California Nuclear Engineer Bus. & Prof. Code § 6701 Attorney Fee Petition Mechanics: PE-NNNNN License Verification at search.dca.ca.gov as Primary Welch Anchor (THE ONLY § 6701 Nuclear Engineer Page — DISTINCT FROM Mechanical Engineer PE-ME, Electrical Engineer PE-EE, and Chemical Engineer PE-ChE; ANS 8.1 Nuclear Criticality Safety Advisory [PURE KETCHUM]; IEEE 323 Class 1E Equipment Qualification Advisory [PURE KETCHUM]; TWO PRIVATE STANDARDS ON ONE PAGE), CCP § 1021.5 Private Attorney General Mandatory Attorney Fees

Statutory Framework: Bus. & Prof. Code § 6701 and PE-NE Licensure — Nuclear Engineer Requirements, Scope of Practice, ANS 8.1 [PURE KETCHUM], IEEE 323 [PURE KETCHUM], and CCP § 1021.5 Mandatory Attorney Fees

California Business and Professions Code § 6701 defines nuclear engineering as a branch of professional engineering encompassing nuclear reactor design, core physics calculations, nuclear criticality safety analysis, radiation shielding design, spent nuclear fuel storage engineering, radioactive waste processing and management system design, nuclear instrumentation and control engineering, radiation protection program design, and nuclear facility safety analysis report (SAR) preparation. § 6701 requires that nuclear engineering documents — including criticality safety analysis reports, shielding design calculations, fuel storage criticality evaluations, NRC license application technical specifications, and nuclear system safety function analyses — be prepared and stamped by a PE-NE licensed by BPELSG. § 6785 prohibits unlicensed engineering practice; a person who prepares and stamps nuclear engineering documents without a valid PE-NE commits § 6785 unauthorized engineering practice, actionable under CCP § 1021.5 when the litigation vindicates a public interest in safe nuclear facility operations protecting California workers, nuclear facility operators, and the public. Primary Welch anchor: BPELSG License Verification at search.dca.ca.gov (PE-NNNNN module, PE-NE designation).

ANS 8.1 NUCLEAR CRITICALITY SAFETY IN OPERATIONS WITH FISSIONABLE MATERIALS OUTSIDE REACTORS ADVISORY — THE ONLY PAGE IN THE FEE-PETITION-MECHANICS SERIES WITH AN ANS 8.1 NUCLEAR CRITICALITY SAFETY ADVISORY — PURE KETCHUM: The American Nuclear Society (ANS, founded 1954, La Grange Park IL) is a private nonprofit professional organization for nuclear engineers and scientists. ANS-8.1-2014 (Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors) establishes the nuclear criticality safety standards for all operations involving fissile materials outside nuclear reactors, including fuel fabrication, fuel processing, spent fuel reprocessing, and research laboratory operations with uranium or plutonium. Key ANS 8.1 requirements applicable to nuclear engineering design documents include: (1) subcritical limit methodology — the upper subcritical limit (USL) must provide at least 95% probability that the system is subcritical in at least 95% of all similar operations (double 95% criterion); computer codes used for criticality calculations must be validated against OECD/NEA benchmark critical experiments within the applicable parameter range; (2) double-contingency principle — process conditions must be designed so that at least two independent and unlikely concurrent changes in process conditions are required before criticality is possible; (3) single-parameter subcritical limits — for fissile solution operations: uranium-235 maximum concentration 12 g/L in infinite homogeneous aqueous system without reflection; 233U mass limit 530 g under full water reflection; (4) safety margin — a minimum margin below the USL to account for uncertainties in calculational bias and bias uncertainty. A nuclear engineer who calculates H/X (hydrogen-to-fissile-atom ratio) criticality safety limits without applying the ANS 8.1 USL double-95% criterion — using nominal values rather than upper subcritical limits with validated bias correction — directly creates an undetected criticality risk. ANS 8.1 standards create no statutory fee-shifting; they are PURE KETCHUM advisories.

