NUREG-1747

Overview and Sutm mary of 

NRC Involvemenft with DOE in 

the Tank Waste Remediation 

SystemoPrivatization (TWRS-P) 

Program 

U.S. Nuclear Regulatory Commission RE 

Office of Nuclear Material Safety and Safeguards 

Washington, DC 20555-0001
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NUREG-1747

Overview and Summary of 

NRC Involvement with DOE in 

the Tank Waste Remediation 

System-Privatization (TWRS-P) 

Program 

Manuscript Completed: June 2001 

Date Published: August 2001 

Prepared by 

A. P. Murray, M. N. Baker, F. H. Burrows, 

S. A. Burrows, L. W. Chang, G. C. Comfort, 

R. E. Shewmaker, W. R. Smith, M. Srinivasan, 

R. K. Struckmeyer, M. Tokar, R. G. Wescott, 

A. Wong 

Division of Fuel Cycle Safety and Safeguards 

Office of Nuclear Material Safety and Safeguards 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001
ABSTRACT 

In 1995, the U.S. Department of Energy (DOE) embarked on an effort to privatize the 

processing through vitrification of 54 million gallons of radioactive waste that has been stored in 

177 underground storage tanks at the Hanford Site. The U.S. Nuclear Regulatory Commission 

(NRC) provided assistance to DOE on the Tank Waste Remediation System-Privatization 

(TWRS-P) program, with a potential transition to NRC regulatory authority at a future time. In 

2000, DOE terminated the privatization approach, and decided to use more traditional 

contracting methods. 

During their reviews, NRC staff analyzed both unmitigated and mitigated consequences from 

potential accident scenarios at the proposed facility. NRC staff's efforts identified several key 

areas of uncertainty, such as melter failure modes and frequencies, that would require further 

study before more refined analyses could be performed. The reviews also identified several 

open issues, including the need for significantly more detailed design information and safety 

analyses, and greater defense-in-depth. In particular, the design, at the time of termination of 

the privatization, was found to be very preliminary and corresponded to perhaps a 15 percent 

level of design. 

This report summarizes NRC's participation in and observations on the TWRS-P program and 

identifies issues from the NRC's perspective.

°,. NUREG-1747
TABLE OF CONTENTS 

A BST RA C T .............................................................. iii 

EXECUTIVE SUMMARY ..................................................... ix 

ABBREVIATIONS AND ACRONYMS .......................................... xv 

GLOSSARY OF TERMS .................................................... xvii 

1.0 INTRODUCTION AND BACKGROUND .................................... 1 

1.1 Hanford and the Tank W astes ...................................... 1 

1.2 The Tri-Party Agreement ......................................... 3 

1.3 DOE Program and Contracts ....................................... 6 

1.4 DOE Regulatory Approach ......................................... 9 

1.5 NRC Involvement via the Memorandum of Understanding ................ 13 

1.6 Overview of this Document ....................................... 13 

2.0 NRC PARTICIPATION AND ACTIVITIES .................................. 15 

2.1 Tank Waste Remediation System Section ............................ 15 

2.2 Support Activities by the Center for Nuclear Waste Regulatory Analyses .... 16 

2.3 TW RS-P Standard Review Plan ................................... 17 

2.4 NRC Participation in Phase IA ..................................... 22 

2.5 NRC Participation in Phase IB-1 ................................... 24 

3.0 NRC AREAS OF REVIEW ............................................. 29 

3.1 Organization and General Information ............................... 29 

3.1.1 Tank Waste Remediation System-Privatization Organization ........ 29 

3.1.2 Meetings and Reports on Safety and Regulatory Activities .......... 30 

3.1.3 Inspections .............................................. 33 

3.1.4 Sum m ary ............................................... 35 

3.2 Seismic and Structural Considerations ............................... 37 

3.2.1 Tank Waste Remediation System-Privatization: Part A ............ 37 

3.2.2 Tank Waste Remediation System-Privatization: Part B-1 ........... 43 

3.2.3 Present Status ........................................... 46 

3.3 Hazards and Safety Analyses ..................................... 47 

3.3.1 Standards Approval Package (Safety Requirements Document and 

Hazardous Analysis Report) and Initial Safety Analysis Report ...... 47 

3.3.2 Design Safety Features Submittal ............................ 48 

3.3.3 Related Activities and Topical Meetings ........................ 49 

3.3.4 NRC Analysis and Evaluation ................................ 50 

3.3.5 Present Status of Hazards and Safety Analyses ................. 54

V NUREG-1747
TABLE OF CONTENTS (cont.) 

3.4 Radiation Safety and Dose Assessment Methodology ................... 65 

3.4.1 Regula:tlory Limits for Occupational and Public Doses Applicable 

to Tank Waste Remediation System-Privatization .............. 65 

3.4.11 Introduction ...................................... 65 

3.4.1 .,2 Definitions ....................................... 66 

3.4.13... Lim its ........................................... 68 

3.4.1.4 Monitoring Thresholds .............................. 75 

3.4.2 Radiation Protection Program ............................... 75 

3.4.2.1 Introduction ...................................... 75 3.4.2.2 Description of the Radiation Protection Program .......... 77 

3.4.2.3 Regulatory Basis .................................. 78 

3.4.2.4. Organization of the RPP for Design .................... 81 3.4.2.5 Comparison of NRC and DOE ALARA Guidance .......... 86 

3.4.2.6 Radiation Protection Aspects from the NRC 

Standard Review Plan ............................ 87 

3.4.2.7 Sum m ary ........................................ 91 

3.5 Criticality Safety ............................................... 93 3.5.1 Criticality Safety Program Document and Requirements ........... 93 3.5.2 Interim Criticality Safety Evaluations ........................... 93 

3.5.3 Uncertainties and Unresolved Nuclear Criticality Safety Issues ...... 94 3.5.4 Conformance with NRC Regulations .......................... 94 

3.6 Process and Chemical Safety ..................................... 97 

3.6.1 Standards Approval Package (Safety Requirements Document and 

Hazard Analysis Report) and Initial Safety Analysis Report ....... 97 3.6.2 Design Safety Features Submittal ............................ 98 

3.6.3 Related Activities and Topical Meetings ........................ 99 

3.6.4 NRC Assessment of the Present Status ........................ 99 

3.7 Fire Protection .............................................. 109 

3.7.1 Standards Approval Package (Safety Requirements Document and 

Hazard Analysis Report) and Initial Safety Analysis Report ...... 109 3.7.2 Topical Meeting on Fire Protection (February 1999) .............. 109 

3.7.3 Level 1 Meeting on Fire Protection (September 1999) ............ 110 3.7.4 Level 1 Meeting on Fire Protection (March 2000) ................ 112 3.7.5 Present Status with Reference to Point Paper .................. 113 

3.8 Explosion Protaection Issues ...................................... 115 

3.8.1 Standar-ds Approval Package (Safety Requirements Document and 

Hazard Analysis Report) and Initial Safety Analysis Report ..... 115 

3.8.2 Design S-,afety Features Topical Meeting-Hydrogen Control 

(January 1999) ........................................ 115

NUREG-1747 vi
TABLE OF CONTENTS (cont.) 

3.8.3 Explosive Hazards Topical Meeting I (August 1999) .............. 116 

3.8.4 Explosive Hazards Topical Meeting II (September 1999) .......... 116 

3.8.5 NRC Assessment of Present Status .......................... 117 

3.9 Environmental Protection ........................................ 119 

3.9.1 Environmental Impact Statement ............................ 119 

3.9.2 Standards Approval Package (Safety Requirements Document and 

Hazard Analysis Report) and Initial Safety Analysis Report ...... 119 

3.9.3 Design Safety Features ................................... 120 

3.9.4 Related Activities and Topical Meetings ....................... 120 

3.9.5 NRC Assessment of the Present Status ....................... 122 

3.10 Quality Assurance and Management Measures ....................... 123 

3.10.1 Introduction ............................................ 123 

3.10.2 DOE QA Requirements and Guidance ........................ 123 

3.10.3 NRC QA Regulations and Guidance .......................... 124 

3.10.4 Comparison of QA Regulatory Requirements and Guidance ....... 127 

3.10.5 BNFL Inc. QA Program for TWRS-P ......................... 128 

3.10.6 NRC Assessment of Current Status and Conclusions ............ 131 

4.0 NRC OBSERVATIONS AND CONCLUSIONS ON THE REGULATORY 

A PPROAC H ..................................................... 133 

4.1 Programmatic Influence Upon DOE Regulatory Activities ............... 133 

4.2 Inadequate Maintenance of Design and Authorization Basis Documents .... 135 

4.3 Use of a Risk-Based Approach ................................... 136 

4.4 Limited Use of NRC Regulations and Guidance ....................... 138 

5.0 POTENTIAL ISSUES FOR TRANSITION TO NRC REGULATION .............. 141 

5.1 Em ergency Planning ........................................... 141 

5.2 Regulatory Authority and Classification of Radioactive Wastes ........... 141 

5.3 Safeguards and Security ........................................ 142 

5.4 Regulation of Emissions ........................................ 142 

5.5 Co-located W orker (CLW ) ....................................... 142 

5.6 W aste Ownership ............................................. 143 

5.7 Payment for Transition to NRC Regulation and NRC Regulatory 

O versight .................................................. 143 

5.8 Tri-Party Agreem ent ............................................ 143 

5.9 DOE Stop W ork Authority ....................................... 143 

5.10 DOE Safety Oversight After Regulatory Transition .................... 144 

5.11 Application of NRC (10 CFR Part 2) Hearing Requests ................. 144 

5.12 Contractual Obligations for Feed Delivery ........................... 144 

5.13 Communication Plan with Stakeholders ............................. 144

vii NUREG-1747
TABLE OF CONTENTS (cont.) 

5.14 Cost Benefit of NRC Regulation ................................... 144 

5.15 Role of the E'efense Nuclear Facilities Safety Board ................... 145 

5.16 Price-Anderson Indemnification ................................... 145 

5.17 Timing to Avoid Delays from Regulatory Transition .................... 145 

5.18 DOE FinancIal Interests ......................................... 145 

5.19 Quality Assurance ............................................. 145 

5.20 Impacts on Otfher Hanford Site Facilities ............................ 146 

5.21 Responsibility for Occupational Safety .............................. 146 

5.22 Tailoring of Requirements ....................................... 146 

5.23 Achievemeni: of Adequate Safety .................................. 146 5.24 Effect of Schedule Delays On Overall Risk .......................... 147 

5.25 Resource Allocation Across TWRS-Related Facilities .................. 147 

5.26 Backfit/Retrofit ................................................ 148 

5.27 Scope of NR(C: Regulation ....................................... 148 

5.28 Location of the Regulator ........................................ 149 

5.29 M ulti-Step Licensing ............................................ 149 5.30 Differences in 10 CFR Part 20 and 10 CFR Part 835 ................... 149 

5.31 Cost of Contractor Documentation ................................. 149 

5.32 The Role of [)'DE Orders Versus NRC Rules ......................... 150 5.33 The Role of National Environmental Policy Act in Regulatory Transition . .. . 150 

5.34 Role of the ACRS or ACNW in Regulatory Oversight Transition .......... 150 

5.35 Impact on Fut:ure Generations .................................... 151 

5.36 Public Input to Regulatory Decisions ............................... 151 

6.0 FUTURE ITEMS, EVENTS, AND CONCERNS ............................. 153 

7.0 MAIN REFERENCE"S ................................................ 155 

APPENDICES 

APPENDIX A: SUMMARY OF ISSUES ........................................ 157 

APPENDIX B: MOU BETWEEN DOE AND NRC ................................ 199 

APPENDIX C: POINT PAPER SUMMARIES ................................... 207 

APPENDIX D: DISCUSSIOI\I OF POTENTIAL ISSUES FOR REGULATORY 

TRANSITIO N ................................................ 211 

APPENDIX E: LISTING OF NRC CORRESPONDENCE WITH DOE ON TWRS-P ....... 265

NUREG-1 747 viii
EXECUTIVE SUMMARY 

Overview and Summary of NRC Involvement with DOE in the 

Tank Waste Remediation System-Privatization Program 

THE TANK WASTE REMEDIATION SYSTEM-PRIVATIZATION PROGRAM 

In 1995, the U.S. Department of Energy (DOE) embarked on an effort to privatize the 

processing through vitrification of 54 million gallons of radioactive waste that has been stored in 

177 underground storage tanks at the Hanford Site. Under the initial phase of the Privatization 

plan, fixed-price waste treatment services for processing a portion of the waste was to be 

supplied, on leased land at the Hanford Site, by contractor-owned, contractor-operated facilities 

under a fixed-priced contract. 

DOE established a dedicated Regulatory Unit (RU) led by a Regulatory Official (RO) at the 

DOE Richland Operations Office with regulatory authority exclusive to the regulation of TWRS

P contractors. The RO reported directly to the Manager of DOE/Richland Operations Office 

(RL) at a level equivalent to the DOE Program Manager for TWRS. The RU planned on 

following the five principles of good regulation as articulated by the NRC - independence, 

openness, efficiency, clarity, and reliability. The basic concept of DOE's regulatory approach at 

TWRS-P was that the contractor is responsible for achieving adequate safety, complying with 

applicable laws and regulations, and conforming with top-level safety standards and principles 

stipulated by DOE. Consistent with applicable laws and regulations, the contractor is required 

to tailor the exercise of this responsibility to the specific hazards associated with its activities, 

and is encouraged to do this in a cost-effective manner that applies best commercial practices. 

TWRS-P contractors have the responsibility to identify and recommend to DOE the set of 

standards, regulations, and requirements necessary to ensure adequate safety at the proposed 

facilities. This constitutes a risk-based, integrated safety management (ISM) process. DOE's 

responsibility is to execute the regulatory process, including authorization of contractor actions 

and confirmation that the contractor activities are performed safely and within approved limits. 

The authority of the RU to regulate a TWRS-P contractor is derived from the terms of the 

TWRS-P contract ("regulate by the contract"). 

The U.S. Nuclear Regulatory Commission (NRC) provided assistance to DOE on the TWRS-P 

program for 3 1⁄2 years under a Memorandum of Understanding (MOU) signed in January 1997. 

The MOU provided for NRC to acquire an understanding of the wastes and potential treatment 

processes, assist the DOE in performing reviews in a manner consistent with the NRC's 

regulatory approach for commercial nuclear facilities, and develop an effective regulatory 

program for the potential transition to NRC regulatory authority at a future time. In May 2000, 

DOE abandoned the privatization approach for cost reasons, and declared its intent to pursue 

a more conventional, management and operations (M&O) style contract for the design, 

construction, and operation of the waste treatment facilities. The M&O contractor may or may 

not use the designs, technologies, and approaches already developed. With this contract 

change, DOE also signaled its intent to self-regulate the facilities for the foreseeable future, 

without any schedule for transition to NRC regulatory authority. As a result of these changes, 

the NRC has re-evaluated its role in the program. The NRC has determined that DOE's 

decision to terminate the privatization contractor and pursue the M&O approach effectively

ix NUREG-1747
completes the MOU by its own terms. Consequently, NRC participation has ceased. However, the NRC remains willing to discuss possible arrangements with DOE for NRC involvement in 

the Hanford tank waste programs in the future, under a new MOU, if there is again a need for 

NRC expertise and if the NIRC sees benefit in such involvement, such as the potential transition 

to NRC regulatory authority. 

This report summarizes NRC,'s participation in and observations on the TWRS-P program, identifies potential open issues from the perspective of the NRC staff, and notes concerns with deviations from the NRC's regulatory approach. Unless otherwise noted, this report is based 

upon the design as it existed in June 2000. 

Section 1 provides more information on the program. 

NRC PARTICIPATION 

When NRC began its involvement, the TWRS-P program was initially designed to begin with a relatively small pilot plant approach and facility for early processing of the wastes. Such an approach would have allowed verification of design and technical approaches with minimal 

economic, programmatic, and safety risks, and would still have resulted in the processing of 

some of the waste materials. However, due to programmatic changes, including concerns 

regarding the feasibility of privately financing a short-term facility, DOE decided to pursue a much larger, full-scale facility instead of a pilot plant. This decision greatly increased the flow 

rates and radiochemical inventories for the proposed facility and contributed to several design basis/authorization issues encountered during the program. 

In carrying out its responsibiiities under the MOU, NRC staff participated with DOE in technical 

reviews and meetings of various contractor submittals including, for example, Safety Requirements Documents (SRDs), Hazard Analysis Reports (HARs), Initial Safety Analysis Reports (ISARs), Design Safety Features (DSF) submittals, and the Firm Fixed Price (FFP) submittal. NRC staff also reviewed numerous other documents on specific features and 

concerns (e.g., seismic design, quality assurance/quality control (QA/QC), radiological plans, fire protection, chemical safety, etc.) and attended many safety and regulatory meetings with 

DOE and DOE's contractors (e.g., monthly Topical Meetings, design review meetings, etc.). Oral and written comments were provided by NRC staff to DOE as a result of these reviews and 

participation in these meetings. NRC staff also assisted DOE in the development of appropriate regulatory guidance and the NRC issued a final Standard Review Plan (SRP) for TWRS-P facilities for use in any future NRC regulatory oversight of the TWRS-P facilities. 

While participating in this program, NRC staff became fully cognizant of the waste issues, design requirements, safety, and regulation of the proposed facility, thus meeting the NRC's 

primary objectives of the MOU. 

To supplement their reviews of contractor and DOE documentation, and to gain better insights into the environmental, safety, and health (ES&H) characteristics of the proposed TWRS-P 

facilities, NRC staff analyzed both unmitigated and mitigated consequences from potential accident scenarios at a generic TWRS-P facility with similar characteristics and operations to 

those proposed by the TWRES-P contractors. The analyses used a conservative approach considered to be suitable for safety categorizations and preliminary designs. Several scenarios involving large radiochemical inventories (in tanks), flammable gases, organic ion exchange 

NUREG-1747 x
resin interactions, glass melters, and cold chemical effects were found to have potential 

accident consequences to the workers and the public of significant severity and risk (1 E-2/yr to 

1 E-4/yr). Under such circumstances, the NRC staff concluded that accident prevention 

(reduced probability) and mitigation (reduced consequences) would become necessary, 

requiring the identification of improved design approaches and items relied on for safety. 

Ideally, processes and approaches proposed for tank waste processing would incorporate 

robust designs with redundant features. The NRC staff noted that suitable process accident 

prevention and mitigation methods exist that are compatible with the regulations and offer the 

potential for reducing process accident risk to more acceptable levels (circa 2E-6/yr). 

Furthermore, the NRC staff found that relatively standard nuclear industry methods (e.g., high 

efficiency particulate air filter systems) could be used for risk reduction; with the possible 

exception of the melter areas, no unique or new risk reduction methods that might require 

qualification appeared to be necessary. NRC staff's efforts identified several key areas of 

uncertainty, such as melter failure modes and frequencies, unique design features, and 

corrosion resistant requirements, that would require further study before more refined analyses 

could be performed. The insights gained from these efforts allowed the NRC staff to better 

understand the potential safety issues and risk control strategies associated with the TWRS-P 

program. 

Chapters 2.0 and 3.0 of this report provide more information on the NRC involvement in TWRS

P and NRC assessment of several areas of review. 

POTENTIAL OPEN ISSUES 

As a result of the NRC staff's technical review of documentation and participation in meetings 

with DOE and the contractors, several concerns and potential open issues were identified. 

These include the need for significantly more detailed design information and safety analyses, 

and greater defense in depth. In particular, the design at the time of termination of the 

privatization contract was found to be very preliminary and corresponded to perhaps a 15 

percent level of design. The NRC staff has identified over two-dozen significant issues and 

over fifty specific topics in the current design and approach that would require further efforts 

and analysis to achieve adequate closure; these issues are discussed further in Chapter 3.0 of 

this report and in Appendix A. These significant issues include both programmatic aspects of 

TWRS-P (e.g., maintenance of design/authorization basis, level of detail) and technical issues 

(e.g., large volumes of tankage and radionuclide inventories, combined chemical and 

radiological hazards, melter corrosion). DOE, as the current regulator, has also identified 

similar issues. 

The melters present several issues, due to their size, capacities, and surface area fluxes, all of 

which would become the largest for radwaste vitrification in the world. However, the 

experiential base, particularly from the perspective of potential ES&H concerns, is limited. 

Towards the end of the program, the Contractor identified the need for high nickel alloys for 

corrosion resistance in areas of the melter that would usually be made of more conventional 

materials (e.g., carbon steel) in existing vitrification facilities. This was based upon testing a 

one-third scale melter, but no further information was made available. The Contractor also 

presented analyses that implied a relatively high level of risk to the worker (circa 1 E-3/yr) from a 

melter offgas/NOx scenario. The melter designs also have several unique attributes, including 

a thin gap between the cooling coils and the outer steel casing, and drainage holes. More

Xi NUREG-1747
information and analyses would be required to ascertain the safety ramifications if these melter 

designs are used by the new contractors. 

DOE prescribed an expeditedl schedule at the beginning of the program, with limited flexibility. Consequently, throughout the length of the program, the NRC and DOE staff technical reviews 

were held 1to tight schedules (typically a 2 week turnaround for a multivolume submittal) which 

frequently resulted in the inability to identify action items and plans, and achieve full closure on 

a number of the issues. Consequently, resolution of several significant design and safety issues (such as those discuLssed in Chapter 3.0 and Appendix A of this report) may not occur 

for some time. In addition, the likely impacts from further contractor changes are unclear but would likely imply more uncertainties and more design changes that, in turn, could raise more 

issues and the corresponding need for additional time for review and resolution prior to 

proceeding into construction and operation. 

POTENTIAL CONCERNS IKiTH DIFFERENCES FROM THE NRC'S REGULATORY 

APPROACH 

The working relationship between the NRC and the DOE has evolved during the program, and DOE has acknowledged the value added by the NRC's involvement. In the opinion of the NRC 

staff, there are several significant concerns which appear to be having a deleterious effect upon DOE's regulatory approach. These are discussed further in Chapter 4.0 of this report. The 

most notable of these are summarized as follows: 

1. The influence of programmatic issues (including cost, schedule, and capacitv) uMon the 

regulatory review activities: Programmatic issues, including economics, arose in several 

regulatory and safety-related meetings, usually in conjunction with the discussions 

regarding safety-related components and systems, and defense-in-depth. Short schedules 

(typically 2 weeks) were established by DOE for reviewing large, multivolume submittals 

and may have impacted the depth and quality of the reviews, including the identification of 

safety and regulatory concerns. (DOE-Headquarters also expressed concerns about 

schedule pressures - see page 28.) In addition, DOE's programmatic concerns and 

desires emphasized higher throughputs (potentially a four-fold or larger increase) and/or 

additional/larger facilities to increase waste processing rates and reduce schedules and 

costs. Many design changes were made to accommodate this emphasis on cost and 

schedule. However, the safety analyses do not appear to have reviewed the potential impacts from such higher throughputs or additional/larger facilities. The emphasis also 

appeared to contribute to the deferral of some issues to subsequent reviews in order to 

maintain the schedule. Thus, the design work often continued with the potential for less 

than adequate consideration and closure of regulatory issues, such as ALARA (as low as 

reasonably achievable) (,e.g., see Chapter 3.0 and Appendix A of this report). 

