW BEGINNING TO BE APPLIED AS PARTS OF LARGER PRA STUDIES. HOWEVER, THE RESULTS OF APPLYING THESE SEISMIC PRA TECHNIQUES ARE STILL HIGHLY UNCERTAIN, AND THE INSIGHTS GAINED SO FAR FOR SPECIFIC REACTORS ARE DIFFICULT TO GENERALIZE. THIS ...

    SBIR Phase II 1984 Nuclear Regulatory Commission
  4. SEISMIC PRA

    SBC: Future Resources Associate Inc            Topic: N/A

    N/A

    SBIR Phase I 1983 Nuclear Regulatory Commission
  5. OF REACTOR THERMAL-HYDRAULICS CODES.

    SBC: International Energy            Topic: N/A

    OF REACTOR THERMAL-HYDRAULICS CODES. THESE CODES HAVE INTERNATIONAL APPLICABILITY. TRADING THESE CODES, WITH APPROPRIATE SUPPORT, TO OTHER COUNTRIES IN EXCHANGE FOR USEFUL DATA OR OTHER RESOURCES WOULD ENHANCE THEIR UTILITY AND REDUCE THE COST TO THE U.S. OF GENERATING EQUIVALENT DATA. THE OBJECTIVE OF THE PROPOSED EFFORT IS TO PROVIDE A SOUND TECHNICAL BASIS FOR SUCH AN EXCHANGE BY: IDENTIFYING C ...

    SBIR Phase I 1983 Nuclear Regulatory Commission
  6. COMMON CAUSE FAILURE ANALYSIS (CCFA) IS AN IMPORTANT ELEMENTOF A COMPLETE PROBABILISTIC RISK ASSESSMENT (PRA) SINCE COMMON CAUSE FAILURE IS OFTEN A DOMINANT FACTOR IN THE RISK POSED TO PUBLIC HEALTH AND SAFETY FROM NUCLEAR POWER PLANT ACCIDENTS.

    SBC: Jfb Associates Inc.            Topic: N/A

    COMMON CAUSE FAILURE ANALYSIS (CCFA) IS AN IMPORTANT ELEMENTOF A COMPLETE PROBABILISTIC RISK ASSESSMENT (PRA) SINCE COMMON CAUSE FAILURE IS OFTEN A DOMINANT FACTOR IN THE RISK POSED TO PUBLIC HEALTH AND SAFETY FROM NUCLEAR POWER PLANT ACCIDENTS. A MAJOR PROBLEM ASSOCIATED WITH PERFORMING A CFFA IS THE ABSENCE OF SUPPORTING QUALITATIVE OR QUANTITATIVE COMPONENT FAILURE DATA. EXTENSIVE AMOUNTS OF CO ...

    SBIR Phase I 1983 Nuclear Regulatory Commission
  7. PRESSURIZED WATER REACTOR TRANSIENT ANALYSIS MODEL

    SBC: Levy S Inc            Topic: N/A

    THIS PROPOSAL DESCRIBES A PROGRAM FOR THE DEVELOPMENT OF A PRESSURIZED WATER REACTOR (PWR) TRANSIENT ANALYSIS MODEL FOR USE ON A MICROCOMPUTER. THE INTENDED APPLICATION OF THIS PWR MODEL IS FOR ANALYSIS OF PWR SYSTEM RESPONSE, INCLUDING FORCED AND NATURAL CIRCULATION OPERATION. THE PRIMARY USER OF SUCH A MODEL WOULD BE A UTILITY OR NUCLEAR REGULATORY COMMISSION (NRC) ENGINEER OR PLANT OPERATOR PER ...

    SBIR Phase II 1984 Nuclear Regulatory Commission
  8. PRESSURIZED WATER REACTOR TRANSIENT ANALYSIS MODEL

    SBC: Levy S Inc            Topic: N/A

    N/A

    SBIR Phase I 1983 Nuclear Regulatory Commission
  9. PROBABILITY OF FLOODS WITH VERY LONG RETURN PERIODS

    SBC: Linsley Kraeger Associates Ltd            Topic: N/A

    N/A

    SBIR Phase I 1984 Nuclear Regulatory Commission
  10. FOLLOWING A POST-IRRADIATION THERMAL ANNEAL CYCLE

    SBC: Materials Engineering &            Topic: N/A

    THIS RESEARCH WILL DETERMINE THE CHARACTERISTICS OF FATIGUE INITIATION LIFE, FATIGUE CRACK GROWTH RATES, AND SLOW-STABLECRACK EXTENSIONS FOLLOWING A POST-IRRADIATION ANNEAL. FATIGUE INTITIATION LIFE WILL BE DETERMINED WITH SMOOTH SPECIMENS OF 308 STAINLESS STEEL WHICH HAVE BEEN ANNEALED AND THEN CYCLED IN PRESSURIZED, HIGH-TEMPERATURE WATER. SUBMERGED ARC WELD STEEL (LINDE 80 FLUX) WILL BE PARTIAL ...

    SBIR Phase I 1983 Nuclear Regulatory Commission
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