IEEE 323-2003 QUALIFYING ELECTRICAL EQUIPMENT FOR NUCLEAR POWER GENERATING STATIONS ADVISORY — THE ONLY PAGE IN THE FEE-PETITION-MECHANICS SERIES WITH AN IEEE 323 CLASS 1E EQUIPMENT QUALIFICATION ADVISORY — PURE KETCHUM: The Institute of Electrical and Electronics Engineers (IEEE, founded 1963, Piscataway NJ) is a private nonprofit professional organization. IEEE 323-2003 (Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations) establishes the requirements for qualifying Class 1E (safety-related) electrical equipment used in nuclear power generating stations. Key IEEE 323 requirements applicable to nuclear engineering design documents include: (1) equipment qualification (EQ) program — documentation of the qualified life for each Class 1E component including normal aging, simulated accident environment (LOCA/HELB: temperature, pressure, humidity, radiation dose, chemical spray), seismic qualification by test or analysis (IEEE 344), and post-accident operability demonstration; (2) aging management — thermal aging simulation per Arrhenius model for polymeric materials; radiation aging dose per 10 CFR 50.49 integrated design basis dose calculation; (3) maintenance and surveillance — qualified life replacement schedules and periodic testing requirements documented in the EQ file for each component; (4) documentation — an EQ file for each Class 1E component listing the qualified life, bounding analysis conditions, test and analysis reports, and replacement schedule. A nuclear engineer who specifies Class 1E electrical instrumentation for nuclear power plant safety systems without IEEE 323 equipment qualification documentation commits a documented nuclear engineering practice deficiency — the facility cannot demonstrate that safety-related instruments will perform their required function during and following a design basis accident. IEEE 323 standards create no statutory fee-shifting; they are PURE KETCHUM advisories.

Three-Track Harm: Criticality Excursion H/U Ratio Error Workers Hospitalized ($285K), Radiation Shielding Error Dose Rate 2.5× Limit NRC NOV ($195K), and Class 1E IEEE 323 Qualification Documentation Absent NRC Penalty ($95K)