2. Maintenance of desiqn/authorization basis (license): Throughout most of the design effort 

(about 2 years), the design and safety teams of the Contractor (BNFL Inc.) worked quasi independently, partly because of the previously mentioned emphasis on schedule. This led 

to inconsistencies between the design and the safety documentation. Changes in 

fundamental aspects of the design occurred in this time period, potentially without 

adequate consideration of regulatory needs, such as inventories and source terms for the

NUREG-1747 xii
safety analyses. The design and authorization basis documents were not updated and 

few amendment requests were submitted to the RU. Ultimately, the RU delivered a 

Corrective Action Notice (CAN) on this concern in March 2000, which, in turn, led to a 

Corrective Action Plan (CAP) that was agreed to between the RU and the Contractor. A 

significant number of amendment requests were received by the RU from the Contractor 

after the CAN. However, by this time, most of the preliminary design activities in this 

phase of the program were completed and this timing of the CAN provided little room for 

regulatory review of the modifications and regulatory impact upon the design and design 

activities. Ironically, the flexible regulatory framework may have contributed to this 

situation; DOE postponed inspections for about a year that could have identified this 

situation and that could have potentially endorsed a CAP for correcting the situation. 

3. The application of a risk-based approach to the development of the design without 

additional considerations: Risk-based analyses were used as the basis for the ISM 

process, which includes hazards identification, consequence estimation, and control 

mitigation. This is essentially a completely fluid process without a minimum level of 

requirements and, as practiced on TWRS-P, did not appear to adequately address 

unknowns, uncertainties, errors, proven practices, future plans, and experience. The ISM 

approach applied to TWRS-P resulted in less conservatism, reduced margins, and less 

defense-in-depth during the preliminary design phase - a phase when margins and 

conservatism would normally be relatively larger. There appeared to be more emphasis on 

the process, and less on the results. The Integrated Safety Management Plan (ISMP) is 

an iterative process which includes reassessing assumptions (e.g., of radionuclide 

concentrations) that can reduce a hazard below a limit into a bin with less reliability and 

safety requirements. Once this occurs, the scenario represents less risk with fewer safety 

requirements, and, again, since ISMP focuses on events with higher risk, this scenario and 

the underlying assumptions may be subject to less scrutiny. An unintended consequence 

of ISM at TWRS-P accrued from this circular logic: ISM focused on the higher risk areas by 

challenging assumptions; assumptions that were changed and resulted in lower 

consequences and risks may not have been revisited or re-evaluated as thoroughly. Thus, 

reduced consequence estimates may have resulted, and control strategies and equipment 

may not have been adequately identified. In contrast, the NRC regulatory approach 

applies a risk-informed, performance-based approach with defense-in-depth, appropriate 

levels of conservatism, and a minimum set of standards and requirements that are codified 

in the regulations. 

4. Limited use of NRC regulations and guidance: DOE has adopted the NRC principles of 

good regulation in the documents that form the basis for the TWRS-P regulatory approach. 

However, DOE has not adopted the use of NRC regulations and guidance for TWRS-P, 

such as 10 CFR Part 70 and the TWRS Standard Review Plan. Instead, standards, codes, 

and regulations were selected by the Contractor and approved for their use by the RU with 

the application of ISM. 

REGULATORY TRANSITION ISSUES 

The NRC and the DOE have previously discussed issues related to the potential regulatory 

transition of TWRS-P to NRC regulation in the near future. Many of these issues are 

summarized in Chapter 5.0 and discussed in detail in Appendix D of this report, and have been

xiii NUREG-1747
discussed between the DOE and NRC over the 3-year length of the program. DOE is converting the contracts to a'tn M&O arrangement for TWRS, which has been renamed the 

RPP/WTP (River Protection Project/ Waste Treatment Plant). The NRC staff believes that 

enabling legislation from Congress is desirable for the NRC to regulate either a privatized TWRS-P facility or an RPP/WTP and that any resulting issues are resolvable. From the viewpoint of the NRC staff, -rost of the regulatory issues would be addressed by the legislation that enables NRC regulator( authority over the TWRS/WTP facilities or NRC external regulation of DOE facilities, and by continued refinement and detailing of the proposed facility designs. The remaining issues relate to DOE programmatic activities and not regulation. 

THE FUTURE TWRS/WTP PROGRAM 

As previously noted, DOE has terminated the current privatization contract and approach and elected to follow an M&O contracting approach. According to the Request for Proposals (RFP), DOE plans to regulate these 'TWRS facilities. The specific features of the regulatory approach and the balancing of programmatic and safety issues are not identified as of this writing (December 2000). The mearns to follow, address, and close the design, safety, and regulatory issues identified from the NF:tC reviews and summarized in this report also have not been presented at this time. The RIFP does include significant incentives for a contractor to reduce 

costs but does not mention safety as an evaluation factor. 

The DOE has acknowledged the value added to the program by the NRC's participation. However, the NRC has terminated its involvement with the TWRS Project and has deployed staff to other projects. The NRC has also informed DOE that it is willing to discuss possible arrangements for NRC involvement in the future, if there is again a need for NRC expertise and if the NRC sees a benefit frcrn such involvement, such as the potential transition to NRC 

regulatory authority

NUREG-1747 xiV
ABBREVIATIONS AND ACRONYMS

ACGIH American Council of Government and 

Industrial Hygienists 

ACNW Advisory Committee on Nuclear Waste 

ACRS Advisory Committee on Reactor Safeguards 

AEA Atomic Energy Act 

AIChE American Institute of Chemical Engineers 

ALARA As Low as Reasonably Achievable 

ALl Annual Limit on Intake 

AMSQ Office of Assistant Manager for 

Environmental, Safety, Health, and Quality 

ANS American Nuclear Society 

ANSI American National Standards Institute 

API American Petroleum Institute 

ARCHIE Automated Resource for Chemical Hazard 

Incident Evaluation 

ARF Airborne Release Fraction 

ASC ALARA Subcommittee 

ASME American Society of Mechanical Engineers 

BARCT Best Available Radionuclide Control 

Technology 

BAT Best Achievable Technology 

BDC Baseline Design Criteria 

BEI BNFL Engineering Inc., United Kingdom 

BIO Basis for Interim Operation 

BNFL Inc. British Nuclear Fuels Limited Inc. 

CAA Clean Air Act 

CAN Corrective Action Notice 

CAP Corrective Action Plan 

CAR Construction Authorization Request 

CBA Cost Benefit Analysis 

CCB Consumable Changeout Box 

CCPS Center for Chemical Process Safety 

CERCLA Comprehensive Environmental Response, 

Compensation, and Liability Act 

CFR Code of Federal Regulations 

CLW Co-Located Worker 

CNWRA Center for Nuclear Waste Regulatory 

Analysis 

COCO Company-Owned, Company-Operated

DAC 

DBE 

DCS 

D&D 

DEAR 

DF 

DID 

DNFSB 

DOE 

DSF 

DST 

DWPF

Derived Air Concentrations 

Design Basis Earthquake 

Distributed Control System 

Decontamination and Decommissioning 

Department of Energy Acquisition 

Regulations 

Decontamination Factor 

Defense-in-Depth 

Defense Nuclear Facilities Safety Board 

United States Department of Energy 

Design Safety Features 

Double Shell Tanks 

Defense Waste Processing Facility

EDG Emergency Diesel Generator 

EH Office of Environment, Safety, 

and Health/DOE 

EIS Environmental Impact Statement 

EM Office of Environmental Management/DOE 

EP Environmental Protection OR 

Emergency Plan 

EPA U.S. Environmental Protection Agency 

ERDF Environmental Restoration Disposal Facility 

ERPG Emergency Response Planning Guideline 

ERPP Environmental Radiation Protection Program 

ES&H Environment, Safety, and Health 

FFP Firm Fixed Price (Contract) 

FFRDC Federally Funded Research and 

Development Center 

FHA Fire Hazards Analysis 

FTE Full Time Equivalent

GAO 

GDP 

GIGO 

GOCO 

GTCC 

HAB 

HAR 

HAZOP 

HEPA 

HFD 

HLW 

HVAC 

HWMA

General Accounting Office 

Gaseous Diffusion Plant 

Garbage-In, Garbage-Out 

Government-Owned, Contractor-Operated 

Greater than Class C 

Hanford Advisory Board 

Hazards Analysis Report 

Hazards Operability (method or analysis) 

High Efficiency Particulate Air 

Hanford Fire Department 

High Level Waste 

Heating, Ventilation, and Air Conditioning 

Hazardous Waste Management Act

I&C Instrumentation and Control 

IDLH Immediate Danger to Life and Health 

IEEE Institute of Electrical and Electronics 

Engineers 

IG Implementation Guide 

INEEL Idaho National Engineering and 

Environmental Laboratory 

IROFS Items Relied on for Safety 

ISA Integrated Safety Analysis 

ISAR Initial Safety Analysis Report 

ISM Integrated Safety Management 

ISMP Integrated Safety Management Plan 

ISO International Organization for Standardization 

ITS Important to Safety

LAW 

LCAR 

LCO 

LLW 

LMAES 

LNT

Low Activity Waste 

Limited Construction Authorization Request 

Limiting Condition of Operation 

Low Level Waste 

Lockheed Martin Advanced Environmental 

Systems 

Linear No Threshold

XV NUREG-1747
MC&A 

MCCMT 

Ml 

M&O 

MOA 

MOU 

MOX

Material Control and A,:counting 

Miscellaneous Cold Chemicals Mix Tank 

Mechanical Integrity 

Management and Operations (contract) 

Memorandum of Agreement 

Memorandum of Understanding 

Mixed Oxide

NDT Nondestructive Testincl 

NEI Nuclear Energy Institute 

NEPA National Environmental Policy Act 

NESHAPS National Emissions Standards for 

Hazardous Air Pollutants 

NFPA National Fire Protection Association 

NIOSH National Institute for Occupational 

Safety and Health 

NMSS Office of Nuclear Matejial Safety and 

Safeguards/NRC 

NRC U. S. Nuclear Regulatciy Commission 

NRPB National Radiation Prol:ection Board 

OAR Operations Authorizatior Request 

ORP Office of River Protecticri/DOE 

OSHA Occupational Safety arid Health 

Administration 

OSR Office of Safety Regulalion/DOE OR 

Operational Safety F.equirements 

PEL Permissable Exposure Levels 

P&ID Piping and Instrumentation Drawings 

PPE Personnel Protective Elquipment 

PRA Probabilistic Risk Analysis OR 

Probabilistic Risk As-;essment 

PSAR Preliminary Safety Analysis Report 

PSHA Preliminary Seismic Hazard Analysis

QA 

QAP 

QAPIP 

QC

Quality Assurance 

Quality Assurance Prccjram 

Quality Assurance Prog cram Implementation 

Plan 

Quality Control

SER Senior Expert Reviewers OR 

Safety Evaluation Report 

SGI Safeguards Information 

SIP State Implementation Plan 

SIS Safety Instrumentation Systems 

SNF Spent Nuclear Fuel 

SNM Special Nuclear Material 

SQEP Suitably Qualified and Experienced Person 

SRD Safety Requirements Document 

SRP Standard Review Plan 

SRS Savannah River Site 

SSC Structures, Systems, and Components 

SST Single Shell Tanks 

TAN Test Area North 

TBD To Be Determined 

TEDE Total Effective Dose Equivalent 

TLV Threshold Limit Value 

TPA Tri-Party Agreement 

TRU Transuranic (waste or isotopes) 

TSD Treatment, Storage, and Disposal 

TWA Time Weighted Average 

TWRS Tank Waste Remediation System 

TWRS-P Tank Waste Remediation System

Privatization 

TWRS/ Tank Waste Remediation System/ 

WTP Waste Treatment Plant

UBC 

UCNI 

UF 

WAC 

WDOH 

WPPSS 

WTP 

WVDP

Uniform Building Code 

Unclassified Controlled Nuclear Information 

Ultrafiltration 

Washington Administrative Code 

State of Washington Department of Health 

Washington Public Power Supply System 

Waste Treatment Plant 

West Valley Demonstration Project

RCM Radiological Control Manual 

RCRA Resource Conservation Recovery Act 

of 1976 

RF Respirable Fraction 

RFP Request for Proposals 

RG Regulatory Guide 

RIPB Risk-Informed Performance Based 

RL Richland Operations Office/DOE 

RMP Risk Management Program 

RO Regulatory Official 

RPP River Protection Project/DOE OR 

Radiation Protection Plan 

RPP/WTP River Protection Project/ 

Waste Treatment Plan-: 

RU Regulatory Unit/DOE 

SAP Standards Application Fackage OR 

Standards Approval Package 

SCR Selective Catalytic Reduction

NUREG-1747 xvi
GLOSSARY OF TERMS

BNFL Inc.

"Contractor" 

Defensive Waste 

Processing Facility 

Office of Environmental 

Management 

High Level Waste 

Incidental Waste 

Low Activity Waste 

Low Level Waste 

Richland Operations 

River Protection Project 

Regulatory Unit

British Nuclear Fuels Limited, Inc., the U.S. subsidiary of BNFL plc 

and a TWRS-P contractor for Phases IA and IB-I. Often referred to 

as the "Contractor" in this report. 

This term refers to a specific, TWRS-P contractor (usually BNFL Inc.) 

in this report. 

A high level waste vitrification plant at the Savannah River Site 

DOE Headquarters Office responsible for DOE Complex-wide 

environmental and waste management activities. 

A term used to describe special nuclear fuel, first-cycle special 

nuclear fuel processing wastes and concentrates, and (for Tank 

Waste Remediation System) the solid phases in the tank wastes. 

High level waste requires disposal in a geologic repository. 

A radioactive waste stream(s) comprised primarily of contaminated 

materials produced incidental to high level waste processing, such as 

spent resins, loaded filters, broken melters and equipment, and 

treated low activity waste. If the radiation levels are sufficiently low 

(generally interpreted as meeting the criteria for low level waste in 10 

CFR Part 61), incidental waste may be treated and sent to disposal in 

near-surface facilities. 

A term used by DOE to describe the predominantly liquid portion of 

tank waste. Untreated low activity waste (LAW) is similar to high level 

waste in terms of environment, safety and health effects and requires 

geologic disposal. Treated LAW may be capable of meeting 

incidental waste criteria and, thus, it may be suitable for near surface 

disposal like low level waste. 

A term for radioactive wastes that pose significantly lower radiological 

risks and of relatively short duration, such that the wastes are 

generally suitable for near-surface disposal per 10 CFR Part 61. 

A DOE office, located adjacent to the Hanford site. This office reports 

to the DOE Office of Environmental Management. 

DOE, Office of River Protection's program for activities that protect 

the Columbia River. 

Office of Safety Regulation of the River Protection Project/Waste 

Treatment Plant Contractor-DOE element responsible for regulating

xvii NUREG-1747
Tank Waste Remedia

tion Systern-Privatization

the treatment and vitrification facilities. This office reports to DOE's 

Richland Operations Office. 

Refers to the private capital financed processing and vitrification 

facilities under the previous contracts held by BNFL Inc. and 

Lockheed Martin Advanced Environmental Systems.

Tank Waste Remedia Re:flerring to the processing and vitrification facilities that will be built 

tion Systern-Privatization under the new contract(s) at Hanford. 

/Waste Treatment Plant

NUREG-1747 xviii
1.0 INTRODUCTION AND BACKGROUND 

1.1 HANFORD AND THE TANK WASTES 

The U.S. Department of Energy (DOE) established the Tank Waste Remediation System 

(TWRS) program at the Hanford site to manage, retrieve, treat, encapsulate/immobilize, and 

disposition radioactive waste materials from the 177 underground waste storage tanks onsite in 

a safe, environmentally sound, and cost effective manner. These tanks primarily contain high 

level wastes (HLW) and chemical species from processing spent nuclear fuels for more than 40 

years at the site (see Chapter 7.0, Main Reference 1). There are 149 single shell tanks (SSTs) 

and 28 double shell tanks (DSTs). There are several tank sizes but the average tank has about 

one million gallons of capacity. Both SSTs and DSTs are manufactured from carbon steels. 

However, the DSTs are newer, have more provisions for monitoring the wastes, and include an 

annulus for leak detection and confinement. To date, no DST has been confirmed to leak. In 

contrast, approximately 67 SSTs have been confirmed as leakers. 

The tank contents consist of mixtures of materials from some eight major processes. Some of 

the wastes date back to 1944. Even though the radiation levels are high (typically exceeding 

100 R/hr in the tank dome spaces and through riser connections), the great majority of the 

waste constituents are nonradioactive and contain some 240,000 tonnes of processed 

chemicals. The tanks hold approximately 54 million gallons of waste and amount to over 200 

million-plus curies of radioactivity, primarily from cesium and strontium but with smaller 

contributions from other fission products and transuranic (TRU) isotopes. Physically, the tank 

contents exist as liquids, sludges, salts, saltcakes, and mixtures thereof, and some tanks 

periodically release gas mixtures. The SSTs contain primarily sludges and saltcakes with 

relatively little liquids - most of the liquid phase has been removed due to concerns about 

potential leaks. The DSTs contain most of the liquids but also have solid phases. The wastes 

stored in the tanks are defined as high level waste (HLW; per 10 CFR Part 50, Appendix F) and 

hazardous waste (per RCRA - Resource Conservation and Recovery Act - with various codes). 

DOE categorizes the wastes to simplify contractual and processing approaches 1 . DOE uses 

the term LAW to denote "Low Activity Waste." Table 1 presents summary information on the 

composition of LAW. LAW is predominantly a liquid phase with soluble species such as 

nitrates and cesium; it may also contain up to 2 percent suspended solids or solids otherwise 

entrained by the waste transfers. Three envelopes of LAW have been defined; Envelope A is 

"standard," Envelope B contains higher levels of cesium, and Envelope C contains higher levels 

of strontium and TRU. The contract (Footnote 1) identifies ranges for chemical and radioactive 

species in these LAW envelopes. LAW would come from the liquid phases of the DSTs and 

from solids washing operations. From a regulatory perspective, LAW is still HLW and has high 

radiation levels requiring handling within shielded structures. DOE identifies the solid phases 

as "HLW," defined as Envelope D. Table 2 provides summary compositional information on 

HLW. Envelope D contains cesium, strontium, and TRUs as the radionuclides. Metal oxides, 

hydroxides, nitrates, phosphates, and aluminates constitute the bulk of the chemical species. 

The contract (Footnote 1) provides ranges for the chemical and radioactive species in Envelope 

1 See, for example, Department of Energy (U.S.) (DOE). Contract No. DE-RP-96RL1 3308, "TWRS Privatization." 

DOE: Richland, Washington. August 1998.

1 NUREG-1747
D. Envelope D is assumed Io be transferred as a slurry in concentrations up to 20 percent, from the removal of solid phases from the SSTs and DSTs. The solids in the LAW envelopes 

would have a composition similar to Envelope D. 

Figure 1 provides a conceptual overview of DOE's approach to tank waste treatment. LAW 

envelopes would be transferred to a treatment plant. The LAW would be pretreated to separate the radionuclides (primarily, ,;esium, strontium, technetium, and TRU, and the suspended solids) from the remainder of the waste envelope. The separated radionuclides would be stored for an 

interim period of several years. Pretreatment reduces the level of radioactivity in the treated 

LAW to levels commensurate with near-surface disposal requirements (essentially equivalent to 

the Class AVB/C definitions (A low level waste in 10 CFR Part 61). The less radioactive, treated 

LAW would be vitrified and placed into stainless steel containers for long term storage or 

disposal at Hanford. The HI.UN (Envelope D) would be treated and washed using a filter or 

other device to separate the liquid phase from the slurry. The liquid phase would be routed to 

pretreatment and combined with the LAW, primarily for cesium and technetium removal. The 

treated HLW would be combined with the separated radionuclides from LAW processing and 

vitrified in an HLW melter. The HLW glass would be stored at Hanford in stainless steel 

canisters until subsequent disposal in an HLW repository. 

Table 1: Summary Information on LAW Radionuclide Composition 

Maximuii Ratio, Bq/mole Sodium Curies/Liter at 10 Molar Sodium 

Radionuclide v i i o J v _______Enveloe SA J Envelope B jEnvelope C ][EveoA A• Envelope B J Eneop = 

TRU 4.8E5 4.8E5 3.0E6 1.3E-04 1.3E-04 8.11 E-04 

Co-60 6.1E4 6.1E4 3.7E5 1.65E-05 1.65E-05 1.OE-04 

Sr-O 4.4E7 4.4E7 8.0E8 1.19E-02 1.19E-02 2.16E-01 

Te-99 7.1 E6 7.1 E6 7.1 E6 1.92E-03 1.92E-03 1.92E-03 

Cs-1 4.3E9 2.OE10 4.3E9 1.16E+00 6.OOE+O0 1.16E+00 - - ,, _(contract max.) 

Eu-154+ 1.2E6 1.2E6 4.3E6 3.24E-04 3.24E-04 1.16E-03 

Eu-I 55 1 

No contribution from the suspended and entrained solids is included in this table. 

LAW envelopes may contain up to 2 percent solids, which are assumed to be 

HLW solids. The solids contribution to radiotoxicity is significant and amounts to 

approximately 90 percent of the unit liter dose.

NUREG-1747 2
Table 2: Summary Information on HLW Radionuclide Composition (Reference 6)

1.2 THE TRI-PARTY AGREEMENT 

The Hanford Federal Facility Agreement and Consent Order (also known as the Tri-Party 

Agreement, or TPA)2 is a legal agreement between the DOE, the U.S. Environmental Protection 

Agency (EPA), and the State of Washington Department of Ecology (usually referred to as 

2 Environmental Protection Agency (U.S.)(EPA). EPA Docket Number 1089-03-04-120; Ecology Docket Number 

89-54; "Hanford Federal Facility Agreement and Consent Order." EPA: Washington, D.C. May 15, 1989.

NUREG-1747

Isotope Ciliter Isotope J Ci/liter Isotope Cl/liter 

H-3 1.30E-04 Od-li1M (NS) Eu-152 9.60E-04 

0-14 1.30E-05 Sn-1 1 9m (NS) Eu-1.54 1.04E-01 

Fe-55 (NS) Sn-i21m (NS) Eu- 55 5.80E-02 

Ni-59 (NS) Sn-I28 3.00E-04 U-233 1.80E-06 

Co-60 2.OOE-02 Sb-124 (NS) U-235 5.00E-07 

Ni-63 (NS) Sb-1126 (NS) U-236 (NS) 

Se-79 (NS) Sb-126m (NS) U-238 (NS) 

Sr-90 2.00E+01 Sb-i25 6.40E-02 Np-237 1.48E-04 

Y-90 (NS) Te-125mn (NS) Pu-238 7.OOE-04 

Nb-93m (NS) 1-129 5.80E-07 Pu-239 6.20E-03 

Zr-93 (NS) Cs-134 (NS) Pu-240 (NS) 

Tc-99 3.O0E-02 Cs-135 (NS) Pu-241 4.40E-02 

(NS) Cs-137 2.OOE+01 Pu-242 (NS) 

Rh-i 06 (NS) Ba- i37m (NS) Am-241 1.80E-01 

Pd-107 (NS) Ce-i44 (NS) Am-242 (NS) 

Ag-1ibm, (NS) Pr-i (NS) Am-242m (NS) 

Cd-i13m (NS) Pr-144m (NS) Am-243 (NS) 

ln-1 13m (NS) Pm-147 (NS) Cm-242 (NS) 

Sn-113 (NS) Sm-i151 (NS) Cm-243/244 6.00E-0

(NS) = Not Specified in the new contract. 

Feed concentration contains between 10 and 200 grams of unwashed solids per liter of solution. Values in the 

table are based upon the upper limit of 200 grams/liter, which is approximately a 20% slurry (the actual value is 

closer to 15 percent).

3
Ecology). The legal authority for the TPA arises from Resource Conservation Recovery Act of 1976 (RCRA) and the Comprehensive Environmental Response, Compensation, and Liability Act (CERCLA). The TPA contains provisions for the overall environmental management of the Hanford site, and includes provisions for the management of hazardous waste treatment, 

storage, and disposal (TSD) units and related permitting requirements. Consequently, the TPA 

incorporates provisions for TSD of the tank wastes. The TPA defines the respective roles, responsibilities, and interrelationships between the three parties. It delineates authorities, identifies enforcement provisions, and provides for dispute resolution among the parties. 

The TPA has an action plan for compliance that establishes milestones for the Hanford Site 

Cleanup. These milestones constitute a minimally acceptable level of progress. Failure to 

meet the milestones can result in lawsuits and fines against DOE. Currently, the most relevant 

milestones applicable to the tank wastes are: 

1. M-50-04: Start hot operations of HLW pretreatment facility by June 30, 2008. 

2. M-50-04-T01: Submit c:, iceptual design of HLW pretreatment facility by March 31, 1998. 

3. M-50-04-T02: Initiate definitive design of pretreatment facility by November 30, 1998. 

4. M-50-04-T03: Start construction of HLW pretreatment facility by June 30, 2001. 

5. M-51-00: Complete vitrification of Hanford HLW (tank waste) by December 31, 2028. 

6. M-51-03: Initiate hot ope;rations of the HLW vitrification facility by December 31, 2009. 

7. M-51-03-T01: Submit conceptual design of the HLW vitrification facility by December 31, 

1998. 

8. M-51-03-T02: Initiate definitive design of the HLW vitrification facility by December 31, 

1998. 

9. M-51-03-T04: Complete construction of the HLW vitrification facility by December 31, 

2007. 

10. Milestone M-61 -00: Complete pretreatment and immobilization of the Hanford LAW by 

December 2028. 

These milestones make for a tight schedule. The current design would treat half the waste by circa 2050, while the DOE program desires to have all of the waste treated by 2028, a 

difference in processing capacity of at least a factor of four 3. 

3 "Waste Management 2000," presentations by Mike Lawrence, BNFL Inc., and Dick French, U.S. Department of 

Energy. March 2000.

NUREG-1747 4
Figure 1: Overview of Hanford Processing Approach

HLW 4 

Vitrification

Precipitates + Cs/Tc

5 NUREG-1747
1.3 DOE PROGRAM AND CONTRACTS

DOE was pursuing a privatization initiative at Hanford for the construction and operation of contractor-owned, contractc,r-operated facility or facilities for treating these tank wastes and meeting the TPA milestones'. The concept was for DOE to enter into two or more firm fixed price (FFP) contracts for the: contractor(s) to build and operate a facility to treat the tank wastes 

according to DOE requirements. A minimum of two contractors would ensure that the 

government would receive ihe lowest price for waste TSD. The TWRS-Privatization Program was divided into two phases. Phase I was a proof-of-concept, commercial demonstration scale 

effort with the following objectives: 

1. Demonstrate the technical and business viability of using privatized contractors to treat 

Hanford tank wastes. 

2. Define and maintain adequate levels of radiological, nuclear, and process safety. 

3. Maintain environmental protection and compliance. 

4. Substantially reduce life-cycle costs and time for treating the wastes. 

The original plan called for processing between 6 and 13 percent of the tank waste in a pilot/demonstration facility as Phase I, and a subsequent, larger program would process the balance of the tank wastes as Phase Ii. Phase I consisted of two parts. Part A consisted of a 20-month development period to establish appropriate and necessary technical, operational, regulatory, business, and firiancial elements. This required the privatization contractors to select safety standards and requirements, formulate integrated safety management plans, and 

to generate conceptual designs and initial safety analyses, all of which would require approval by DOE. The contractors worked for 16 months to develop these items and the remaining 4 

months were used by DOE bor evaluation. 

DOE awarded contracts of $;;27 Million (each) for Part A to two teams, one led by British Nuclear Fuels Limited Inc. (BNFL Inc:.i, and the other led by Lockheed Martin Advanced Environmental 

Systems (LMAES). The BN1F'L Inc. team proposed a conceptual approach based upon the 

following operations (Chapter 7.0, Main References 2-5): 

1. Strontium/TRU coprecipilation from LAW. 

2. Suspended solids/stronl:ium/TRU removal by ultrafiltration from LAW. 

3. Two columns in series, organic ion exchange recovery of cesium from LAW. 4. Two columns in series, organic ion exchange recovery of technetium from LAW. 

5. Optional loading of radiocesium onto crystalline silicotitanate (CST). 

6. LAW vitrification in a joU le-heated melter. 

7. HLW washing and concentration by ultrafiltration. 

8. HLW vitrification in a joule-heated melter. 

9. NOx treatment by select:ve catalytic reduction (SCR) using anhydrous ammonia. 

4 Department of Energy (U.S.) (DO)E). RL/RU-2000-20, Rev. 0, "Regulatory Unit Position on Important to Safety Work Authorization for the RPP-W'-P interim Design Period." DOE: Richland, Washington. July 3, 2000.

NUREG-1747 6
Some aspects of the conceptual design were more detailed than other parts. For example, 

each ion exchange column would enclose 3.15 cubic meters, with an aspect ratio of about 7. In 

use, the columns would only contain about one cubic meter of resin each, for an effective 

aspect ratio of about 1.5-2. The approach included four ion exchange columns for cesium 

removal, arranged as two trains of two columns each. The technetium arrangement was 

identical. The design included two HLW and three LAW melters, and around 300,000 gallons 

of tankage. 

In terms of technology, BNFL Inc. relied upon proprietary organic ion exchange resins 

("superligands") and ultrafilters for separation of the radionuclides and solids from the LAW in 

pretreatment. Ultrafiltration was also planned as part of the treatment of HLW. BNFL Inc. 

proposed a liquid-fed slurry ceramic melter with joule heating for HLW vitrification, similar to the 

approach used at the West Valley Demonstration Project (WVDP) in New York State for 

vitrifying that site's HLW. LAW vitrification also used a joule-heated melter for the proposed 

facility, but it would be considerably larger than the HLW design. As an alternative, BNFL Inc. 

proposed storage and return of suspended solids to DOE without HLW vitrification; the cesium 

would be returned on CST. BNFL Inc. identified technology development in order to address 

unknowns about the new technologies involved. 

LMAES proposed a similar conceptual process approach with several variations (see 

Chapter 7.0, Main References 6-9): 

1. Suspended solids removal from LAW by centrifugation. 

2. Three columns in series, organic ion exchange recovery of cesium from LAW, regenerated 

by nitric acid and caustic. 

3. Three columns in series, organic ion exchange, polishing step recovery of cesium (from the 

effluent of the preceding step), with electrical regeneration. 

4. An inorganic, "guard" bed for cesium removal (on the effluent of the preceding step). 

5. Optional loading of radiocesium onto CST. 

6. Removal of technetium from LAW by electroplating. 

7. Strontium and TRU removal using ozone destruction of organics followed by precipitation 

and centrifugation. 

8. LAW vitrification in a joule-heated melter, augmented by fired burners (based upon 

oxygen-propane combustion) during startup and glass pouring. 

9. HLW washing and concentration by centrifugation. 

10. HLW vitrification in a cold-crucible, induction-heated melter. 

11. NOx treatment using SCR and ammonia from aqueous solution.

7 NUREG-1747
Each ion exchange column would be approximately 0.67 cubic meters, containing about 0.6 

cubic meters of resin, with a working aspect ratio of approximately four. The approach 

effectively used six columns; in series to remove the radioactive cesium. The design 

incorporated one HLW and three LAW melters, and around 200,000 gallons of tankage. 

In terms of technology, ion exchange technology was planned for cesium separation from LAW, 

using resins tested by Savannah River Site (SRS), with CST used as a guard column. 

Centrifuges separated suspended solids from the LAW. Cold crucible induction melting, similar 

to the process at La Hague (France) was planned for HLW vitrification. For LAW vitrification, 

LMAES proposed to use a direct fired melter, undergoing development on other types of DOE 

wastes. The approach planned to use electrolysis for technetium removal (by plating) and for 

recycling some of the ion exchange regeneration solutions. 

Phase I, Part A has been completed; the contractors each submitted a System Requirements 

Document (on standards), Hazards Analysis Report, Integrated Safety Management Plan, and 

Initial Safety Analysis Report, The DOE generated safety evaluation reports on the contractor 

submittals (Main References 10 and 11), indicating many concerns and open issues on each 

contractor's approach. For IMAES, the DOE review team concluded that the approach 

described in the ISAR would be capable of achieving subsequent authorizations for 

construction, operation, and deactivation, provided some 37 open issues would be resolved in 

the next regulatory submittal (Construction Authorization Request (CAR) - essentially a 

Preliminary Safety Analysis Report (PSAR)). Of these 37 issues, development of the 

pretreatment technology to maturity and its subsequent safe implementation were identified as 

the most significant uncertainties. For BNFL Inc., the DOE review team concluded that no 

individual regulatory issue wc uld prevent subsequent safety authorizations. However, the 

review identified 90 represenl:ative open regulatory issues that would require resolution, and the 

nature and number greatly challenged the reviewers to reach a consensus on the viability and 

sufficiency of some of the applroaches the Contractor proposed to achieve and maintain 

adequate safety through design and management practices. 

In May 1998, DOE said BNFL. Inc. had presented a superior plan for TWRS-P because the 

LMAES proposal posed an '"nacceptably high technical risk" of failure to meet DOE's cleanup 

goals 5 . In September 1998, DOE entered into a revised contract with BNFL Inc. In addition to 

modifying the original intent of two contractors and competition, DOE changed the program to 

accommodate larger production scale facilities in Phase I that would have a 30-40 year useful 

life. Phase I, Part B-1 involved a 24-month facility design phase that would advance the design 

to approximately a "30 percent level" and have the Contractor prepared to start construction 

and obtain financing'. Furthermore, in this 24-month period, DOE and BNFL Inc. were to refine 

the technical requirements, s.-,ubmit regulatory permitting applications, and finalize fixed unit 

prices for treated wastes. This Phase I, Part B-1 was estimated to be worth approximately 

$350 million, using a fixed lee type of contract. Facility construction and operation was to occur 

in a subsequent, planned Phase I, Part B-2. For Part B-2, the Contractor would only receive 

payment for waste actually processed and vitrified. This part of the effort was projected to cost 

DOE approximately $6.9 billicon (see Footnote 4, page 22). In total, Part B was expected to 

5 See, for example, The Energy Daily, May 22, 1998 and Inside Energy/with Federal Lands, May 25, 1998. 

6 U.S Department of Energy/Hanford Press Release, July 21, 1998.

NUREG-1747 8
require 5-8 years for design and construction activities, and another 5-10 years for the 

processing of the initial, 6-13 percent of the tank waste. 

At the time of this writing, Phase II planned to enlarge and utilize the Phase I facilities instead of 

constructing and using entirely new facilities. Hence, the proposed Phase I facilities would be 

relatively large. This diminished the distinction between the two phases and implied that the 

regulatory framework for Phase I would continue into Phase II. The Phase II activities extended 

the waste processing time frame by another 10-30 years, depending upon the facility 

capacities, waste retrieval activities, and difficulties encountered. Potentially, plant operation 

could continue until circa 2050. 

DOE anticipated the total value of the Part B work to be in the $7-10 billion range 7 (also see 

Main Reference 12). As part of the major design submittal (Main Reference 13), BNFL Inc. 

identified potential costs as high as $15 billion if the privatization route was continued. On May 

8, 2000, DOE decided to stop the privatization initiative for Hanford tank waste primarily for 

economic reasons, and elected to terminate the BNFL Inc. contract (Footnotes 5-7). Project 

management and quality assurance concerns were also mentioned as reasons for contract 

termination. The program name was also changed from Tank Waste Remediation System 

(TWRS) to River Protection Project (RPP). While the specific details are still evolving, DOE is 

using a transition contractor (CH2MHILL Hanford Group) to continue the design efforts while a 

competitive procurement is conducted for a new contractor for completion of the design and 

construction of the facility. A separate, additional contract would be released for facility 

operations. All of these new contracts would utilize M&O (management and operations) style 

contracts with cost-reimbursement and incentive clauses. 

1.4 DOE REGULATORY APPROACH 

DOE is a self-regulating agency on nuclear safety matters. DOE's goal in proceeding with the 

radiological, nuclear, and process safety regulation of TWRS-P contractors is to establish a 

regulatory environment that will permit privatization to occur on a timely, predictable, and stable 

basis with attention to safety consistent with that which would accrue from regulation by an 

external agency8 . This same policy states that DOE is patterning its regulation of TWRS-P 

contractors to be consistent with the NRC's regulatory approach. The Manager of the Richland 

Operations Office (DOE/RL) has the responsibility for safety of activities on the Hanford site. 

The policy and the related Memorandum of Agreement (MOA) 9 established a dedicated 

Regulatory Unit (RU) led by a Regulatory Official (RO) at the DOE Richland Operations Office 

with regulatory authority exclusive to the regulation of TWRS-P contractors. The RO would 

report directly to the Manager of DOE/RL at a level equivalent to the DOE Program Manager for 

TWRS. In implementation, the RU is to follow the five principles of good regulation as 

7 See Footnote 4 of this document and The Energy Daily, May 15, 2000 

8 Department of Energy (U.S.)(DOE). DOE/RL-96-25, Rev. 0, "Policy for Radiological, Nuclear, and Process Safety 

Regulation of TWRS Privatization Contractors." DOE: Richland, Washington. July 3, 1996. 

9 Department of Energy (U.S.)(DOE). DOE/RL-96-26, Rev. 0, "Memorandum of Agreement for the Execution of 

Radiological, Nuclear, and Process Safety Regulation of TWRS Privatization Contractors." DOE: Richland, 

Washington. July 3, 1996.

9 NUREG-1747
articulated by the NRC - independence, openness, efficiency, clarity, and reliability. The RU 

has four organizational elements: the Standards and Requirement Group, the Activities 

Authorization Group, the Verification and Confirmation Group, and the Senior Technical Team. 

The Manager of DOE/RL als.=o has a three-member, Senior Expert Review (SER) panel for 

periodic assessment of the RU and major issues. 

The basic concept of DOE's regulatory approach is that the contractor is responsible for 

achieving adequate safety, c'omplying with applicable laws and regulations, and conforming with 

top-level safety standards and principles stipulated by DOE1

". Consistent with applicable laws 

and regulations, the contracl:or is required to tailor the exercise of this responsibility to the 

specific hazards associated with its activities, and is encouraged to do this in a cost-effective 

manner that applies best commercial practices. TWRS-P contractors have the responsibility to 

identify and recommend to DOE the set of standards, regulations, and requirements necessary 

to ensure adequate safety. DOE's responsibility is to execute the regulatory process, including authorization of contractor actions and confirmation that the contractor activities are performed 

safely and within approved limits. 

The authority of the RU to regulate a TWRS-P contractor is derived from the terms of the 

TWRS-P contract ("regulate by the contract"). In addition to the regulatory concept, the 

following three radiological, nuclear, and process safety related documents are incorporated 

into the contract: 

1. "DOE Process for Radiological, Nuclear, and Process Safety Regulation of the TWRS-P 

Contractor," DOE/RL-96-0003. 

2. 'Top-Level Radiological, Nuclear, and Process Safety Standards and Principles for the 

TWRS-P Contractor," ICDOE!RL-0006. 

3. "Process for Establishing a Set of Radiological, Nuclear, and Process Safety Standards 

and Requirements for tI'e TWRS-P Contractor," DOE/RL-96-0004. 

The two, nonradiological doc'uments are: 

1. "Industrial Hygiene and S',afety Regulatory Plan," RL'REG-2000-04. 

2. "Regulatory Unit Position on Regulation of the Contractor's Industrial Hygiene and Safety 

Program," RL/REG-99-I 1. 

The RU has further explained the regulatory process as "bottoms up," starting with the 

contractor establishing the ,saý of standards needed to achieve safety, comply with applicable laws, achieve the DOE top-level standards and principles (DOE/RL-96-006), and follow an 

10 Department of Energy (U.S.)(DC'E). DOE/RL-96-0005, Rev. 1, "Concept of the DOE Regulatory Process for 

Radiological, Nuclear, and Process Safety for TWRS Privatization Contractors." DOE: Richland, Washington. July, 

1998.

NUREG-1747 10
integrated safety management process (ISMP, in DOE/RL-96-0004) 11. The regulator would 

subsequently approve the contractor's set of standards. A clear, central concept of ISM is that 

the contractor should tailor the design and safety requirements to the specific hazards of the 

activities and operations at a facility. DOE/RU policy endorses tailoring via the following 

process: 

1. Identify applicable requirements. 

2. Define the scope of the work or operations to be analyzed. 

3. Analyze the hazards. 

4. Propose, analyze, select, and implement controls. 

5. Perform the work or operations (does not apply at the design stage). 

6. Assess, feedback, and improve/modify (as appropriate). 

The Contractor conducted two ISMP cycles on the TWRS-P design prior to contract 

termination. 

DOE defined the following, minimum set of regulatory process elements: 

1. The top-level standards and principles. 

2. Standards identification (including the ISMP). 

3. DOE review and approval. 

4. Initial safety analysis and review. 

5. Construction authorization. 

6. Operating authorization. 

7. Regulatory oversight, including inspection of design, construction, or operating activities. 

8. Deactivation authorization. 

9. Independent oversight of TWRS regulation by DOE. 

10. Public information and involvement. 

(Note that the contract was terminated in the design phase prior to the CAR.) 

The RU has published guidance, policy, and position documents to assist its staff in these 

regulatory elements. Of these, the CAR review guidance (Main Reference 14) would be used 

in a manner similar to that of a standard review plan. This guidance was used in the review of 

the Firm Fixed Price submittal (Chapter 7.0, Main Reference 13). 

The RU also established an inspection program for the Phase 1B-1 activities. As of this writing, 

thirteen inspections were conducted over the 20-month period. 

DOE conducted two external assessments of the RU regulatory program and related activities. 

DOE, Office of Environment, Safety, and Health (EH), assessed the readiness of the RU in 

199812. The EH team reviewed TWRS-P related documents and interviewed personnel from 

11 Department of Energy (U.S.)(DOE). RL/REG-98-21, "Regulatory Unit Position on Implementing and Assuring 

Compliance with Integrated Safety Management." DOE: Richland, Washington. August 26, 1998. 

12 Department of Energy (U.S.)(DOE). DOE/EH-0569, "Assessment of Tank Waste Remediation System

Privatization Regulatory Unit Readiness." DOE: Richland, Washington. April 1998.

11 NUREG-1747
the RU, DOE/RL, DOE-Headcquarters, public, stakeholders, and the two Phase IA contractors. 

The review identified several weaknesses in the areas of management and organization, interfaces, staffing, technic'al standards and requirements, authorization process (contract consistency), document reviews, and inspection and enforcement. One weakness noted that 

the regulatory approach with two contractors resulted in two ISMP systems, different deliverable 

formats, and different submittal schedules. The RU subsequently developed a corrective 

action plan in response to these weaknesses' 3

. The RU only disagreed with the finding of one 

weakness; review schedule pressures. The EH team expressed concerns about the lack of 

specific provisions for delaying a review schedule when it cannot be supported by the RU. The 

EH team noted that RU directives and DOE documentation confirm the RU's responsibilities for meeting schedules, except when the contractor information is insufficient. The EH team further 

stated that review schedule pressures may cause reviewers to be less thorough, may compromise safety, and may give the public the impression that production takes precedence 

over safety and that the RU is not fully independent. In response, the RU noted that the 

revised contract adequately protects both the DOE and the Contractor from delays in the 

schedule caused by the exti-nsion of reviews. 

A second external assessmt,.mt of the RU and the regulatory process was conducted by the 

Senior Expert Reviewers (SER)14 . The SER panel reviewed documents, observed RU 

operations, and conducted interviews with four senior officials from DOE and BNFL Inc. (there 

were no interviews of NRC officials). The SER report considered the performance of the RU in planning and executing a first-of-a-kind regulatory concept exemplary. The report noted 

concerns about the level of detail, the Topical Meeting process, the need for submittal of any proposed changes in fundamental design safety features in the same time frame as the design finalization, and NRC reviews. On the latter, the SER panel concluded that the NRC reviews do 

not appear to have identified significant differences between the safety requirements invoked 

by the RU or the NRC, but rcted a significant difference in the regulatory system being developed by the NRC staff for possible application to TWRS-P (essentially 10 CFR Part 70 

and NUREG-1702), as compared to the DOE/RU regulatory system or process. 