• UNLICENSED NUCLEAR ENGINEER STAMPS CRITICALITY SAFETY ANALYSIS FOR URANIUM HEXAFLUORIDE (UF⊂6;) CONVERSION FACILITY WITHOUT PE-NE LICENSE (BPELSG CONFIRMS NO PE-NNNNN PE-NE IN DATABASE) — ANS 8.1 UPPER SUBCRITICAL LIMIT (USL) METHODOLOGY NOT APPLIED — CALCULATIONAL BIAS CORRECTION FOR VALIDATED CODE OMITTED — H/U RATIO SUBCRITICAL LIMIT CALCULATED AT NOMINAL RATHER THAN USL CONDITION — ACTUAL USL AT EXISTING PROCESS CONDITIONS 15% BELOW NOMINAL CALCULATION — CRITICALITY EXCURSION DURING NORMAL OPERATIONS — TWO WORKERS RECEIVE PROMPT RADIATION DOSE EXCEEDING 10 CFR 835 100 REM ANNUAL OCCUPATIONAL DOSE LIMIT — HOSPITALIZED WITH ACUTE RADIATION SYNDROME SYMPTOMS — $285,000 WORKERS’ COMPENSATION AND REGULATORY RESPONSE COSTS. A California research facility operated a uranium hexafluoride (UF₆) conversion process requiring a criticality safety analysis (CSA) as part of its NRC facility license. The CSA was prepared and stamped by a licensed Chemical Engineer (PE-ChE) with no BPELSG PE-NE designation (BPELSG confirms no PE-NNNNN PE-NE for the defendant). ANS 8.1 Section 4.3 requires that criticality safety calculations using computer codes be validated against appropriate OECD/NEA benchmark experiments, with the validated bias and bias uncertainty incorporated into the USL using the double-95% subcritical limit methodology. The PE-ChE used the SCALE 6.1 criticality code with published material cross-sections but did not perform the required code validation against benchmark critical experiments in the applicable parameter range (aqueous U(IV) solutions, H/U ratio 500–1,000, enrichment 3.5% 235U). The USL applicable to the process — accounting for calculational bias and bias uncertainty — was 15% lower than the nominal calculated k-eff. The CSA declared the process subcritical at the nominal concentration limit without applying the ANS 8.1 USL margin. During a normal operations campaign, a valve position indicator malfunction caused an inadvertent solution concentration to accumulate 12% above the nominal limit; the combined concentration error and USL non-application resulted in a criticality excursion (k-eff exceeding 1.0 transiently) lasting approximately 90 seconds. Two workers in the process cell received prompt neutron and gamma doses estimated at 120–180 rem — above the NRC 10 CFR 835 annual occupational dose limit of 5 rem and the emergency dose limit of 25 rem — and were hospitalized for observation and treatment for acute radiation syndrome symptoms. The attorney retained a licensed PE-NE and ANS criticality safety specialist to opine on the ANS 8.1 USL methodology omission and causation of the criticality excursion.
• UNLICENSED ENGINEER STAMPS RADIATION SHIELDING CALCULATION FOR INDUSTRIAL RADIOGRAPHY VAULT WITHOUT PE-NE — SHIELD WALL THICKNESS CALCULATED USING EXPOSURE RATE CONSTANT FROM OUTDATED REFERENCE (1970s DATA) INSTEAD OF CURRENT NCRP 151 TABLE FOR COBALT-60 — CALCULATED DOSE RATE AT OCCUPIED AREAS 2.5× HIGHER THAN DESIGN VALUE AFTER COMMISSIONING — WORKERS IN ADJACENT OFFICE RECEIVING ESTIMATED 500 mrem/WEEK VS NRC 10 CFR 20.1201 OCCUPATIONAL LIMIT 100 mrem/WEEK (2 mSv/week) — NRC NOTICE OF VIOLATION (NOV) UNDER 10 CFR 20 — RADIOGRAPHY VAULT CLOSED FOR SHIELDING UPGRADE — $195,000 UPGRADE AND REGULATORY RESPONSE COSTS. A California non-destructive testing company constructed a new industrial radiography vault for gamma radiography work using a 40 Ci Cobalt-60 source. A shielding calculation was required for NRC license amendment to authorize use of the new vault. The shielding calculation was prepared and stamped by a licensed Mechanical Engineer (PE-ME) with no BPELSG PE-NE designation. The PE-ME calculated the required shield wall thickness using an exposure rate constant (Γ) for cobalt-60 from a 1975 health physics handbook. The correct cobalt-60 exposure rate constant from NCRP Report 151 (2005 edition) is Γ = 13.0 R•cm²/mCi•hr, reflecting updated cross-section values from the ENDF/B-VII nuclear data library. The PE-ME’s calculation used Γ = 10.6 (outdated value), producing a calculated shield thickness of 9.0 cm of lead for the control room barrier — 18% less than the 10.9 cm of lead that the correct Γ value would have required. After vault commissioning, NRC inspection measurements at the adjacent occupied office area showed a dose rate of 2.5 mrem/hr — annualizing to approximately 500 mrem/week for occupancy factor of 1.0. NRC issued a Notice of Violation under 10 CFR 20.1201 (occupational dose limits); the company was required to increase shield thickness and demonstrate compliance. Facility closure, shielding upgrade, and regulatory response costs totaled $195,000. The attorney retained a licensed PE-NE and health physicist to opine on the shielding calculation deficiency.
• UNLICENSED ENGINEER SPECIFIES SEISMIC CATEGORY I INSTRUMENTATION FOR NUCLEAR POWER PLANT AUXILIARY FEEDWATER SYSTEM WITHOUT PE-NE — IEEE 323 CLASS 1E EQUIPMENT QUALIFICATION (EQ) FILES NOT ESTABLISHED FOR 14 INSTRUMENTS — NRC INSPECTION FINDS EQ DOCUMENTATION ABSENT — NON-CITED VIOLATION (NCV) ISSUED UNDER 10 CFR 50 APPENDIX B CRITERION III — SIGNIFICANCE DETERMINATION PROCESS GREEN FINDING — REGULATORY PENALTY AND CORRECTIVE ACTION PROGRAM COSTS — $95,000. A California nuclear power plant was replacing 14 transmitters and pressure switches in its Auxiliary Feedwater (AFW) system — a seismic Category I, Class 1E safety system required to supply feedwater during a loss-of-coolant accident (LOCA) or station blackout event. The engineering change (EC) package specifying the replacement instruments was prepared and stamped by a licensed Electrical Engineer (PE-EE) with no BPELSG PE-NE designation. The PE-EE specified commercially available process instruments that met the applicable seismic qualification requirements of IEEE 344 (confirmed by vendor seismic test data). However, the PE-EE did not establish IEEE 323 equipment qualification (EQ) files for the 14 instruments documenting qualified life, thermal aging analysis (Arrhenius model for polymeric insulation), radiation aging dose (per 10 CFR 50.49 integrated design basis dose for AFW harsh environment zone), LOCA/HELB qualification testing, and replacement schedule in the plant’s EQ program. NRC resident inspectors identified the missing EQ file documentation during a routine inspection six months after installation. The finding was classified as a non-cited violation (NCV) under 10 CFR 50 Appendix B Criterion III (design control), with a Significance Determination Process (SDP) result of Green (low safety significance). The corrective action program (CAP) required: retroactive EQ file preparation by a licensed PE-NE, thermal and radiation aging analyses for all 14 instruments, and a probabilistic risk assessment review to bound safety significance. Regulatory response and corrective action costs totaled $95,000. The plant retained a licensed PE-NE to opine on the IEEE 323 EQ documentation requirement omission.