The Office of River Protection (ORP) was established at the Hanford site in December 1998. 

As directed by Congress, DOE is using ORP to focus management responsibility and 

accountability. ORP has assumed the former RL/TWRS role of overseeing and directing TWRS, TWRS-P, and activit:!es related to the tank wastes. ORP reports directly to DOE 

Headquarters and not to DCIEYRL. ORP manages the TWRS-P contract and regularly interacts 

with the contractors on pro~gram issues. At the time of this writing (December 2000), the RU has become part of the ORF' and reports to ORP management. The RU has been renamed the 

Office of Safety Regulation (OSR), and the OSR is responsible for regulating the radiological, nuclear, and process safety of the facility being built to take waste currently stored in 

underground tanks at Hanford and process it to a glass form. From the new contract' 5 , the 

13 Department of Energy (U.S.)(CCiE). RLIREG-98-15, "Regulatory Unit (RU) Readiness Assessment (DOE/EH

0569) - RU Response and Corrective Actions." DOE: Richland, Washington. July 1998. 

14 Department of Energy (U.S.)(DOE). 00-RU-0005, "Report of an Assessment of the Regulatory Unit for the River 

Protection Project Privatization Contract," Senior Expert Reviewers, September 1999. 

15 Contract Number DE-AC27-01 RV1 4136, "Bechtel National Inc.: Design, Construction, and Commissioning of the 

Hanford Tank Waste Treatment and Immobilization Plant."

NUREG-1747 12
contractor is required to follow an integrated safety management (ISM) process, with approvals 

from DOE (presumably the OSR). The ORP organization also includes an "Office of Assistant 

Manager for Environmental, Safety, Health, and Quality" (AMSQ). The actual missions and 

working relationships of OSR and AMSQ within the ORP are not clear at this time. 

For the interim design period, the authority of the RU to regulate the RPP-WTP contractor was 

derived from a recent memorandum1 6 and stated in an RU position paper17 . 

1.5 NRC INVOLVEMENT VIA THE MEMORANDUM OF UNDERSTANDING 

DOE also considered the potential for external regulation, and entered into a Memorandum of 

Understanding (MOU) with the Nuclear Regulatory Commission (NRC) for cooperation and 

mutual support, with the possibility of transitioning to NRC regulation sometime in the future 

Attachment B). The MOU had two main purposes: 

1. The DOE to acquire the capability to implement a program of nuclear safety and 

safeguards regulation consistent with the NRC's regulatory approach. 

2. The NRC to acquire sufficient knowledge and understanding of the physical and 

operational situation at the Hanford waste tanks and the processes, technology, and 

hazards involved in Phase I activities to enable the NRC (a) to assist DOE in performing 

reviews in a manner consistent with NRC's regulatory approach and (b) to be prepared to 

develop an effective and efficient regulatory program for the licensing of DOE contractor

owned and contractor-operated facilities that will process waste at Hanford during Phase II. 

Prior to the termination of the privatization contract, DOE and the NRC were negotiating a 

revised MOU that recognized the reuse of the Phase I facilities for Phase II and maintained the 

NRC role in the near-term. However, the revised MOU implied a delay in regulatory transition, 

if it were to occur at all. In addition, the impact of the termination of the privatized contract and 

the use of new M&O-style contracts would have diminished or ultimately even eliminated NRC's 

participation, as it is likely that DOE will self-regulate the proposed facilities without transition to 

independent regulation by the NRC. Consequently, the NRC has decided not to participate in 

the near-term but has stated it is willing to discuss possible arrangements for NRC involvement 

in the TWRS project in the future, if there is again a need for NRC expertise and if the NRC 

sees benefit in such involvement, such as the potential transition to NRC regulatory authority. 

1.6 OVERVIEW OF THIS DOCUMENT 

Chapter 2.0 summarizes the NRC participation and activities. 

16 Huntoon, C.L., U.S. Department of Energy, memorandum to R.T. French, U.S. Department of Energy, 

"Maintaining Safety and Quality through Contract Transitions." May 23, 2000. 

17 Department of Energy (U.S.)(DOE). RL/REG-2000-20, Rev. 0 "Regulatory Unit Position on Important to Safety 

Work Authorization for the RPP-WTP Interim Design Period." DOE: Richland, Washington. July 3, 2000.

13 NUREG-1747
Chapter 3.0 discusses areas of design and safety that the NRC would typically review from a 

potential license application on TWRS-P, and provides an assessment of the current status 

based upon the BNFL Inc. documentation through June, 2000. 

Chapter 4.0 presents NRC observations and conclusions. 

Chapter 5.0 summarizes potential issues for transition of TWRS-type facilities to NRC 

regulation. 

The attachments provide information on the major technical issues from the perspective of the 

NRC staff, point paper summaries, the DOE/NRC MOU, potential issues for regulatory transition, and an index listing of correspondence between the NRC and DOE on the program.

NUREG-1747 14
2.0 NRC PARTICIPATION AND ACTIVITIES 

2.1 TANK WASTE REMEDIATION SYSTEMS SECTION 

In order to support the Memorandum of Understanding, the U.S. Nuclear Regulatory 

Commission (NRC), in October, 1996, established the Tank Waste Remediation System 

(TWRS) Section within the Special Projects Branch of the Fuel Cycle Safety and Safeguards 

Division, in the Office of Nuclear Materials Safety and Safeguards. This Section had the 

following specific tasks: 

1. Provide technical support to the U.S. Department of Energy (DOE) for activities related to 

the DOE TWRS privatization (TWRS-P) activities at Hanford. 

2. Develop sufficient knowledge and understanding of the physical and operational situation 

at the Hanford waste tanks and the processes, technologies, and hazards involved in 

Phase I activities to (a) assist DOE in performing reviews in a manner consistent with 

NRC's regulatory approach and (b) be prepared to develop an effective and efficient 

regulatory program for the licensing of DOE contractor-owned and contractor-operated 

facilities that will process Hanford wastes in Phase II. 

3. Conduct safety, safeguards, and environmental reviews concerning TWRS-P operations 

and processes. 

4. Review and comment on DOE regulations and guidance for the regulation of activities 

related to Phase I of the tank waste remediation activities at Hanford. 

5. At the request of DOE, participate as appropriate in the development of guidance for use 

by DOE, including guidance based upon industry standards (e.g., American Nuclear 

Society, American National Standards Institute, American Society of Mechanical 

Engineers). 

6. Review existing NRC regulations and guidance for their potential applicability to TWRS-P 

and its potential licensing in the future by NRC, including identification of potential 

modifications and additions. 

7. Identify potential differences between the DOE and NRC regulatory approaches as they 

apply to TWRS-P and notify DOE of these potential differences. 

To accomplish these tasks, the NRC recruited senior technical talent from both within and 

outside of the agency for the TWRS Section with experience in high level waste (HLW), HLW 

chemistry and processing, vitrification, Hanford, DOE, and regulatory activities. The following 

positions were filled: 

1. Section Leader 

2. Senior Onsite Technical Representative 

3. Senior Chemical Process Engineer 

4. Fire Protection Specialist

is NUREG-1747
5. Senior Nuclear Process Engineer 

6. Quality Assurance Spe,;ialist 

7. Structural/Construction Engineer 

8. Chemical Engineer 

9. Metallurgical Specialist/Engineer 

10. Nuclear Process Engineer 

11. Chemical/Materials Engineer 

12. Mechanical Engineer 

13. Environmental Engineer 

14. Health Physicist 

15. Licensing Assistant 

The staff had an average experience of approximately 20 years. Staff would rotate into 

Hanford for extended time periods in order to support the DOE-Regulatory Unit (RU), equivalent to a full time position onsite. In addition, a Senior Onsite Representative was located at 

Hanford, in the same office area as the RU. The onsite representative interacted with the RU 

on a daily basis, participating:l in meetings and presentations involving DOE and the contractors. 

The remaining staff were located in the Rockville, Maryland, offices of the NRC Headquarters. This staff visited the Hanford site many times for attending meetings and presentations by DOE 

and the contractors. An annual average of 57 trips were made in support of the RU and the 

TWRS-P program. Activities included review, analysis, and comment on contractor submittals, guidance documents, procedures, RU documents, etc. Usually, several members of the NRC 

Headquarters staff would rotate to Hanford for TWRS-related meetings for a week each month. 

Several major review activitie•s required members of the staff to remain in the Richland area for extended time periods. The staff at NRC Headquarters also communicated frequently with their DOE counterparts via phone, E-mail, and teleconferencing. NRC staff accompanied RU 

personnel on inspections. NR:!C staff reviewed NRC regulations and guidance related to TWRS-P. A standard review plan was prepared, issued, and finalized (see Section 2.3). NRC 

staff attended conferences aend training sessions related to TWRS technical areas, such as quality assurance and hazards analysis. Several NRC staff members also visited existing HLW 

processing facilities in Savannah River (Defensive Waste Processing Facility), West Valley (West Valley Demonstration Project), United Kingdom (Sellafield), and France (La Hague). 

2.2 SUPPORT ACTIVITIJES BY THE CENTER FOR NUCLEAR WASTE REGULATORY 

ANALYSES 

The Center for Nuclear Waste Regulatory Analyses (CNWRA) was established in 1987 as an 

NRC-sponsored, federally funded research and development center (FFRDC), in order to 

provide technical assistance and research in support of the NRC HLW program. Because of 

the similarities in the waste forms and processing requirements, the NRC decided to fund 

support work related to TWRý.E_; at the CNWRA. The CNWRA program had the following three 

main objectives: 

1. Provide technical assistance to the NRC for developing a regulatory framework and 

associated guidance that would be used to license a TWRS-P facility. 2. Assist the NRC in the review of contractor submittals. 

3. Develop an experience base for future activities in the program (e.g., construction and 

operations).

NUREG-1747 16
In support of these main objectives, the CNWRA effort had the following tasks: 

1. Familiarization, regulatory development, and safety reviews. 

2. Preoperational reviews of pilot processing facilities. 

3. Operational reviews of pilot facilities and reviews of safety bases. 

4. Revision of the regulatory framework. 

The CNWRA primarily operated under Task 1 in support of the NRC. The CNWRA produced 

analyses and reports on Hanford chemistry, risk/safety analyses, waste solidification, and 

process experience. This included participation in site visits to operating vitrification facilities at 

Savannah River, La Hague (France), and Sellafield (United Kingdom). The CNWRA analyzed 

consequence criteria, potential doses and risks, and issues, such as fires, explosion, and 

radiolysis. The CNWRA also assessed TWRS-P chemistry and processes, including several 

reports on the tank wastes, vitrification, separations, and a spreadsheet model of pretreatment 

processes. 

2.3 TWRS-P STANDARD REVIEW PLAN 

As part of its activities, the NRC has published the NUREG-1702, "Standard Review Plan for 

the Review of a License Application for the Tank Waste Remediation System Privatization 

(TWRS-P) Project" (Chapter 7.0, Main Reference 15). This provides NRC guidance for the 

review and evaluation of health, safety, and environmental protection in applications for licenses 

for remediation of radioactive tank waste at Hanford. The guidance is also applicable to the 

review and evaluation of proposed amendments and license renewal applications. Specific 

filing requirements for license applications and for issuance of such licenses are in 10 CFR Part 

70, "Domestic Licensing of Special Nuclear Material," as revised 18. Although 10 CFR Part 70, 

as revised, does not specifically include a TWRS-P facility in 10 CFR 70.60, "Applicability," the 

regulation specifies applicable facilities which include, "any other activity that the Commission 

determines could significantly affect public health and safety." 

The principal purpose of the Standard Review Plan (SRP) is to ensure the quality and 

uniformity of staff reviews and to present a well-defined base from which to evaluate proposed 

changes in the scope, level of detail, and acceptance criteria of reviews. The SRP also should 

be used as the basis for the review of requests by licensees for changes in their licenses. 

Thus, the SRP, at any point in time, can provide a basis for the review of proposed new or 

renewal applications, and amendments to existing licenses, as well as modifications to the SRP 

resulting from new NRC requirements and licensee initiatives. 

Another important purpose of the SRP is to make information about regulatory reviews widely 

available and to improve communication and understanding of the staff review process. 

Because the SRP describes the scope, level of detail, and acceptance criteria for reviewers, it 

can serve as regulatory guidance for applicants who need to determine what information should 

be presented in a license application. 

18 Nuclear Regulatory Commission (U.S.), Washington, D.C. "Domestic Licensing of Special Nuclear Material (10 

CFR Part 70)." Federal Register Vol. 64, No. 146. pp. 41338-41357. July 30, 1999.

17 NUREG-1747
The responsibility of the stalf in the review of a license application, renewal application, or license amendment for a TWRS-P facility is to determine that there is reasonable assurance 

that the facility can and will b9 operated in a manner that will not be inimical to the common 

defense and security, and will provide reasonable protection of the health and safety of workers 

and the public, and the environment. To carry out this responsibility, the staff evaluates 

information provided by an atpplicant and through independent assessments determines that 

the applicant has demonstriated a reasonable safety program that is in accordance with regulatory requirements. To facilitate carrying out this responsibility, the SRP clearly states and 

identifies those standards, criteria, and bases that the staff should use in reaching licensing 

decisions. 

The staff believes that a TWRS-P facility is an activity that could significantly affect public health 

and safety.. This belief is due to the presence of multi-kilogram quantities of special nuclear material (SNM) (slightly enricned uranium and plutonium), kilocurie quantities of transuranics 

(TRUs), megacurie quantities of fission products, and numerous chemical species in the 

wastes, and the need for a large plant with relatively high flow rates, large inventories, 

numerous process steps, anid energetic operations. Application of the specific technologies and equipment in a highly radioactive environment would be new and at a magnitude beyond 

any other applications and without significant pilot plant testing, thus increasing the unknowns 

and uncertainties. The melters would become the largest of their respective types in the world and have unique features. Therefore, if the NRC became the regulator, it would plan to invoke the requirements found in Subpart H of 10 CFR Part 70 for this type of facility. As such, 10 CFR Part 70, as revised, reqluires that an applicant submit a complete description of the safety 

program for the possession and use of SNM to show how compliance with the applicable requirements will be accomplished. The Safety Program Description must be sufficiently detailed to permit the staff to obtain reasonable assurance that the facility is designed and will be operated without undue risk to the health and safety of workers or the public. Prior to 

submission of the program Clescription, an applicant should have analyzed the facility in sufficient detail to conclude that it is designed and can be operated safely. The Safety Program Description is the principal document with which the applicant provides the information needed 

by staff to understand the basis for conclusion. When reviewed and approved by the staff, and incorporated in the NRC license by reference, the Safety Program Description, in its entirety and in its parts, is the safebt basis on which the license is issued and may not be changed 

except through the process defined in 10 CFR 70.72. 

The requirements in 10 CFR Part 70 specify, in general terms, the information to be supplied in 

a Safety Program Descriptioi. The specific information to be submitted by an applicant and evaluated by staff is identified in this SRP. Prospective applicants should study the topic areas 

treated in this document (generally, chapter headings) and the sections within each topic area, specifically the sections headed "Areas of Review" and "Acceptance Criteria." A license 

application should contain a Safety Program Description that addresses all the topics in the 

Table of Contents of this SRP, in the same order as presented in this document. 

In this SRP, information is provided to assist the licensing staff and the applicant in 

understanding the underlying objective of the regulatory requirements, the relationships among NRC requirements, the licensing process, the major guidance documents NRC staff has 

prepared for licensing facilities under 10 CFR Part 70, and the details of the staff review 

process set out in individual SRP sections. Analyses by the staff are intended to provide regulatory confirmation of reasonable assurance of safe design and operation. A staff

NUREG-1747 18
determination of reasonable assurance leads to a decision to issue or renew a license or to 

approve an amendment. In the case of a staff determination of inadequate description or 

commitments, the staff should inform the applicant of what is needed and the basis upon which 

the determination was made. 

The "Acceptance Criteria" delineated in this SRP are intended to communicate the underlying 

objectives but not to represent the only means of satisfying that objective. An applicant should 

tailor its safety program to the features of its particular facility. If approaches different from the 

SRP are chosen, the applicant should identify the portions of its application that differ from the 

design approaches and acceptance criteria of the SRP and evaluate how the proposed 

alternatives provide an acceptable method of complying with the Commission's regulations. 

The staff retains the responsibility to make an independent determination of the adequacy of 

what is proposed. 

The major topics addressed within the Safety Program Description of a facility license 

application are addressed in separate SRP chapters, as follows: 

0 Chapter 1: General Information 

* Chapter 2: Organization and Administration 

* Chapter 3: Integrated Safety Analysis 

* Chapter 4: Radiation Safety 

* Chapter 5: Nuclear Criticality Safety 

* Chapter 6: Chemical Safety 

* Chapter 7: Fire Protection 

* Chapter 8: Emergency Management 

* Chapter 9: Environmental Protection 

* Chapter 10: Decommissioning 

* Chapter 11: Management Measures 

* Chapter 12: Plant Systems 

The applicant's integrated safety analysis (ISA) is the central focus for the selection of design 

and operational safety measures and the management control systems that assure the 

availability and reliability of those measures. It is the ISA that provides a comprehensive 

evaluation and presentation, useful to both the applicant and the NRC, of the distribution of risk 

among the many activities ongoing at the TWRS-P facility. The NRC expects to be able to use 

the ISA summary to focus its resources on the dominant risks of facility design and operation 

and the safety controls and assurances necessary to ensure that those controls remain 

available and reliable. Accordingly, staff reviewers should conduct a coordinated review of the 

ISA summary and focus on the portions of the summary that are applicable to each of the 

technical areas treated in the chapters of the SRP. The acceptance criteria in each of the SRP 

chapters are the criteria that apply to the dominant risks of operation. The applicant has the 

opportunity to justify lesser criteria for those design and operational features that can be shown 

to represent lesser risk than the accident or failure sequences that pose the dominant risks. 

While recognizing the fundamental importance of the ISA to understanding the risk at a facility, 

certain SRP chapters are less dependent on ISA outcomes than others. The chapters 

concerning radiation safety, environmental protection, emergency management, and 

decommissioning, for example, contain acceptance criteria that are set primarily by existing 

regulations and have not been affected by issuing the revision to 10 CFR Part 70. Finally, for

19 NUREG-1747
new facilities (that have not already been designed, built, licensed, and operated), certain 

baseline design criteria have been specified in 10 CFR 70.64, "Requirements for New Facilities 

or New Processes at Existing Facilities." These criteria identify safety considerations that an 

applicant must address in its, facility design. The ISA for the complete facility design may indicate when reduced levels of assurance may be acceptable. 

Each chapter in the SRP includes sections, which are described below in general terms. A more 

detailed description of the application of the baseline design criteria is given in the discussion of 

Section 4, "Acceptance Criteria," below. 

Section 1., PURPOSE OF REVIEW: This section is a brief statement of the purpose for and objectives of reviewing the subject areas. It emphasizes the staff's evaluation of the ways the 

applicant can achieve identified performance objectives and ensures through the review that 

the applicant has used a multi-disciplinary, systems-oriented approach to establishing designs, controls, and procedures within individual technical areas. 

Section 2., RESPONSIBILITY FOR REVIEW: This section identifies the organization and 

individuals by function, withir NRC, responsible for evaluating the subject or functional area 

covered by the SRP. If reviewers with expertise in other areas are to participate in the 

evaluation, they are identified by function. In general, the Licensing Project Manager has 

responsibility for the total review product, a safety evaluation report for an application. However, an identified technical specialist should have primary responsibility for a particular review topic, usually an SRP chapter. One or more specialists may have supporting responsibility. In most situations the review is performed by a team of specialist reviewers 

including the lead reviewer for the ISA and the project manager. Although they individually perform their review tasks, the reviews are extensively coordinated and integrated to ensure 

consistency in approach and to ensure risk-informed reviews. The project manager oversees 

and directs the coordination of the reviewers. The reviewers' immediate line management has 

the responsibility to ensure that an adequate review is performed by qualified reviewers. 

Section 3., AREAS OF REVIEW: This section describes the topics, functions, systems, structures, equipment, and components, analyses, data, or other information that should be reviewed as part of that particular subject area of the license application. Because the section 

identifies information to be reviewed in evaluating the adequacy of the application, it identifies 

the acceptable content of an applicant's submittal in the areas discussed. The areas of review 

identified in this section obviate the need for a separate Standard Format and Content Guide. 

The topics identified in this section also set the content of the next two sections of the SRP. 

Both Section 4, "Acceptance Criteria," and Section 5, "Review Procedures," should address, in 

the same order, the topics set forth in this section as areas to be reviewed. This section also identifies the information needed or the review expected from other NRC individuals to permit the individual charged with primary review responsibility to complete the review. 

Section 4., ACCEPTANCE CI-:,ITERIA: This section contains a statement of the applicable NRC criteria based on regula~i:ory requirements, and the bases for determining the acceptability of the applicant's commitments relative to the design, programs, or functions within the scope of 

the particular SRP section. Technical bases consist of specific criteria such as NRC regulations, regulatory guides, NUREG reports, industry codes and standards, and branch 

technical positions. To the extent practicable, the acceptance criteria identify, as objectively or

NUREG-1747 20
quantitatively as is feasible, specific criteria and other technical bases that are to be satisfied. 

The acceptance criteria (including branch technical positions or other information) present 

positions and approaches that are acceptable to the staff. They are not considered the only 

acceptable positions or approaches. Others may be proposed by an applicant. 

It is NRC's intent that the SRP present acceptance criteria for each technical function area 

(e.g., nuclear criticality safety, fire safety, radiation safety), and for the management measures 

(e.g., quality assurance, maintenance, audits and assessments), that allow an applicant to 

provide a level of protection commensurate with the accident risk inherent in the process 

activities proposed. For example, at process stations (or for an entire process or sub-process) 

for which the inherent risk to workers, the public, or the environment is demonstrably small, the 

applicant needs to provide only those design and operating controls which assure that small 

risk. The key element in the regulatory transaction involving presentation by an applicant, and 

review and approval by the NRC, is an adequate demonstration of acceptable control of risk by 

the applicant, which then supports a competent and informed review by NRC staff. The starting 

point for the applicant's demonstration of acceptable control of risk is the ISA. 

The applicant's ISA summary (described in and reviewed under Chapter 3 of this SRP) is the 

primary supporting rationale for the safety level of design and operational features. There are, 

however, design and operational features and management controls that may be required 

independent of the ISA results presented by an applicant. This is to meet the requirements of 

10 CFR 70.64 for new facilities or new processes at existing facilities, or, for all facilities, other 

NRC requirements such as 10 CFR Parts 20 and 51. The level of detail presented in the ISA 

summary and in other parts of the application represents the safety basis committed to by the 

applicant, and it is that basis that is subject to the provisions of 10 CFR Part 70, as revised, 

regarding changes that a licensee may make to the facility without prior NRC approval. NRC 

should find an application acceptable if an applicant commits to the design features and 

management measures defined by the acceptance criteria within this SRP. The criteria in this 

SRP represent the design features or management measures that support an NRC finding of 

reasonable assurance of adequate protection, independent of any iSA findings or conclusions 

that could lead to NRC approval of reduced levels of assurance for certain design features or 

management measures where the associated risk does not warrant the same high level of 

assurance. 

An applicant for license renewal or an amendment for an existing facility responding to the 

requirements of 10 CFR Part 70, may propose structures, systems, and components (SSC) or 

management measures that meet less stringent acceptance criteria than described in the SRP 

based on supporting analyses from the applicant's ISA. The ISA may be used to justify a 

reduced level of assurance for particular items relied on for safety, that are associated with 

lesser risk accident sequences, as defined by the applicant's analysis of likelihood and 

consequences pursuant to 10 CFR Part 70, as revised. The SRP criteria shown in this SRP 

apply to those SSC and management measures that are involved in the higher risk accident 

sequences as defined in Part 70, as revised. 

For proposed new facilities or amendments for new processes proposed at existing facilities, 

the acceptance criteria described in the SRP apply for design purposes and should be 

addressed in the applicant's licensing submittal for all SSC and management measures and 

that section's requirement to comply with the baseline design criteria (BDCs) of Part 70, as 

revised. The BDCs are consistent with risk-informed regulation, in that, for new processes or

21 NUREG-1747
new facilities, NRC recognizes that good engineering practice dictates certain minimum requirements be applied as design and safety considerations, generally independent of the risk based information ultimately obtained through the ISA. However, the applicant may use the ISA summary to justify reduced criteria for some SSC and management measures consistent with an ISA summary for a f:acility final design. Proposed reductions in the level of assurance should be considered by the NRC staff and, if accepted, should also constitute compliance with 

the BDCs. 

Section 5., REVIEW PROCE--DURES: This section describes how the review should be performed. It describes procedures that the reviewer should follow to achieve an acceptable scope and depth of review and to obtain reasonable assurance that the applicant has provided appropriate commitments to ensure that it will operate the facility safely. This includes identifying licensee commitments to verify and could include directing the reviewer to coordinate with others having review responsibilities for other portions of the application than that assigned to the reviewer. This section should provide whatever procedural guidance is necessary to evaluate the applicant's level of achievement of the acceptance criteria. 

Section 6.. EVALUATION FINDINGS: This section presents the type of positive conclusion that is sought for the particular re--ýview area to support a decision to grant a license or amendment. The review must be adequale to permit the reviewer to support this conclusion. For each section, a conclusion of this type should be included in the staff's Safety Evaluation Report (SER) in which the staff publishes the results of its review. The SER should also contain a description of the review, including aspects of the review that received special emphasis; matters that were modified by the applicant during the review; matters that require additional information or will be resolved in the future; aspects where the plant's design or the applicant's proposals deviate from the crteria in the SRP; and the bases for any deviations from the SRP 

or proposed exemptions frorn the regulations. Staff reviews may be documented in the form of draft SERs that identify open issues requiring resolution before the staff can make a positive finding in favor of the license issuance or amendment. 

Section 7., REFERENCES: This section lists references that should be consulted in the review 

process. However, they may not always be relevant to the review, depending on the action and 

approaches proposed by the applicant. 

2.4 NRC PARTICIPATION IN PHASE IA 

NRC staff have participated in various reviews of the TWRS-P program. Under Phase IA, the staff reviewed the following major documents from each of the two contractor teams (BNFL Inc. and Lockheed Martin Advanced Environmental Systems (LMAES); see Chapter 7.0 for main 

reference for citations): 

1. Hazards Analysis Report (HAR) - 1997. 

2. Standards Requirements Document (SRD) - 1997. 

3. Integrated Safety Management Plan (ISMP) - 1997. 

4. Initial Safety Analysis FReport (ISAR) - 1998.

NUREG-1747 22
Numerous open items or items requiring additional information and clarification were identified 

by the NRC's review of these documents. Focusing on BNFL Inc. (because of its selection by 

DOE for continuing into Phase IB - see Section 1.3), a total of 409 comments were generated 

by the NRC review of these documents, and were subsequently identified and sent to DOE in 

interagency letters. DOE subsequently forwarded most of these comments to the Contractor 

for informational purposes. The NRC staff developed and maintained databases to track the 

issues from the reviews of the SRD and ISAR. The NRC staff identified some 245 comments 

as being open from the perspective of the NRC as a potential regulator of the facility. 

From the point of view of the NRC staff, the most significant issues were found to be the 

following19 : 

1. Incompleteness of the description and documentation. 

2. Key unit operational and process information not included. 

3. Inadequate configuration management and documentation organization. 

4. Exclusion of safety analyses for the facility workers. 

5. Inadequate design class description and SSC categorization. 

6. Limited analysis of non-process accidents and external events. 

7. Lack of conservatism, particularly at this early stage of design. 

The DOE review reached similar conclusions. DOE divided its comments of the ISAR into the 

following four main categories in March 1998 (see Chapter 7.0, Main Reference 10): 

1. Adequate Safety Basis Not Demonstrated: There were 25 issues relating to the adequacy 

of the proposed facility's safety basis. Resolution was to be required during the first 

6 months of Phase lB-1. 

2. Inadequate Classification of SSC's: There were seven issues to ensure that the ISMP 

described in the Contract and in the BNFL Inc. ISMP (plan) were implemented. Resolution 

was to be required during the first 6 months of Phase lB-1. 

3. Incomplete or Conflicting Elements of the Authorization Basis: There were 30 issues that 

showed incomplete, inadequate assessments and attention to detail. Resolution of these 

items would be required prior to the submittal of the Construction Authorization Request 

(CAR). 

4. Incomplete Design or Operational Information: There were 28 issues related to the limited 

information available in the Phase IA (conceptual) design stage. Resolution of these items 

would be required prior to the submittal of the CAR. 

19 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, 

October 17, 1997.

23 NUREG-1747
2.5 NRCO PARTICIPATION; IN PHASE IB-1

DOE selected BNFL Inc. as the sole contractor for Phase IB-1 due to concerns about the status 

of technology development in the LMAES proposal. Phase lB-1 focused on addressing the 

main issues from Phase IA aid extending the level of design from the preliminary conceptual design level of 3-5 percent to around 25-30 percent. DOE identified the HAR, SRD, and the 

ISAR submitted in Phase IA as the authorization basis for the continuing design efforts. The 

NRC staff participated in major reviews of the following (see Chapter 7.0, main reference 

section for report citations): 

1. The Design Safety Features (DSF) submittal (four volumes, Spring, 1999). 

2. Topical Meetings and their associated reports (essentially monthly). 

3. Other Technical Reportis (e.g., the RPP - Radiation Protection Plan, QAPIP - Quality 

Assurance Program Im olementation Plan). 

4. Periodic Design Reviews. 

5. April 24, 2000, submittal (Firm Fixed Price submittal). 

Most of the NRC staff commrients on these Part B-1 documents mirror the ISAR comments, and 

are discussed further in Sections 3.1-3.10 and Appendix A. A large fraction of these comments 

center around the lack of detail and supporting information on safety-related issues, potentially 

uneven amounts of redundant equipment and backup systems, and inadequate maintenance of 

the design/authorization basis. From the perspective of the NRC staff and based upon subsequent NRC reviews of the additional information and documents generated so far in 

Phase lB-1, a significant number of the original, specific comments (about 100) tracked in the 

databases remained open. A list of potential critical issues was forwarded to the RU2

". One 

large database was subsequently developed for tracking the issues. For this report, these 

issues have been updated and condensed into a form that are largely design independent and 

have been placed in Attachment A. DOE opted to discuss several of these issues with the 

Contractor via the "Topical Meeting" process (see Section 3.1). Most of these meetings had at least one NRC representative in attendance. DOE subsequently decided to close many of 

these issues by deferring then to the future review of the CAR/Preliminary Safety Analysis Report instead of keeping them open in an Action Plan or SER. DOE/RU anticipates a long list 

of issues from the CAR review activities 21. Thus, as of this writing, potentially significant and 

safety-related issues remain open. 

Review areas of major involvement have included the following: 

* Organization and general information. 

20 Leach, M.N., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "Potential 

Critical Technical Issues For Construction Authorization Resolution," January 21, 2000. 

21 Hoadley, D.A., U.S. Nuclear Recgulatory Commission, Note to File, "Minutes of March 16, 2000 Teleconference 

with DOE/RU," March 21, 2000.

NUREG-1747 24
* Seismic and structural considerations. 

* Hazards and safety analyses. 

* Radiation safety and dose assessment methodology. 

* Criticality safety. 

* Process and chemical safety. 

* Fire protection. 

* Explosion protection. 

* Environmental protection. 

* Quality assurance (QA) and management measures. 

These are discussed in more detail in Chapter 3.0. A common concern of the reviews of the 

NRC staff is the potential lack of conservatism and margin in many of these areas. Thus, if the 

facility is designed right at a limit in a particular area, it may be significantly restricted for future 

changes and modifications such as increased capacities or unforeseen changes in the wastes 

and chemistries. In addition, the design and safety limits are derived from the ISMP. For 

TWRS-P, ISMP has invoked a circular logic that can magnify the impact of changes, particularly 

when a scenario consequence result slightly exceeds a limit. ISMP is an iterative process 

which includes reassessing assumptions (e.g., of radionuclide concentrations) that can reduce 

a hazard below a limit into a bin with less reliability and safety requirements. Once this occurs, 

the scenario represents less risk with fewer safety requirements, and, since ISMP focuses on 

events with higher risk, this scenario and the underlying assumptions may be subject to less 

scrutiny. 

The NRC staff participated with the RU in inspections of Contractor activities and programs. 

The NRC staff members provided input to the RU inspection teams on the following programs: 

Personnel Training and Qualification, Design Process, as low as reasonably achievable/RPP, 

QA, and Employee Concerns. These inspections identified a number of design, QA, and 

project management issues requiring BNFL Inc. actions, including procedure compliance, 

design basis documentation, corrective actions, and document control, which are further 

discussed in the relevant parts of Chapter 3.0. 

The NRC staff has also provided input on guidance and regulatory documents to DOE and the 

RU, and commented on issues that have arisen. Appendix C contains summaries of Point 

Papers generated by the NRC. Appendix E provides a listing of correspondence with the DOE. 

Several major guidance and regulatory items reviewed by the NRC staff have included the 

Limited Construction Authorization Request (LCAR) Review Guidance, the CAR Review 

Guidance, the co-located worker issue, and risk and regulatory comparisons. The latter two 

items are discussed further in Chapter 4.0. 

The NRC has conducted independent assessments and analyses, including the following: 

1. The Standard Review Plan (SRP - NUREG-1702) was developed, issued as a draft for 

public comment, and issued in final form (see Section 2.3). 

2. Point/Position Papers on QA, process safety, fire, and materials aspects (Appendix C). 

3. Information papers on radiation protection. 

4. CNWRA analyses and reports include:

25 NUREG-1747
A. "Survey of Solidific'ation Process Technologies," April 1998.

B. "Process Hazards, and Safety Issues for TWRS-P." 

C. Vol. 1 -Low activity waste Feed Makeup, Solidification, and Offgas Treatment, 

July 1999. 

D. Vol. 2 - Auxiliary Support Systems and Process Control Systems, August 1999. 

E. HLW Chemistry Manual. 

F. Chemistry of TWRS Waste Pretreatment and Technology Report. 

G. PRETREAT Spreadsheet Model for TWRS Pretreatment Flowsheet. 

H. Glass Melt Chernistry and Product Qualification Report (in progress). 

I. Chemical Simulation of TWRS Pretreatment Process (in progress). 

J. "Classification of Process Systems Used in the Tank Waste Remediation System

Privatization Design," June 2000. 

The TWRS-P facility represents a radiochemical facility with a relatively large radionuclide 

inventory. Table 3 compares the potential radionuclide inventories at TWRS-P locations 

(calculated from Tables 1 arco 2) with selected radionuclide quantities at a nuclear power plant (calculated from the Radiological Characteristics Database22 ); the TWRS facility is likely to 

handle comparable quantities of radioactive cesium, strontium, and technetium in significantly 

more mobile physical and chemical forms (e.g., as nitrates and aqueous solutions), as 

compared to ceramic oxide fuels in power reactors. In addition, while a reactor has more 

energy for potentially energetic scenarios during operations, including scenarios with delays of 

hours and days before the radionuclide release occurs, the TWRS-P facility has more stored 

chemical energy for prompt potential events directly involving the radionuclides in their mobile 

forms. Consequently, the Ti'WRS/Waste Treatment Plant may have some requirements that 

are more similar to reactor facilities than to commercial fuel fabrication facilities. 

22 U.S. Department of Energy, (U.33.) (DOE). "Radiological Characteristics Database." DOE: Richland, 

Washington. 1995.

NUREG-1747 26
Table 3: Comparison of Curie Quantities between TWRS-P and a Commercial Reactor 

1,000 MWe Nominal Reactor (PWR, in cu1. es 

Bounding for TW RS-P Facility 

Radio- (cuies) 30,000 MWD1MTIHM 60,000 MWD2MTIHM 

nuclide 6 0. 

LAW HLW Cs Cs Resin OST Core, SNF Dry Core, SNF DRY 

10OKgaJ 10OKgal 1IKgal 100O liters coled cooled cooled cooled 

_ _ _ (Env. B) __ _ _ __ _ _ _ _ _ _ _ __ _ _ __ _ __ _ _ _ _ _ _ _ _ _ 

Cs-137 8.6E6 1.1E6 1.32E6 72,000 3E5 9.2E6 1E6 1.8E7 1.9E6 

(2.3E6) (7.57E06) (no (no (no 

change) change) change) _____ 

Tc-99 1,000 1,700 (0) (0) 0 1,200 140 2,200 260 

(727) (11,400) _ _ 1__ ____ _ _ _ _ _ _ _ 

Sr-90 6,300 1.2E6 (0) (0) 0 6.6E6 7.2E5 1 .2E7 1 .3E6 

(4,500) (7.57E06) ____ _____ ___ ____ 

TRU 8,700; 17,200 (0) (0) 0 1.3E7 1.2E6 2E7 1.8E6 

70 from (87,000) 

solution 

(8,700; 

49 from 

- ! solution) 

Note: Reactor core nominally contains 100 MTIHM and SNF cask nominally contains 

12 MTIHM. 

PWR values calculated using the Radiological Characteristics Database from 

Footnote 22. 

TWRS-P values calculated from Tables 1 and 2 for the original contract, with 5 

percent and 10 percent suspended solids for the LAW and HLW respectively. The 

values in parentheses are calculated from the new contract, using 2 percent and 20 

percent as the suspended solids concentrations for LAW and HLW respectively. The 

HLW has not been washed. Non-TRU, LAW values do not include the solids 

contribution; for the new contract, inclusion of the solids contribution would 

respectively add 7.6E5, 1.1 E3, and 7.6E5 curies to the cesium, technetium, and 

strontium values. Recent discussions have not included CST columns.

27 NUREG-1747
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NUREG-1747 28
3.0 NRC AREAS OF REVIEW 

3.1 ORGANIZATION AND GENERAL INFORMATION 

3.1.1 TANK WASTE REMEDIATION SYSTEM-PRIVATIZATION ORGANIZATION 

The BNFL Inc. team was approximately 550 people at its peak, with some additional support 

from outside contractors and consultants. Approximately 50 key personnel from BEI (BNFL 

Engineering Incorporated, United Kingdom) were in lead positions. 

The organization's management functions were adopted from the British Nuclear Fuels Limited 

parent company. These functions were characterized as the 'TWRS-P Business Model." 

A total of eight key business processes were identified: 

Business Process' Responsibilty~ 

Management and Administration Project Manager 

Corporate policies General Manager 

Strategies Quality Assurance Manager 

Quality Program 

Resources Human Resources Manager 

Human Resources 

Training 

Customer Service General Manager 

DOE Interface and Controls DOE Client Interface 

Client (DOE) Commitment Tracking 

Change Control 

Stakeholder Interface 

Procurement Procurement Manager 

Local Purchase 

Plant and Equipment 

Service Contracts 

Main Subcontracts (R&D) 

Finance and Accounts Financial Manager 

Control of Funds 

Financial Audit 

Financial Risk 

Financial Delegation 

Project Management Project Manager 

Schedules 

Costs 

Value Engineering Options 

Project Meeting Structure 

Project Risk Management 

Design and Engineering Engineering Manager 

Engineering Design Process Environmental Safety and Health Manager 

Environmental Safety and Health Procedures 

and Processes 

Technical Risk Management 

Construction 

Operations Operations Manager 

(To be developed) II

29 NUREG-1747
Each business process performed in accordance with established methods. The hierarchy of 

instruction was as follows: general instructions were included in Procedures, with details in 

Codes of Practice. Other direction was provided by Design Guides, sometimes called Desktop Guides. Procedures, Codes of Practice, and Design Guides were not submitted for U.S. 

Nuclear Regulatory Commission (NRC) review but, as part of inspection accompaniments with 

the RU, some Contractor documents and procedures were compared with activities for 

consistency. Some disconnects were observed, apparently due to the relative fast schedule of 

the program and the rapid aggregation of the project team from multiple locations. 

Performance of some management functions ("business processes") such as Corporate Policy & Strategy, Finance & Accounts, DOE Interface, Schedule & Costs, and Operations were 

transparent to NRC technical reviewers. 

Performance of other managlement functions was visible to NRC technical reviewers because 

these departments had active roles in many of the design and safety submittals that were 

reviewed (Safety Requirements Document [SRD], Hazards Analysis Report [HAR], Integrated 

Safety Management Plan [ISMP], etc.) and presentations (Design Reviews and Topical Meetings) that were attended. However, it was not clear that environmental, safety, and health 

(ES&H) and quality assurance (QA) functions were sufficiently independent and had the 

authority to ensure that their issues were adequately addressed. Specific examples of these 

points are discussed in Sections 3.1.3 and 3.10. 

3.1.2 MEETINGS AND REPORTS ON SAFETY AND REGULATORY ACTIVITIES 

The NRC was not fully involved with and did not comment on all meetings and reports. During Phase 1, Part A, there were lhree main deliverable packages for addressing safety and 

regulatory concerns; these were the Standards Application Package (SAP) ( consisting of the 

SRD and the HAR), the ISMP, and the Initial Safety Analysis Report (ISAR). No revisions were 

made to these documents and a final version reconciled to the U.S. Department of Energy (DOE) and NRC comments was not issued in Part A. The SRD was revised twice in late 1998 

and the HAR was to be revised in the Fall of 2000 (both in Part B-1 of Phase I). In addition, the 

organization of the document:s and references was difficult to follow; for example, the ISAR 

referred to a Technical Repcr: which in turn referenced three binders of poorly organized, handwritten material. The Pairt B-1 Contract only identified two main deliverables for regulatory and safety purposes: (1) the Design Safety Features (DSF), also called the six month submittal 

and (2) the Construction Authorization Request (CAR), which would include the Preliminary 

Safety Analysis Report (PSAFI), at 26 months into the Contract. The NRC and the 

DOE/Regulatory Unit (RU) also looked at the Firm Fixed Price (FFP), 20-month submittal for 

informational purposes about the design and approach. The FFP was not intended to address 

safety and regulatory concerns. The DSF submittal only represented examples that were not 

part of the design process. Thus, the FFP submittal represented the first formal update of the 

design and safety features sqnce the ISAR, a period of 2 years. The FFP submittal itself was 

really a collection of reports. Integration between these multiple reports was not a clear 

process. Consequently, revisions addressing comments and concerns were not produced in a 

timely manner, and there were significant time periods between these documents during which 

there was no contractually required safety information. 

As a result of the limited number of safety deliverables in Part B, the Design Review process 

was instituted by the DOE/RU as part of their approach towards project oversight. The

NUREG-1747 30
objectives of the design review process were to (a) verify fulfillment of the TWRS-P contract 

requirement to review the evolving design, (b) ensure that safety-related design aspects were 

integrated between Contractor engineering disciplines, (c) develop a clear understanding of 

evolving safety-related design aspects, and (d) provide a reasonable expectation that the CAR 

would be acceptable. Design reviews were performed primarily as a "status meeting" as the 

discipline engineers exchanged information, and NRC reviewers attended as many as possible. 

DOE and NRC staff members participated in design reviews as observers and not as reviewers. 

The quality of the NRC-observed design reviews was variable. On the positive side, minutes 

and action items were kept and assembled for later review in the library. However, on the 

negative side, greater rigor, meeting focus, and conduct would seem to be needed. Documents 

were usually provided to the meeting attendees at the beginning of the meeting, which did not 

allow for adequate review time. The documents/handouts usually consisted of just the 

overheads for the presentation, without a formal report with revision numbers. No revision 

changes and written notes/rationale were apparent in the presentations, although these were 

sometimes mentioned verbally. There appeared to be an emphasis for avoiding action items, 

using terms such as "design refinemenf' for items that would seem more like actual action 

items. It was not clear how action items were tracked, particularly between interrelated design 

activities and reviews. Although not required, there was little apparent involvement of personnel 

from the Contractor's safety organizations. The design teams frequently appeared to be 

working separately from the safety reviewers, without consideration of the design/authorization 

basis, safety, and the SRD. In some meetings, over half of the questions are asked by 

DOE/RU and NRC personnel. Finally, design reviews and documentation were not formally 

reviewed and commented upon by either DOE or the NRC. Only general feedback was 

provided by DOE to the Contractor. 

Topical Meetings were instituted by the RU in Phase IB, in an effort to keep informed of 

progress and potentially bring closure to the many issues that had been raised during Phase IA. 