Distinct From Mechanical Engineer (§ 6701 PE-ME), Electrical Engineer (§ 6701 PE-EE), Chemical Engineer (§ 6701 PE-ChE), and NRC-Licensed Reactor Operator

• DISTINCT FROM Mechanical Engineer (Bus. & Prof. Code § 6701 PE-ME designation) — A Mechanical Engineer (PE-ME) is licensed to perform HVAC system design, process piping, pressure vessel engineering, and mechanical equipment engineering; a Mechanical Engineer may not stamp nuclear criticality safety analyses, radiation shielding calculations, or nuclear facility safety analysis reports without PE-NE designation; a Nuclear Engineer (PE-NE) may not stamp HVAC system designs, process piping calculations outside nuclear code compliance, or conventional pressure vessel specifications without PE-ME designation: nuclear facilities require both PE-NE-stamped nuclear safety documents and PE-ME-stamped mechanical system designs; neither designation authorizes practice of the other.
• DISTINCT FROM Electrical Engineer (Bus. & Prof. Code § 6701 PE-EE designation) — An Electrical Engineer (PE-EE) is licensed to perform power distribution system design, arc flash hazard analysis, and electrical equipment specifications; an Electrical Engineer may not stamp IEEE 323 Class 1E equipment qualification documentation, nuclear instrumentation design, or Class 1E cable raceway routing analyses requiring nuclear engineering judgment without PE-NE designation; a Nuclear Engineer (PE-NE) may not stamp electrical power distribution designs, arc flash hazard studies, or protective device coordination studies without PE-EE designation: nuclear power plant Class 1E electrical systems typically require both PE-EE-stamped power distribution documents and PE-NE-stamped safety function qualification documents; the PE-EE license does not authorize preparation of IEEE 323 EQ files.
• DISTINCT FROM Chemical Engineer (Bus. & Prof. Code § 6701 PE-ChE designation) — A Chemical Engineer (PE-ChE) is licensed to perform chemical process design, reaction engineering, and process safety analysis; a Chemical Engineer may not stamp nuclear criticality safety analyses (ANS 8.1), spent nuclear fuel storage criticality evaluations, or radiation protection engineering design documents without PE-NE designation; a Nuclear Engineer (PE-NE) may not stamp chemical process flow diagrams, P&IDs requiring PE-ChE stamp authority, or industrial chemical reaction engineering without PE-ChE designation: uranium solution processing facilities involve both PE-ChE process chemistry and PE-NE criticality safety; the ANS 8.1 criticality safety analysis requires PE-NE stamp authority regardless of the chemical engineering expertise of the preparer.
• DISTINCT FROM NRC-Licensed Reactor Operator (10 CFR Part 55) — An NRC reactor operator license authorizes operation of a licensed nuclear facility at the control room level; it does NOT authorize preparation and stamping of nuclear engineering design documents, criticality safety analyses, shielding calculations, safety analysis reports, or Class 1E equipment qualification documentation: NRC operations licensing and BPELSG PE-NE serve entirely different functions; an NRC-licensed operator who prepares and stamps nuclear engineering design documents without PE-NE commits § 6785 unauthorized engineering practice regardless of NRC operator licensure.

Billing Gaps 1–3 Summary (13.25 hrs/yr = $3,975–$6,625/yr)

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