The Topical Meetings were held every month. The results of a Topical Report on the issue 

were generally submitted in advance of the meeting. After one of the early meetings, an 

agreement was made to hold a pre-topical meeting approximately 2 weeks prior to the Topical 

Meeting to ensure that the level of detail in the presentation was satisfactory. An effort was 

made to have at least two NRC reviewers at each topical meeting. While there were many 

letter-type Topical and other reports generated for the Topical Meetings, it was not clear how 

these documents inter-related with each other and would be incorporated into safety analysis 

reports. A clear route to closure was not apparent. The NRC communicated to the DOE-RU its 

concerns regarding closure of issues 23. In particular, the NRC noted the RU was focusing on 

the Contractor demonstrating an understanding of the standards identification process (see 

Section 1.4) and the Contractor's implementation of this process, while the NRC concentrated 

on determining whether or not the Contractor's approach to safety incorporated appropriate 

levels of conservatism as well as whether the design approaches could later prove to pose 

potentially unacceptable levels of risk. The NRC further commented that the RU and the 

Contractor would benefit from closing issues in the near term and well before the CAR; 

otherwise, the deferrals may increase the time and costs associated with the review of the 

23 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, June 25, 

1999.

31 NUREG-1747
CAR. The RU identified the- following, thirteen significant, unresolved issues from the Topical 

Meetings 24 : 

1. Criticality control. 

2. Explosion hazards (hycdrogen). 

3. Fire protection. 

4. Design safety features (unmitigated definition, source terms, uncertainties, etc.). 

5. Dose assessment. 

6. Emergency response plan. 

7. Seismic analysis and criieria. 

8. Technology development and plans. 

9. Explosion hazards (non-hydrogen). 

10. Iodine-129 evaluation. 

11. Cesium storage tank cooling. 

12. Seismic PRA dose consequence. 

13. ISM cycle 1. 

At the request of the DOE/Ofrfice of River Protection (ORP), the RU reviewed the FFP submittal 

with assistance from NRC staff25 . While noting the FFP submittal was not intended as a safety deliverable and that some additional design information exists outside of the FFP document, the 

review determined that the facility and process design information fell far short of the 

information required for the CAR. Significant additional design information would be needed, including further developmenl: of the design and information on safety equipment. The 

following, short list of items were presented: 

1. Limited identification of items relied on for safety (IROFS). 

2. Control logic and diagrams not provided. 

3. System descriptions were at a summary level, and codes and standards were not identified 

for most structures, systems, and components, (SSCs). 

4. Materials of construction were not identified for most SSCs. 

5. Process information focused on normal operations-information for startup, shutdown, 

off normal, and fault conditions was not provided. 

6. Operating parameters (t:emperature, flow rates, pressures, etc.) were not routinely included 

for routine operations arid not mentioned for startup, shutdown, off normal, and fault 

conditions. 

7. Information was rarely pr'ovided for system pressure, pressure drops, instantaneous flow 

rates, and specific gravity (this information would be needed to assess tanks and the 

adequacy of line/pump sizes, etc.). 

24 Gibbs, D.C., U.S. Department o'a E--nergy, letter to M. Bullock, BNFL Inc., March 24, 2000. 

25 Gibbs, D.C., U.S. Department of Energy, letter to R. French, U.S. Department of Energy, May 25, 2000.

NUREG-1747 32
8. The description focuses on operability with little attention to safety issues. 

9. Information regarding analysis and the validation/verification of software for structural and 

seismic considerations was not provided. 

10. Important-to-safety (ITS) areas of the design are not complete or do not have the detail to 

support safety analyses (CAR/PSAR). 

11. The levels of conservatism and the design margins are not mentioned for operations or 

safety parameters. 

The RU subsequently provided this information to the Contractor and included a more detailed 

report by review area26. 

3.1.3 INSPECTIONS 

NRC reviewers also participated in several inspections which were conducted by the DOE/RU: 

these included inspections of BNFL Inc.'s Employee Concerns Program, Training and 

Qualification Program (two inspections), Quality Assurance Program (two inspections), and 

Design Process. 

One of the earliest inspections was of the Employee Concerns Program. Inspectors found the 

program to be in place, but no formal concerns had been voiced. 

An inspection of the Contractor's Training and Qualification Program was conducted early in the 

project. The Contractor was hiring staff at the time of the inspection. Training and Qualification 

procedures were only recently issued, but appeared to be functioning. It was determined that 

the Contractor's staff was qualified for their positions and responsibilities. Two important 

program weaknesses were identified: (1) the positions of Training Manager and Training 

Specialist were not filled at the time of the inspection; and (2) the Contractor's program for 

position-specific training was not fully developed. The second inspection of the Training and 

Qualification program confirmed that deficiencies in management and position-specific training 

still remained. 

An inspection of the design process and authorization basis was initially scheduled in the Spring 

of 1999, then postponed by the RU because of concerns that the Contractor would do poorly as 

the authorization basis was not being followed nor maintained. Subsequently, the postponed 

inspection was canceled - again, over RU concerns that the Contractor would not pass the 

inspection. The Contractor design process was inspected in January, 2000. Three (3) areas of 

Findings were identified during the inspection: (1) examples where the Contractor had not 

followed procedures; (2) procedures did not address or consider testability or inspectability in 

the design; (3) the Quality Assurance organization was not effectively reviewing the design 

program. Limited QA overview of the design process and continued instances of procedural 

noncompliance were causes for concern. These and continuing concerns about the 

management of the authorization and design basis resulted in a Corrective Action Notice (CAN) 

and conference in March, 2000. A highly interactive QA organization that self-identified 

26 Gibbs, D.C., U.S. Department of Energy, letter to P. Strawbridge, BNFL Inc., June 2, 2000.

33 NUREG-1747
problems, such as the procedure noncompliance issues identified in this inspection, coupled with a robust corrective acticn program, were two elements required to assure the development of a safe design and mainta3in the authorization basis. 

DOE/RU and the Contractor recognized deficiencies in QA, and conducted several self

assessments, including a Rcot Cause Analysis. This was viewed as a weakness in Project/Management, and not simply a QA problem. Some of the deficiencies include a failure 

to follow procedures, lack o1: knowledge of requirements, failure to implement inspection programs, inadequate docu mentation of design activities, and deficient design or contract technical and standards bases. The failures noted were program/project implementation (management) problems, which would be detected during inspections. 

DOE/RU issued a CAN27 on March 14, 2000, and held a public meeting in March 2000. Contractual and regulatory documents specify that changes in facility design will either be consistent with the existing authorization basis or the authorization basis will be revised based 

upon a determination of the potential impacts of the proposed changes. The RU determined that, to the contrary, the Contractor had neither established nor implemented a program or 

process to ensure that the authorization basis was maintained current with respect to the facility design. In short, the facility design was proceeding without consideration of the authorization basis and safety issues. This condition existed throughout Part B1 of the TWRS-P project, from contract award (August: 1998) through two inspections (October and November 1999) through several meetings witn the Contractor until the CAN was issued. The RU and the Contractor held several other meetings and the RU ultimately accepted the Corrective Action Plan (CAP)28. The RU had also planned an inspection on the self-assessment and the CAP 29

. However, the contract termination resulted in the indefinite rescheduling of the inspection and 

CAP ". 

The DOE/RU regulatory program has a significant amount of flexibility and discretion for regulatory decisions and acl:ions. This allowed the postponement and subsequent cancellation of an inspection that would have resulted in a corrective action notice a year earlier than when it actually was issued, and, as a result, most of the design activities occurred outside of the authorization basis and SRE'. The RU subsequently received a large number of authorization basis amendment requests (ABARs) in the last 3 months of the contract. The CAN was not issued until 1 month before the FFP. If the CAN had been issued a year earlier, there would be 

more confidence that the design is within the authorization basis. In the NRC regulatory regime, it is unlikely that such an inspection would have been postponed and then canceled if the licensee was known to be outside their design basis. 

27 Gibbs, D.C., U.S. Department ci: Energy, letter to M. Bullock, BNFL Inc., March 14, 2000. 

28 Gibbs, D.C., U.S. Department of Energy, letter to M. Bullock, BNFL Inc., April 25, 2000. 

29 Gibbs, D.C., U.S. Department of Energy, letter to M. Bullock, BNFL Inc. March 27, 2000. 

30 Gibbs, D.C., U.S. Department c,1 Energy, letter to P. Strawbridge, BNFL Inc., July 26, 2000.

NUREG-1747 34
3.1.4 SUMMARY

The Contractor organization experienced rapid growth during Phase IA and Phase IB-1. 

Adherence to procedures and protocols, and integration of program activities and 

documentation (authorization basis) were not fully developed. The emphasis on schedule and 

cost may be contributing to these concerns. In the near term, this is unlikely to change 

because of new contracts and new contractors.

35 NUREG-1747
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NUREG-1 747 36
3.0 NRC AREAS OF REVIEW 

3.2 SEISMIC AND STRUCTURAL CONSIDERATIONS 

3.2.1 TANK WASTE REMEDIATION SYSTEM-PRIVATIZATION: PART A 

The privatization concept developed by the U.S. Department of Energy (DOE) for the Tank 

Waste Remediation System-Privatization (TWRS-P) project set forth the contractual 

requirements for the two contractors in Part A, and was expanded on in a series of four top

level documents. These were known as DOE/RL-96-003, -004, -005 and -006. Briefly these 

documents addressed the following topics for the two contractors, LMAES and BNFL Inc.3

" 

-003: the DOE Regulatory Process for TWRS-P 

-004: Process for Establishing a Set of Standards and Requests 

-005: Concept of the Regulatory Process for TWRS 

-006: Top-Level Principles 

Based upon these documents, it was the responsibility of each contractor to devise a 

pretreatment process, a vitrification process and all the auxiliary and supporting processes, and 

the plant design so that the treatment of the tank wastes would result in a product that will meet 

the acceptance criteria for high level waste disposal. The hazards specific to each of the 

contractors' processing activities were to be identified and evaluated for their impact on the 

safety on each of their proposed facilities throughout its lifetime. The DOE privatization concept 

included the execution of an integrated safety management system that would consider the 

hazards, the consequences of failures, and the mitigation elements that would be provided in 

the facility design and operation. Each contractor, based on implementing these four top-level 

documents, was to identify and recommend for DOE approval, a set of safety standards that 

the contractors would certify that, if properly implemented, would result in adequate safety, 

compliance with applicable laws and legal requirements, and conformance to the DOE

stipulated top-level safety standards and principles. For the NRC, the elements of the defined 

regulatory process with the most importance to the NRC at this stage of the project were the 

Standards Approval, the Initial Safety Evaluation, and the Construction Authorization Request 

(CAR). These elements in a final form would reflect similar aspects that the NRC would 

consider in a preliminary safety analysis report (PSAR) that would be submitted by an applicant 

for a license prior to NRC authorizing construction of a facility. In this instance however, DOE 

was interacting with the contractor (or applicant) in the conceptual design and preliminary 

design phases. These represent a much earlier stage of involvement than the NRC would 

normally be engaged in. DOE, in this situation as being responsible for the cost of the facility 

as well as safety, was motivated to be involved in the entire developmental process for this one

of-a-kind facility. 

The Standards Approval was a step in the Part A contract procedures to allow for DOE review, 

evaluation and subsequent approval, if warranted, of each of the contractors' proposed set of 

radiological, nuclear, and process safety standards and requirements, as each contractor 

defined in their Safety Requirements Document (SRD). In addition, each contractor was to 

31 Note that while there were two contractors involved in the work under Part A, all specific comments relative to this 

part are based only on the submittal by BNFL Inc. since the other contractor LMAES, did not proceed to Part B.

37 NUREG-1747
provide a standards-based inlegrated safety management program that was to be documented 

in an Integrated Safety Management Plan (ISMP). DOE required that the SRD document all 

requirements of applicable laws and regulations, conform to the -0006 top-level document, and 

that the set of standards and requirements identified by a contractor was generated by the 

standards process defined in the -0004 top-level DOE document using the appropriate level of 

expertise. The SRD was also to reflect the fact that all hazards associated with the proposed facility and its operation have been appropriately addressed. The ISMP, developed as a portion of the Standards Approval process, was specified by DOE to include the planning elements of 

the implementation plans required by DOE regulations, particularly 10 CFR Part 830 addressing Nuclear Safety Management:. DOE required that the contractor's selected safety management 

process documented in the ISMP be standards and requirements base, and that it be 

appropriately tailored to the hazards associated with the contractor's proposed facility, its 

operation and its deactivation. 

The only specific guidance provided by DOE that could be considered as a yardstick against which the level of safety of each of the proposed facilities could be judged consisted of the 

radiological Dose Standards Above Normal Background for individuals. These consisted of a 

breakdown for individuals in various categories, under various event scenarios, with varying estimated probabilities of occurrence that were provided in terms of numerical dose values, 

except for two types of individuals (the worker and co-located worker) impacted by the unlikely 

events and the highly unlikely events. The dose limits for these two cases were to be provided by the contractors as ref lectecl by their proposed design. These standards were provided by DOE in the -0006 top-level doacument with the statement that the use of the standards does not 

provide a blanket waiver to safety regulations that apply to DOE activities, but are additional 

considerations that must be addressed by each of the contractors. 

The BNFL Inc. initial SRD submittal (September 26, 1997) that was reviewed identified two 

types of standards along with the source of the standards as follows. Standards used to 

specify certain objectives that: must be achieved, or in other words performance objectives, 

were identified as "safety criteria." It was noted that these standards usually are made up of 

laws and regulation requiremnents. Standards that are a prescriptive method for achieving a 

certain objective were identified as those standards that are industry consensus codes or 

standards such as American Society for Testing and Materials standards. The initial submittal 

also indicated that the conseq jential hazards for the facility consisted of those that were 

chemical and/or radiological in nature. It was also indicated that the hazard analysis would 

identify the need for accident prevention controls or mitigation controls in order to meet the 

performance goals. The mitigation controls were noted as being either engineered or 

administrative with a preference for engineered mitigation features. 

With respect to the engineering and design of these mitigation features, BNFL Inc., as a result 

of the process hazards analys s, elected to identify two classes of structures, systems and 

components - Design Class 1 and Design Class II. 

Design Class I was designated for those structures, systems, and components (SSCs) where 

structural controls were indicated from the preliminary accident analyses as being necessary to 

protect the public from seismic-induced failure of the high level waste (HLW) receipt tanks. 

The Design Class II SSC cate-!gory based on the preliminary accident analyses indicated that 

the workers needed to be protected from the following list of accidents:

NUREG-1747 38
1. Failure of the HLW feed receipt tank. 

2. Self-boiling of the cesium product. 

3. Cesium ion exchange column fire. 

4. Melter failure. 

5. Cesium product storage overpressure or canister drop. 

BNFL Inc. noted that engineered features required to prevent or mitigate these accidents and 

the resulting hazards would be provided in the facility. The engineered features were 

envisioned to consist of those elements providing confinement for the process cells, the 

ventilation system for the process cells and the cell ventilation stack. Other potential Design 

Class II elements included certain radiation monitors, instrumentation and controls including 

interlock logic devices controlling entry into high radiation areas, isolation devices relied on for 

maintaining ventilation confinement control of radioactive material, and shielding in certain 

areas. This information was presented by BNFL Inc. as constituting the "safety criteria" 

standards. 

The consensus codes and standards were the next portions of the Standard Approval process 

and provided a more usable set of standards for the seismic, civil and structure technical 

discipline subject area. BNFL Inc. stated that the intent of the design process was to use 

consensus codes and standards that are recognized, and accepted. Design documents would 

be based on DOE guidance, and US and United Kingdom commercial nuclear and chemical 

industry practices for designing SSCs. In cases where no consensus code or standard exists, 

BNFL Inc. indicated an ad hoc standard will be developed. At the time of the submittal of the 

initial SRD, the need for any such ad hoc standards had not been identified. For the seismic, 

civil and structural discipline areas, BNFL Inc. relied heavily on the consensus codes and 

standards used in the commercial US nuclear power industry along with standards and guides 

that have been developed and incorporated into the DOE organizational hierarchy. In addition, 

guidance documents used by the U.S. Nuclear Regulatory Commission (NRC) in the licensing 

of commercial power reactors were also listed within the SRD. 

Design loads resulting from natural phenomena were provided in tabular forms for each design 

class, i.e. for Design Class I and Design Class II SSCs. The main document referenced in 

these tables was DOE-STD-1 020-94 with Change No. 1, 1996, "Natural Phenomena Hazards 

Design and Evaluation Criteria for Department of Energy Facilities." Additionally, DOE-STD 

1021-93 with Change No. 1, 1996, "Natural Phenomena Hazards Performance Categorization 

Guidelines for Structures, Systems and Components" was referenced, and for straight wind 

loads and snow loads the references were to ASCE 7-95, "Minimum Design Loads for Buildings 

and Other Structures," American Society of Civil Engineers. Tornado loads and tornado 

missiles were noted as not applicable for the Hanford Site for any design class of structure, 

however a wind generated missile was to be considered for Design Class I structures. 

Subordinate to these referenced documents were the implementing codes and standards that 

were identified for both design classes. The basis for the seismic loads for Design Class I was 

noted to be a previous report for the Hanford site that had been encompassed for application to 

all facilities at the Hanford site. This was known as the "Geomatrix Report (1996)" which had 

the title of, "Probabilistic Seismic Hazard Analysis DOE Hanford Site, Washington," WHC-SD

W236A-T1-002, Rev. 1. Based on this document BNFL, Inc. indicated that an equal-hazards 

response spectra for a peak horizontal ground acceleration of 0.24g at 33 Hz and a peak 

vertical ground acceleration of 0.1 6g at 50 Hz was appropriate for the facility. A graphical 

representation of these spectra was provided for the case of 5 percent damping. The codes and

39 NUREG-1747
standards referenced for r)e;sign Class I included ASCE 4-86, "Seismic Analysis of Safety Related Nuclear Structures;" ACI 349-90, "Code Requirements for Nuclear Safety-Related Concrete Structures;" AISC 1\1690-95, "Specification for the Design, Fabrication and Erection of Steel Safely-Related Structures for Nuclear Facilities;" and UBC-94, "Uniform Building Code." 

The basis for the seismic loads for Design Class II was identified as the Uniform Building Code, UBC-94, for Zone 2B and an Importance Factor of 1.0. The reinforced concrete design was to 

follow ACI 318-95, "Building Code Requirements for Structural Concrete," and the structural 

steel design was to follow Al SC-89(ASD), "Specification for Structural Steel Buildings-Allowable 

Stress Design(ASD)." 

NRC comments in this technical area were provided to the DOE/Regulatory Unit (RU) as part of 

the NRC Comments on the I3NFL Inc. Standards Approval Package, transmitted on 10/17/97 

and 10/30/97. The following iitems in the category of hazards and codes/standards were 

addressed and identified as requiring additional information: 

1. Natural events and the 'elated hazards that are beyond the design basis should be 

included in the hazards analysis and the appropriate mitigative measures identified as 

needed should be ident'fied and discussed. 

2. Basis of probability of oc'currence for design precipitation intensities relative to the length of 

meteorological data at tiie Hanford Site needs to be provided. 

3. The Hazards Analysis Report (HAR) indicated that tornados would not be considered 

based on the low annual probability of occurrence (once every 100,000 years) and the low 

wind speed that would be less than the straight wind speeds. This event is still categorized 

as a Frequency 1 event per Table 3-3 of the HAR. In addition, besides the wind velocities 

associated with a tornado, there are rotational velocities and differential pressures that can 

be potentially dangerous,, for closed structures and those with controlled pressure differentials for contamirnation control. The basis for ignoring all effects of tornados, even 

though the tornados may be less intense than a maximum tornado, should be justified 

further. 

4. The HAR indicates that the return period for the design basis earthquake is approximately 2000 years. The basis for the peak horizontal ground acceleration being 0.24 g for this 

return period needs to be provided. Based on the DOE top-level documents this is an 

unlikely event. There should be information provided regarding the consideration and 

treatment of seismic evrenrts in the extremely unlikely category of seismic events which 

have a return period frornl0,000 to 1,000,000 years. 

5. The SRD was noted to reference numerous commercial nuclear power plant standards, 

portions of which may be inappropriate, or conflicting for TWRS-P application. The all

encompassing references to such standards should be re-examined and the specifics 

identified. 

6. The use of non-US consensus codes and standards should be compared to the associated 

US codes and standards. The development, adoption, and use of ad hoc standards 

should be clearly prescribed in a documented procedure.

NUREG-1747 40
7. References in several locations in the document to the Uniform Building Codes (UBC) 

does not identify the year of the UBC, or identifies an old version of the UBC. These 

issues should be clarified with references corrected to the current UBC edition. 

8. The safety classification system of the SSCs into the "design classes" relative to the 

hazards and the consequences does not appear to be consistent. 

As noted by the RU in a letter to NRC, dated 12/5/97, these items were already addressed by 

RU questions to BNFL Inc., or provided information for additional questions the RU asked of 

BNFL Inc. The RU considered Item 6 to be editorial, but NRC disagreed with that position. 

While BNFL Inc. provided the RU additional information in response to some of the above 

issues as well as others, the RU's Evaluation Report of the BNFL Inc. SRD (RLIREG-98-01, 

March 1998) identified issues in this technical discipline area requiring resolution prior to 

commencement of preliminary design as well as others requiring resolution prior to construction 

authorization. With respect to hazards, the top-level Safety Criteria 4.1-3, 4.1-4 and 4.3-3 were 

identified as only addressing the effect of natural phenomena and hazards, but not all 

categories of potential hazards. The evaluation identified that Safety Criteria 4.1-3 and 4.1-4 

also established seismic design criteria for which BNFL Inc. had not provided an adequate 

safety basis, and the criteria only addressed Design Class I and Design Class II SSCs and not 

all SSCs important to safety. The design classification system as proposed by BNFL, Inc. was 

identified as being inadequate and requiring changes that BNFL Inc. had described in a letter, 

dated 2/19/98. Correction of these deficiencies was required by the RU prior to the 

commencement of preliminary design. With regard to the site description and its impact on 

design, the RU's evaluation stated that adequate justification was needed to support the BNFL, 

Inc. hazards approach for natural phenomena hazards design criteria, and this should be more 

than that described in the HAR. The RU stated that BNFL Inc. did not provide an adequate or 

sufficiently detailed safety basis to support the selection of a 0.24 g horizontal acceleration 

earthquake with a 2000-year return period. The RU noted that these were issues that are 

commonly resolved during preliminary design, but essential to resolve prior to construction; 

hence, it was required that these be resolved prior to construction authorization. With respect 

to codes/standards, the RU determined that the implementing codes and standards were 

adequate as subordinate standards. The NRC had noted that two of the referenced standards 

were currently outdated by a later edition and recommended that ACI 349-97 and UBC-97 be 

the referenced standards. This was not reflected in the RU safety evaluation report. 

Another key element within Part A of the DOE contract was the development, by each of 

the contractors, of an Initial Safety Assessment (ISA) for the proposed TWRS-P facility. DOE, 

as part of the contract with its contractors, required that the ISA provide the following elements 

(the NRC also believed these elements to be relevant for its technical safety review and support 

to the DOE RU): 

1. An adequate description of the site, the design of the proposed facility and the operation. 

2. A description of the hazards, the potential design-basis events and the analysis of those 

potential events. 

3. Preliminary acceptance criteria.

41 NUREG-1747
4. Identification and description of SSCs important to safety and the basis for the 

classification. 

5. Evaluations of the proposed design for constructability, operability, reliability, 

maintainability and inspectability. 

6. An Initial Safety Analysis Report (ISAR) that defines the projected safety basis for the 

facility in terms of physical design, SSCs with prescribed safety functions, the operating modes and conditions, a.rong with representative events of off-normal internal events, and 

external events. The results of representative safety analyses considering uncertainties 

should describe how the, facility will meet the SRD requirements as well as how the 

prescribed dose limits are met. 

The BNFL Inc. ISA submittal (January 12, 1998) was reviewed in the technical area of seismic, civil, and structural areas, and a series of comments/questions were developed. In some cases 

the subject of the issue represented a broader focus that these technical disciplines whereas in 

other cases the issue was unique to these disciplines. NRC comments in this technical area 

were provided to the DOE RU in a letter, dated 2/6/98, in support of their review process. A 

total of ten comments/questions resulted from the NRC review in these particular technical 

disciplines. These can be broadly identified as issues related to the design safety classes/categories, to external events (both natural and man-induced), and to the availability of 

information on various buildings and their function in the complex and the hazards they may 

pose. The consolidated issues addressed the following topics 

1. Three design classes are noted in the ISAR with Design Class I and II being for the 

protection of the public and workers respectively, however all the credible scenarios which 

can change the requiredt safety classification do not appear to have been considered. 

2. All potential events considered are not identified in the ISAR and those considered 

incredible are not identified nor is the basis for such a classification provided. 

3. Design Class II SSCs appear to be treated differently in the SRD and the ISAR relative to 

external events. 

4. Other adjacent sites and facilities hazards are not always considered. 

5. Specific design criteria were questioned regarding the basis of the criteria. 

As noted by the RU in a letter to the NRC, dated 2/26/98, the comments were already encompassed by RU comments/questions to BNFL Inc. or revisions were being made to those 

items, or a new comment/question was being sent to BNFL Inc. The RU noted that three of 

the original NRC comments were determined to not address information that DOE believed was 

needed at this stage of the design development process, so they were not considered further. 

The RU Initial Safety Evaluation Report of the BNFL Inc. ISA (Main Reference 10) identified 

issues in this technical discipline as requiring resolution. The issues for the entire facility as a 

result of the ISA were grouped into four categories as follows: (1) adequate safety basis not 

demonstrated, (2) inadequat,' classification of SSCs, (3) incomplete or conflicting elements of 

the authorization basis, and (4) incomplete design or operational information. The schedule for 

resolution was identified in ternns of three distinct time frames as follows: during the first six

NUREG-1747 42
months of the preliminary design (Part B-i), during the PSAR stage, and prior to the submittal 

of the Construction Authorization Request. 

The RU stated that the justification for the use as a design basis earthquake of an event with a 

peak horizontal ground acceleration of 0.24g, with a return period of 2000 years, had not been 

derived or provided based on a comprehensive hazard/consequence analysis. Further, it was 

noted that there was a need to identify and justify how natural phenomena hazard design 

requirements apply to SSCs and their safety functions. The use of specific editions of codes 

and standards as well as specific provisions were identified by the RU regarding tornados, site 

precipitation flooding, and assigned importance factors. The lack of any site-specific 

geotechnical investigation results was also identified and there was a stated need for the facility 

structural classifications to be defined. The RU identified these items for resolution within 6 

months after the preliminary design is initiated. 

As a result of work underway in the NRC in 1998 to revise 10 CFR Part 70, which was 

ultimately expected to be the regulation the TWRS-P facility would have to comply with if the 

private owner-operator were to be licensed by NRC, an Issue Paper was developed and 

provided to the RU. The paper, "Consideration of the Seismic Events for Integrated Seismic 

Analysis of Hanford Tank Waste Remediation System," dated 6/29/98, discussed recently 

issued consensus standards and guidance that could be considered as the baseline minimum 

set of seismic criteria. 

3.2.2. TANK WASTE REMEDIATION SYSTEM-PRIVATIZATION: PART B-1 

When Part B-1 of the contract between DOE and BNFL Inc. began, revisions were being made 

by BNFL Inc. to the previous documents in response to the issues identified. A series of topics 

that identified specific high-interest issues judged to be key issues needing technical resolution 

were scheduled. In this discipline area, the identified topic was "seismic;" however, before the 

first scheduled Topical Meeting was held there were several pre-meetings. NRC staff attended 

such a meeting on 11/6/98 where the topic of discussion focused on BNFL Inc.'s request to 

obtain concurrence from the RU on the peak ground acceleration value for the seismic analyses 

of the required SSCs. The basis for the BNFL, Inc. proposed design basis earthquake of 0.24g 

peak horizontal ground acceleration was identified as DOE-STD-1020-94, with Change No. 1, 

1996, and a report addressing the entire Hanford site entitled, "Probabilistic Seismic Hazard 

Analysis, DOE Hanford Site, Washington," WHC-SD-W236A-TI-002, 2/14/94 through Rev. 1A, 

10/8/96, 02, by Geomatrix Consultants, Inc. Issues that were identified included the fact that it 

was not clear that new data and new concepts on the seismicity of the area may be available 

since the time the Geomatrix report was developed, whether distant events could be more 

dominant in some frequencies than nearby events, the impact on response of reflected 

energies from a bowl/basin effect for the Hanford site, the extent of the peer reviews conducted 

on the Geomatrix report, and the level of BNFL Inc. review/validation of the report to conclude 

that it can be applied to the TWRS-P facility. As a result of meeting on 11/6/98, BNFL Inc. 

submitted additional information on 11/19/98 for use in the Seismic Topical pre-meeting that 

was held on 12/6/98. NRC provided comments to the RU (12/11/98) after a review of the 

additional BNFL Inc. information. The key issues that had been identified included the 

following:

43 NUREG-1747
1. The submittal is a much broader request for approvals than just the horizontal peak ground acceleration value. It also included the response spectra and a document outlining the 

seismic design approach. 

2. Noted that the submittal seemed to contain conflicting information regarding the Design Basis Earthquake and thie Performance Categories of DOE-STD-1020-94, and the return 

period of 2,000 years. 

3. While mentioned in the submittal, it was unclear how the seismic analysis and design of 

SSCs in other than Pedrormance Category 3 were to be addressed. 

On December 14, 1998, a Tc)oical Meeting was held on the subject of the TWRS-P peak ground acceleration in order to discuss the basis for the derived value. However because the 

pre-meeting discussion had included related topics, the RU asked BNFL Inc. to address the 

relationship of the 2000 year return design basis earthquake to the extremely unlikely events 

and the consequences that must be met for the project when considering the Radiation 

Exposure Standards for Workers (i.e., these must be addressed for events with return periods from 10,000 to 1,000,000 years). As a result of the meeting there were several key items left 

for resolution: 

1. The current validity of the Geomatrix Report for use on TWRS-P must be developed including addressing basin effects, site specific attenuation data, uncertainty, and peer 

review. 

2. The use of the DOE series of standards related to DOE-STD-1020-94. 

3. The dip in certain frequencies of the proposed horizontal ground response spectra for 0.24 

g; d.) The basis for the statement that there was a 5 percent chance of building collapse for a beyond Design Bas-is Earthquake. 

Target dates for additional inlormation were established by the RU and BNFL Inc. for February 1999 for all items except a May 1999 date was set for the availability of a document entitled, "Seismic Design Approach/Criteria." NRC comments on the Topical Meeting were contained in 

an NRC Memorandum dated 2J2/99 that reiterated these major outstanding issues. 

A meeting was held on Januaiy 7, 1999 as a follow-up meeting to the December 14, 1998 

Seismic Topical Meeting to focus on the Geomatrix Report and the questions that arose from 

the review of the report by the,, RU and its consultants. A list of questions the RU had 

developed prior to the Seismic: Topical Meeting that remained unanswered as well as additional 

questions developed from the Seismic Topical Meeting were the focus of this meeting. The 

more important issues included the following: 

1. Relationship of geologic structure to the historical seismicity. 

2. Comparison of attenuation relationships used for subduction-zone earthquakes to the 

empirical database in the range of distances to the site. 

3. Method of scaling peak accelerations for subduction-zone earthquakes for the site-specific 

response analyses.

NUREG-1747 44
4. Comparison of input parameters for the probabilistic seismic hazard analysis to the US 

Geological Survey (Frankel, 1996 ) values. 

Most of the issues were satisfactorily resolved in the meeting. However there was a list of nine 

additional issues/questions that remained. These were provided to BNFL Inc. by the RU in the 

meeting minutes issued, dated 2/1/99. At the meeting it was agreed that BNFL Inc. would 

provide the answers to the additional issues on 2/28/99 in a report. The key items in this list 

included the following: justification for the weighting factors used for the various fault models in 

the Preliminary Seismic Hazard Analysis (PSHA) logic tree relative to the geological 

characteristics should be provided and there should be some sensitivity calculations to 

demonstrate the impact of weighting factors on the peak ground acceleration value as well as 

the response spectra. 

On March 18, 1999, BNFL Inc. transmitted a package of seismic documents to the RU with the 

statement that the four enclosures adequately addressed all the comments and questions 

raised by the RU and provided an acceptable basis for the design basis earthquake being a 

peak horizontal ground acceleration of 0.26g. A request for approval of that value was 

requested. The four enclosures were as follows: 

1. "TWRS-P Facility Design Basis Earthquake-Peak Ground Acceleration, Seismic 

Response Spectra, and Seismic Design Approach," Rev. 1, 3/17/99. 

2. "Applicability of DOE Documents to the Design of TWRS-P Facility for Natural 

Phenomena Hazards," Rev. 0, 3/17/99. 

3. "Validation of the Geomatrix Hanford Seismic Report for Use on the TWRS Privatization 

Project," Rev. 0, 3/17/99. 

4. "Approach for Ensuring Compliance with the TWRS-P Radiation Exposure Standards 

under Earthquake Conditions," Rev. 0, 3/17/99. 

The NRC reviewed these documents and provided comments to the RU in a letter dated 

5/10/99. The comments consisted of 10 items related to Enclosures 2 and 4. NRC 

summarized the comments in the transmittal letter by stating the following, "...whether or not the 

adoption of a 2,000 year return period for the seismic design basis event will result in a TWRS

P facility that will meet the DOE radiation exposure standards for this project has yet to be 

demonstrated." Additionally it was stated that, "...while Enclosure 4 to BNFL Inc.'s submittal 

presents and proposes the use of a methodology that might be applicable for such a 

demonstration, the details for its execution and the source for the needed data to conduct such 

an analysis have yet to be determined. Consequently, the possibility remains that some SSCs 

might have to be designed to a seismic level (i.e., peak ground acceleration) above 0.26g, or 

that changes in the stress allowables or load factors might have to be made to provide 

reasonable assurances that allowable exposure levels (dose limits) will not be exceeded with a 

frequency of occurrence per year equal to or greater than one in one million." 

The RU in a May 14, 1999, letter to BNFL Inc. described a conditional acceptance of the peak 

horizontal ground acceleration at 0.26g that stipulated that their enclosed concerns must be 

satisfactorily resolved.

45 NUREG-1747
BNFL Inc. presented a seismic probabilistic risk analysis (PRA) approach at the Topical Meeting in October 1999. This was comprised of several models on different failure modes and releases. Extensive use was made of fragility curves and "small leak before break" arguments, for both civil structures (e.g., cells) and equipment (e.g., tanks). The median dose result (50 percent) for collocated workers corresponded to the SRD limit of 25 rem with a frequency of 1 E 6/yr. However, the analysis indicated the facility would survive and have functioning prevention and mitigation features with earthquakes considerably stronger than 0.26 g. Upon questioning, BNFL Inc. indicated that input values had been assumed to illustrate and refine the 

methodology. Actual data and design values would be used after several months of further 

model development and with actual design parameters and features (i.e., once finalized - after 

the Firm Fixed Price submittal). 

3.2.3 PRESENT STATUS 

At the time of the contract cancellation by DOE, the issue related to the seismic area of review 

was the acceptability of the exact PRA methodology that would be used and the validity of the 

input data, in order to substantiate that the dose limits for the extremely unlikely events will not 

be exceeded. The criteria and methodology that will be used for the civil and structural design 

are outlined in the project documents and are considered to be acceptable once there is 

agreement on the safety classifications of SSCs. The site specific geotechnical data were 

provided in the last BNFL Inc. submittal prior to contract cancellation; however the NRC was not 

provided a copy of that specific material. The NRC has no information regarding DOE 

acceptance of that geotechnical information and its impact on the soil and structural foundation 

designs. In addition, the NFIC does not have any information on the compatibility of current and 

near-term site activities with the proposed design or plans from the new, Management & 

Operations contracts.

NUREG-1747 46
3.0 NRC AREAS OF REVIEW 

3.3 HAZARDS AND SAFETY ANALYSES (ISA/ISM) 

3.3.1 STANDARDS APPROVAL PACKAGE (SAFETY REQUIREMENTS DOCUMENT AND 

HAZARDS ANALYSIS REPORT) AND INITIAL SAFETY ANALYSIS REPORT 

The Safety Requirements Document (SRD) and Hazards Analysis Report (HAR) arrived from 

BNFL Inc. in September 1997 and identified numerous process and chemical safety scenarios 

(many with both radiological and chemical consequences) using a preliminary hazards

operability (HAZOP) method. HAZOP reviewers with expertise in similar types of plants and 

processes were used to bridge areas of incomplete information because the preliminary design 

was still undergoing development and, as such, many open areas existed. The HAZOP used 

an American Institute of Chemical Engineers (AIChE) risk matrix (a binning approach) to assist 

with identifying those potential scenarios that posed a sufficient risk (i.e., consideration of both 

frequency and consequence) to warrant quantitative risk analysis and the consideration of 

safety controls. In general, the list included many entries of a benign nature that obscured 

some of the more hazardous conditions at the proposed facility. In addition, the expert review 

panel inadvertently included the effects of mitigation in some of their assessment, and, 

consequently, the frequency and consequence bins for some events were underestimated. 

Thus, potential items relied on for safety (IROFS) were overlooked. From these reviews, the 

following comments were raised in the hazards and safety analysis area: 

1. Incomplete process and hazard descriptions. 

2. Inconsistency of unmitigated and mitigated analyses. 

3. Emphasis on active instead of passive mitigation and control. 

4. Organization of documents and information. 

5. Inconsistencies in hazards identification, structures, systems, and components (SSC) 

categorization, and design classes. 

6. Relative lack of quantification. 

7. Presence of large tanks and inventories. 

8. Limited consideration of off normal and unanticipated events, and interactions. 

The actual comments may be found in the staff's SRD/HAR comment transmittal letter3 2. 

BNFL Inc.'s Initial Safety Analysis Report (ISAR) was dated January 12, 1998. The ISAR 

represented a more detailed design as compared to the SRD/HAR package. The ISAR 

advanced the design and presented quantitative analyses on several process and chemical 

safety scenarios, including tank failures, crystalline silicotitanate (CST) powder dispersion, ion 

32 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, 

October 17, 1997.

47 NUREG-1747
exchange resin fires, glass spill, anhydrous ammonia leaks, and a nitric acid spill. However, the 

level of the design corresponded to a preliminary, conceptual level, identified as circa "2-3% of 

design" by the U.S. Department of Energy (DOE) and the Contractor. From the perspective of 

safety reviews and regulation, the ISAR did not contain sufficient information for addressing 

many of the concerns and issues. Furthermore, the input values and methods used appeared 

to be nonrepresentative anc would underestimate the potential consequences. From the ISAR 

review, the NRC staff raisec the following comments and questions in the process and chemical 

safety arena: 

1. Lack of conservatism. 

2. Inconsistency of unmitigated and mitigated analyses. 3. Emphasis on active insl:ead of passive mitigation and control. 

4. Organization of documents and information. 

5. Inconsistencies in hazards identification, SSC categorization, and design classes. 

6. Relative lack of quantification. 

7. Presence of large tanks and inventories. 

8. Limited consideration of offnormal and unanticipated events, and interactions. 

The actual, detailed commetils may be found in the NRC staff's ISAR comment transmittal 

letter,3. 

3.3.2 DESIGN SAFETY FE'ATURES SUBMITTAL 

The Design Safety Features (DSF) submittal from BNFL Inc. consisted of two parts. Part 1 

discussed general features used to address general hazards and concerns. In essence, this 

part presented the practical interpretation of the codes and standards from the SRD. 

"Standard" chemical process industry approaches would be used unless radiological limits were 

shown to be exceeded. Pard 2 discussed and analyzed 10 examples in more detail, including loss of confinement events with both radiological and chemical consequences. The 10 

examples were: 

1. Hydrogen generation in the high level waste storage vessels. 

2. Loss of cooling to the cesium storage vessel. 

3. Load drop of a pretreatment pump (out of cell). 

4. High level waste melter feed line failure. 

5. Cooling water contamination. 

6. Sample carrier breakout. 

7. Low activity waste pipe break. 

8. Receipt vessel rupture. 

9. Activity backf low. 

10. Nitric acid handling. 

The following, main comments and questions were raised by the NRC staff regarding safety 

analyses from the DSF review: 

33 Pierson, R.C., U.S. Nuclear ReCgulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, February 

6, 1998.

NUREG-1747 48
1. Concerns about adequate margin, conservatism, and defense-in-depth resulting in 

potential under-estimation of consequences and IROFSs. 

2. Relative lack of information, details, specificity, and numerical values on many features. 

3. Design basis not identified vis-a-vis "best basis." 

4. Uneven approach to dose methodology including the handling of unmitigated events. 

5. Use of optimistic design assumptions and reliabilities. 

6. Uncertainties and inconsistencies not addressed. 

7. Validation and justification of data, assumption, and models. 

The actual comments may be found in the NRC staff's DSF comment transmittal letters. The 

DSF constituted an example of approaches for safety analyses and did not represent an actual 

design or safety document. 

3.3.3 RELATED ACTIVITIES AND TOPICAL MEETINGS 

The DSF submittal and related subjects were discussed at the Topical Meeting that was held in 

Hanford on March 27, 1999. The lack of detail was noted on the design and many of the 

proposed safety features. The April 1999 Topical Meeting presented the dose methodology to 

be used in the Contractor's integrated safety management (ISM) process. Many questions 

were raised and there were concerns about the limited documentation, the relevance of the 

Sellafield database, source term calculations, and the incorporation of decontamination factors 

in unmitigated analyses. Subsequent Topical Meetings and reports used parameters with 

fluctuating parameters and values, often without explanation and justification. For example, the 

October 1999 Topical Meeting used values that were just selected as examples, without any 

stated basis or criteria for selection. The November 1999 Topical Meeting discussed the ISM 

Cycle 1 results while the May 2000 Topical Meeting discussed the ISM Cycle 2 results. Both 

ISM cycles were not based on the design existing at their respective times. 

The Firm Fixed Price (FFP) submittal did not contain any hazards or safety analyses because 

DOE and the Contractor had intended it to be a design basis type of document, suitable for 

supporting cost and safety analyses but not necessarily including them. However, the NRC 

staff participated with DOE in the review of the FFP submittal from the perspective of it 

supporting the Construction Authorization Request/Preliminary Safety Analysis Report 

(CAR/PSAR). The NRC staff agreed with DOE that the facility and process design information 

provided in the FFP falls far short of the level of information required to support the CAR/PSAR 

(see Section 2.5). 

"4 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, March 18, 

1999.

49 NUREG-1747
3.3.4 NRC ANALYSIS AND EVALUATION

The NRC staff used generic and conceptual process approaches proposed by DOE contractors 

to analyze potential risks from process and materials aspects at potential TWRS-P facilities 

(Chapter 7.0, Main References 18 and 19). These analyses identified the following areas of 

concern: 

1. Radiochemical inventories. 

2. Process efficacy. 

3. Organic ion exchange resin/nitrate interactions. 

4. CST drying. 

5. Organic materials. 

6. Radiolysis. 

7. High temperature operations. 

8. Nonradioactive chemical effects upon radiochemical processing. 

Several of these areas of concern involved events that could be analyzed at this early stage of 

design using a conservative bounding approach suitable for an initial assessment of risk, 

determination of relative implcrtance, and the preliminary categorization of SSCs and 

identification of controls. The U.S. Nuclear Regulatory Commission (NRC) staff used 

parameters from the privatization contracts (Chapter 7.0, Main References 20 and 21) to 

estimate inventories and assess materials at risk. NRC and DOE accident handbooks (Main 

References 22 and 23) outlined several scenarios (e.g., spray leak, tank rupture) with 

suggested values for release parameters (e.g., atmospheric release fractions, respirable fractions); the staff primarily selected bounding values as these were recommended by the 

handbooks for preliminary accident analyses on conceptual designs. Dispersion calculations 

focused on the receptor at 100 meters as this is a typical distance to a member of the public in 

NRC licensee analyses, and it approximately corresponds to the likely distance from a surface 

release to the facility fenceline at the proposed, TWRS/Waste Treatment Plant (WTP) location 

on the Hanford site. The calculations assumed a breathing rate corresponding to light activity. This allowed the determination of dose consequences from these various scenarios. 

The NRC staff estimated risk using the LNT (linear no threshold) model for dose 

consequences, without any modification in the risk factor for acute doses above 10 rem. 

Frequencies were based upon published values in the literature (see Chapter 7.0, Main 

References 18 and 19). 

Table 4 summarizes potential consequences from unmitigated events, and Table 5 lists 

potential mitigating controls a'Lnd their beneficial impacts. All of the accidents listed in Table 4 

have potential consequence!-; exceeding the thresholds and guidelines in regulations, including 

the revised 10 CFR Part 70, referred to as the "New Part 70" in Tables 4 and 5. Many of the 

events have frequencies in the 1 E-2 to 1 E-4 range and would be considered to reside in the "unlikely" probability bin. 

Using the U.S. national averaLge for workplace fatalities of 4.8E-5/yr (see Main Reference 18) for comparison, ten process scenarios exceed that national average (at 100 meters). The total 

estimated, unmitigated risk from a generic, TWRS/WTP facility at 100 meters due to these 

incidents involving radionuclides is approximately 2.4E-2/yr, some 500 times larger than the 

U.S. workplace average risk. For contrast and comparison, Table 6 displays additional risk

NUREG-1747 so
comparisons, and shows that the U.S. average background radiation dose dominates individual 

public radiological risk (at 1.8E-4/yr). Table 6 also lists the average risk due to cancer (2E-3/yr 

- see Main Reference 18). By comparison, the potential unmitigated risk from these process 

accidents at a TWRS/WTP facility exceeds the background dose risk by two orders of 

magnitude and the average cancer risk by a factor of ten. Four accident scenarios involving 

two forms of melter failure, and two forms of resin interactions dominate the risk by accounting 

for 90 percent of the total. A large portion of the risk from the two melter accident scenarios 

accrues from rapid thermal volatilization and dispersal of the aqueous cold cap from a 

catastrophic release of the high temperature, molten glass. Limited experimental data and 

experience are available for these melter failure scenarios. If these melter and resin accidents 

are effectively prevented and/or mitigated, the TWRS-P risk decreases to around 1.4E-3/yr, a 

level commensurate with the risk associated with occupational exposure limits, but still some 10 

times greater than the risk due to average background exposure to radiation. Several accident 

scenarios involving tank failures or deflagrations also exhibit the potential for very high doses. 

In the case of chemical storage tank failure, the potential ammonia and nitric acid releases 

would result in irreversible, deterministic health effects around the TWRS-P facility and its 

environs, and would render the facility uninhabitable for operating and control purposes. If 

liquid anhydrous ammonia were used, the affected area could extend out beyond a mile. Thus, 

prevention and mitigation are required to minimize the impact of these chemical effects upon 

radioactive materials. 

The NRC staff investigated the availability of controls for reducing the risks. Fortunately, 

relatively simple and effective, prevention and mitigation methods are available, and Table 5 

displays this situation. Prevention and mitigation methods reduce the total risk to the receptor 

at 100 meters from the TWRS-P facility to about 2.5E-6/yr. This result is about 5 percent of the 

average occupational risk and around 1.4 percent of the risk due to the average background 

dose. Incorporation of prevention and mitigation controls is likely to be acceptable to the 

revised Part 70, although further analysis may be necessary for the melter failure scenarios. 

Consequently, the preventative and mitigating design features are likely to become controls and 

items relied upon for safety, and potential examples are discussed as follows: 

1. Low Activity Waste (LAW) and High Level Waste (HLW) Tank Leak Events: Prevention 

could be accomplished by high quality tanks and components, with 100 percent weld 

inspection and leak testing, and inventory controls. The tanks and the ullage ventilation 

system (including the first high efficiency particulate air (HEPA) filter bank) could become 

the primary mitigative control. The second ullage ventilation system HEPA filter bank and 

the cell, with its ventilation system and (single-stage) HEPA filter bank and sump, could 

constitute likely secondary controls. In order to meet defense in depth requirements, the 

area outside the cell (usually an access or operating corridor) and its ventilation system, 

and a short exhaust stack represent probable tertiary controls. Specific level and spill 

detection instrumentation and alarms may also be identified as controls. Ideally, the 

inventory of liquid LAW and HLW in the TWRS-P facility should be minimized, perhaps by 

some form of an inventory limit (e.g., such as a potential limiting conditions of operation 

(LCO)). A specific design might include several smaller tanks (instead of one large one), 

of which one or more are maintained empty as a dedicated spare(s) in case of a leak in a 

full tank. The requirement for a spare tank or spare tank capacity could become a 

requirement (perhaps another potential LCO).

51 NUREG-1747
2. Cesium Tank Events: Prevention could be accomplished by high quality tanks and 

components, with 100 percent weld inspection and leak testing, and inventory controls. 

The cesium product tank(s) and the ullage ventilation system (including the first HEPA filter 

bank) are likely to become the primary control. The second ullage ventilation system HEPA filter bank and the cell, with its ventilation system and (single-stage) HEPA filter 

bank and sump, constit:ute probable secondary controls. In order to meet defense in depth requirements, the area outside the cell (usually an access or operating corridor) and its 

ventilation system, and a short exhaust stack represent likely tertiary controls. If these are 

(physically) the same iti,,;ms relied upon for safety by the LAW and HLW tanks, then an 

analysis should be performed demonstrating no deleterious effects upon the safety functions from the different tank systems and their events. Specific level and spill detection instrumentation and alarms could also be identified as controls. Ideally, the inventory of liquid cesium solution in the TWRS-P facility should be minimized or even 

eliminated, perhaps with an inventory limit (a potential LCO). It is also possible to add a 

requirement for conversion of the cesium into a solid, such as loading onto CST. A 

specific design might include several smaller tanks (instead of one large one), of which one 

or more are maintained empty as a dedicated spare(s) in case of a leak in a full tank. The 

spare tank or spare capacity could also become a requirement (e.g., another potential 

LCO). 

The cesium solution in 1:h e tank requires cooling to prevent temperature increases and, ultimately, avoid solution boiling. Thus, the cooling means and its backup potentially become items relied upon for safety. This includes the separate cooling coils or jackets (zones), the cooling sources (usually cooling water, with process and firewater backup), the pumps or means of recirculation, the piping, and redundant flow/temperature instrumentation. Interruption of cooling is likely to become an LCO, and the respective temperatures (of the tank and cooling means) Operational Safety Requirements (OSRs). If it could be demonstrated by calculation or experimentation that a tank design is 

adequately cooled by passive means under all credible conditions, then this would reduce 

the number of items relied upon for safety. 

3. Ion Exchange Column Events: Different controls can be used to provide adequate 

assurances of safety with the ion exchange columns. Preventative controls rely upon cooling, detection, and cilution/dispersal of the reacting resin mass. Resin performance improves with lower feed solution temperatures (20-250C or lower); cooler temperatures also reduce the probabili ty of resin degradation reactions. The cooling means, heat 

exchanger, and temperature indicators are likely to become items relied upon for safety. The column itself is likely to be a preventative control, because it confines a significant amount of radioactivity (i.e., once loaded - to avoid a loss of confinement) and the design 

may incorporate specific design features to cool or separate the resin bed and prevent 

runaway resin reactions. If mitigation relies upon an enclosure around the columns, venting to the cell; then the enclosure, vents/relief valves, rupture disks, and the ventilation 

systems (both the process/vessel ullage [2 HEPA banks in series] and cell [1 HEPA bank]) become the primary controls. Analyses should be performed to verify no cross effects 

upon other vessels on the same vent header from potential fumes caused by the 

degradation reactions; if such effects are discovered, then the resin columns should have 

their own ventilation sys:tem. In order to meet defense-in-depth requirements, the area 

outside the cell (usually an access or operating corridor) and its ventilation system, and a

NUREG-1747 52
short exhaust stack represent potential secondary mitigating controls. Specific level and 

spill detection instrumentation and alarms could also be identified as controls. 

4. Tank Deflagration Events: These probably require the same items relied upon for safety 

as with the tank leak events. Hydrogen and flammable gas detection systems constitute 

additional potential controls. Redundant exhaust fans, supply air fans, or nitrogen blanket 

systems may be warranted as controls if more refined calculations or experimental results 

indicate higher evolution rates for flammable gases. 

5. CST Drying Events: Prevention could use OSRs on the CST drying. Temperature and 

humidity instrumentation are likely to become the primary controls, with temperature and 

humidity endpoints determined by experimental testing. The CST canister drying 

enclosure and its filtered vent represent the first potential mitigating control but, in the 

absence of a specific design, are not credited with any source amelioration. The cell and 

its ventilation system (including two HEPA filter banks) likely become the secondary 

control; the first HEPA filter bank is assumed to be rendered ineffective by the deflagration. 

In order to meet defense-in-depth requirements, the area outside the cell (usually an 

access or operating corridor) and its ventilation system, and a short exhaust stack 

represent probable tertiary controls. 

6. Melter Events: Prevention could rely upon melter instrumentation and controls to detect 

maloperation and terminate the radioactive feed to the melter and replace it with process 

water (to maintain the cold cap), and terminate operations prior to melter release or failure. 

The proposed design had an inner gap with unspecified instrumentation; melter electrical 

and thermal sensors would become the probable controls. In the absence of a specific 

design and details, these are not credited with source prevention and reduction in this 

generic analysis by the NRC staff. The melter, its enclosure and its filtered vent would 

likely represent the first potential mitigating control but, in the absence of a specific design, 

are not credited with any source amelioration. The cell and its ventilation system (including 

two HEPA filter banks) would probably become the primary control. In order to meet 

defense-in-depth requirements, the area outside the cell (usually an access or operating 

corridor) and its ventilation system, the melter offgas system, and a short exhaust stack 

represent likely secondary controls. 

7. Chemical Events: Prevention would use high quality tanks and components with 100 

percent weld inspection, and leak testing, and a generous (conservative) corrosion 

allowance. Prevention also could use an enclosure (with a spill basin) for the reduction of 

weather effects (including diurnal thermal cycles) upon both the components and the 

chemicals of concern (principally ammonia and nitric acid). Leak detection sensors and a 

water spray/deluge system probably constitute the primary mitigative method. Exhaust 

fans (on the enclosure) activated by a separate sensor system denote a likely secondary 

mitigative method. The third mitigative method locates the cold chemical storage tank area 

away from the facility and preferably near a cooling pond or its equivalent. Potential 

catastrophic releases of chemicals may affect the operability of the TWRS-P facility, 

rendering the area uninhabitable for a period of time (possibly as long as several hours). 

Consequently, additional safety requirements may be needed for control room air, 

breathing air (self contained breathing apparatus), and/or the ability of the TWRS process 

to operate and shutdown automatically without human assistance.

53 NUREG-1747
The average worker and cancer risks presented in Table 6 include contributions from all 

sources, such as industrial accidents, environmental chemical exposures, and other 

nonradiological contributors. Therefore, acceptable limits for potential contributions from 

radiological risks associatedl with process hazards of TWRS-P are likely to be lower, perhaps a 

few percent of these averages. This preliminary analysis suggests this is indeed the case for a 

TWRS-P facility design incorporating standard nuclear industry prevention and mitigation techniques; the estimated risfk with prevention and mitigation features is 5 percent of the 

average occupational risk and 0.1 percent of the average, public cancer risk. This is consistent 

with discussions in the literal:ure (Reference 18). 

From this review of a generic facility, the NRC staff concluded that safety controls (IROFS) would be needed at the proposed TWRS/WTP facility to meet likely risk goals, and that, with 

the possible exception of the melter areas, no unusual or special controls with unique characteristics would be necessary. More information is needed on the melter designs before 

specific control strategies can be postulated and evaluated. 

In general, designs proposed by the contractors do not consider prevention and controls and 

only incorporate one mitigating means to overcome failures. The designs do not include 

important auxiliary effects in the analyses, such as common mode failures, operability, recoverability, and plant habitability for operators, and means for controlling these effects. 

Obviously, DOE and its contractors will include experimental testing as part of the program leading to the design, construction, and operation of the TWRS-P facility. Few appropriate safety related parameters, such as failure rates, modes, and release fractions, are available for 

HLW processing and vitrification facilities. It would be beneficial if the measurement of such 

safety parameters could be included in the DOE and contractor programs. 

3.3.5 PRESENT STATUS CF HAZARDS AND SAFETY ANALYSES 

An ISM Cycle 3 was originally planned for late Summer of 2000, based upon the FFP design from the April 26, 2000, submittal. However, this is unlikely to occur due to the contract 

termination and the selection of new contractors. 

It should be noted that many of the original questions and comments raised in the SRD, ISAR, DSF, FFP, and other reviews remain without adequate closure. These include the following: 

1. Incomplete process and hazard descriptions and a relative lack of information, details, 

specificity, and numerical values on many features, including IROFSs. 

2. Concerns about adequat1e margin, conservatism, and defense-in-depth resulting in 

potential under-estimation of consequences and IROFSs. 

3. Design basis not identified vis-a-vis "best basis," average, or median. 

4. Uneven approach to dose methodology including the handling of unmitigated events. 

5. Inconsistency of unmitigated and mitigated analyses.

NUREG-1747 54
Table 4: Summary of Unmitigated Events at the Generic TWRS Facility

Unmitigated Part 70 Estimated Likelihood Unmitigated 

Consequence Consequence Frequency (probability) Impact, 

Event Impact, Category (uncontrolled), Bin Risk, 

Receptor at Eventlyr yr-1 

100 meters, 

__________________ rem i " _ _ _ _ ... ..... .. .. . . .. , _. .. ._ "_ _ _ _ 

LAW Tank Failure 3,000-6,300 High 2E-5 Unlikely 3E-5 to 6E-5 

HLW Tank Failure 6,000-12,000 High 2E-5 Unlikely 6E-5 to 1.2E-4 

Cesium Tank - Loss 25,000 High 1 E-6 Highly Unlikely 1.25E-5 

of Cooling/ 

Boiling (1,000 gal) 

Cesium Tank 55 High 2E-5 Unlikely 5.5E-7 

Failure, 1,000 gal 

Basis 

Cesium Loaded 1,400 High 1 E-3 Unlikely 7E-4 

Resin/Nitrate (3% of total, 

Interaction ._unmitigated risk) 

Cesium Eluting, 3,400 High 1 E-3 Unlikely 1.7E-3 

Resin/Nitrate (7% of total, 

Interaction __unmitigated risk) 

H2 Deflag,. 20,000 High 1E-5 Unlikely 1E-4 L6AW Heel , ... I 

55 NUREG-1747
Table 4: Summary of Unmitigated Events at the Generic TWRS Facility (continued) 

FUnmitigated Pat 0 Estimated Lik~elihood Unitigated Consequence Consequence Frequency (probability) Impact, Event Impact, Category (uncontrolled), Bin Risk, Receptor at Event/yr yr-i 100 meters, 

H2 Deflag/ 38,000 High 1 E-5 Unlikely 1.9E-4 

HLW Heel 

H2 Deflag/ 2,500 High 1 E-5 Unlikely 1.2E-5 

Cesium Tank 

Heel 

H2/LAW Tank 2,300 High 1 E-5 Unlikely 1.15E-5 

Deflag/low H2 

H2/LAW Tank 115,000 High 1 E-6 Highly Unlikely 6E-5 

Deflag/high H2 

H2/HLW Tank 216,000 High 1E-6 Highly Unlikely 1.1 E-4 

Deflag/high H2 

CST Drying/ 48,000 High 1 E-5 Unlikely 2.4E-4 H2 Deflag (1% of total, ,, 1 1 unmitigated riskL

NUREG-1747 56
Table 4: Summary of Unmitigated Events at the Generic TWRS Facility (continued)

Unmitigated Part70 Estimated Likelihood Unmitigated 

Consequence Consequence Frequency (probability) impact, 

Event Impact, Category (uncontrolled), Bin Risk, 

Receptor at Event/yr yr-i 

100 meters, 

Melter/Canister 14,500 High 1 E-3 Unlikely 7.3E-3 

Failure, Cold Cap (30% of total, 

Dispersal unmitigated risk) 

Melter/Steam 26,000 High 1 E-3 Unlikely 1.3E-2 

Explosion (54% of total, 

_ _ __ _unmitigated risk) 

Chemical - > ERPG-3 High 1 E-5 Unlikely (not applicable) 

Ammonia Tank 

Failure 

Chemical - > ERPG-3 High 1 E-5 Unlikely (not applicable) 

Nitric Acid Tank 

oFailure nmitigated risk_=2.4E-2/yr 

Total, unmitigated risk 2.4E-2Iyr

57 NUREG-1747
Table 5: Summary of the Impact of Potential Controls at the Generic TWRS Facility 

Evn J Potential Potential I Mitigated I tigated I Likely -tMitigated !0 vent Controls/Items j Mitigated Consequence Consequence j Likelihood Acceptable per J Risk, Relied Upon to Receptor Category (Probability) New Part 70? year-i For Safety at 100 meters, rem _ I _I 

I AW Tank 1. Tank 2. cell/HEPA 3-6 Intermediate 2E-6 Yes 3E-9 to 

Failure 3. Enclosure/sump 6E-9 

14. Spare tank 

HLW Tank 1. Tank 2. cell/HEPA 6-12 Intermediate 2E-6 Yes 6E-9 to Failure 3. Enclosure/sump 1.2E-8 

4. Spare tank 

Cs Tank LOCA, 1 . cell/vent./2 HEPA 25 Intermediate to 1 E-6 Yes 1.3E-8 Boiling/i,000 gal 2. emerg. cooling (first HEPA fails High 

due to moisture) 

Cs Tank Failure, 1 . Tank 2. cell/vent. 0.1 Low 2E-6 Yes 1E-10 1,000 gal 3. Enclosure/sump 

4. Spare tank 

5. Cs as solid 

Cs Resin, 1. Enclosure 1.4 Low 1 E-4 Yes 7E-8 Loaded 2. Cell/vent/HEPA 

Cs Resin, 1 . Enclosure 3.4 Low 1 E-4 Yes 1 .7E-7 Elution 2. Cell/vent/HEPA (7% of total) 

H2 Deflag/ 1. Gas/vent./2 HEPA 20 Intermediate 1 E-6 Yes 1 E-8 

LAW Tank Heel 2. Cell/vent (first HEPA rendered 

_3. Sensor/N2 iniect ineffective_) I I-_1_ 1

NUREG-1747 58
Table 5: Summary of the Impact of Potential Controls at the Generic TWRS Facility (continued)

59 NUREG-1747

Potential Potential Mitigated Mitigated Likely Mitigated 

Event Controls/Items Mitigated Consequence Consequence Likelihood Acceptable per - Risk, 

Relied Upon to Receptor j Category (Probability) New Part 70? year-1 

_______________ For Safety at 100 meters, rem I _....... .. __ .... . ... _" __ _ __ _ __________ 

H2 Deflag/ 1. Gas/vent./2 HEPA 38 Intermediate 1 E-6 Yes 2E-8 

HLW Tank Heel 2. Cell/vent (first HEPA rendered 

3. Sensor/N2 inject ineffective) 

H2 Deflag/ 1. Gas/vent./2 HEPA 2.5 Low 1 E-6 Yes 1.3E-9 

Cs Tank Heel 2. Cell/vent (first HEPA rendered 

3. Sensor/N2 inject ineffective) 

H2 LAW Tank 1. Gas/vent./2 HEPA 2.3 Low 1 E-6 Yes 1.2E-9 

Deflag/ 2. Cell/vent (first HEPA rendered 

Low H2 3. Sensor/N2 inject ineffective) 

H2/LAW Tank 1. Gas/vent./2 HEPA 115 High 1 E-6 Yes 6E-8 

Deflag/ 2. Cell/vent (first HEPA rendered 

High H2 3. Sensor/N2 inject ineffective) 

H2/HLW Tank 1. Gas/vent./2 HEPA 216 High 1E-6 Yes 1.1 E-7 

Deflag/ 2. Cell/vent (first HEPA rendered (4.3% of 

Hiah H2 13. Sensor/N2 iniect ineffective) ....... total)
Table 5: Summary of the Impact of Potential Controls at the Generic TWRS Facility (continued)

Potential Potential Mitigated Mitigated kely Mitgatedl Event Controls/tems Mitigated Consequence Consequence Likelihood Acceptable per Risk, Relied Upon to Receptor Category (Probability) New Part 70? year-i CS TFo r S afety at 100 m eters, rem ,,, . .. ... __ __ __,__- __, ___ 

1. Cell/vent/2 HEPA 48 High !E6 e 2.4E-8 N JI e~QI;;- ; FrIlt-iL~~ i 'rf H-imtý PPA ronruarrri el 

ineffective) 

Melter/Canister/ 1. Cell/vent/2 HEPA 15 Intermediate 1E-4 Further Analysis 7.5E-7 Cap Dispersal 2. Instrumentation (first HEPA rendered Necessary (30% of 

ineffective due to heat) Total) 

Melter/Steam 1. Cell/vent/2 HEPA 26 Intermediate to 1 E-4 Further Analysis 1.3E-6 

Explosion 2. Instrumentation (first HEPA rendered High Necessary (51% of 

ineffective due to blast Total) 

and heat) 

Ammonia Tank 1. Tank < ERPG-1 Low 1 E-6 Yes (0) 

Failure 2. Enclosure 

3. Detect/Sprays 

Nitric Acid 1. Tank < ERPG-1 Low 1 E-6 Yes (0) 

Tank Failure 2. Enclosure _ . .... 3. Detect/Sprayvs 

Total mitigated risk = 2.5E-6/yr

NUREG-1747 60
Table 6: Different Sources of Risk Limits

Risk Source/Basis Dose Equivalent, Frequency, Risk, 

Rem yr-1 yr-1 

Worker Limits 

Part 20, Worker 5 1 2E-3 

Limit 

Part 20, Typical 0.31 1 1.2E-4 

ALARA Value 

U.S. Worker (-) (-) 4.8E-5 

Average, All Causes 

Public Limits 

Part 20, Public Limit 0.1 1 5E-5 

Part 20, D&D and 0.025 1 1.3E-5 

Part 61, Public Limits 

Typical Public Values 

U.S. Average 0.350 1 1.8E-4 

Background 

Background Difference 0.500 1 2.5E-4 

between 

Denver and U.S. 

Average 

Average U.S. Public (NA) (NA) 2E-3 

Cancer Fatality Rate 

Average Public Dose <0.001 1< 5E-7 

from Commercial 

Nuclear Plant

Note: Radiological comparisons assume Linear No Threshold (LNT) theory, with risk factors of 

2,500 rem/fatality for workers and 2,000 rem/fatality for members of the public. These rates are 

kept constant, and not reduced for higher acute doses (e.g., 1,000 rem/fatality for individual, 

acute doses over 10 rem).

61 NUREG-1747
6. Use of optimistic desigrn assumptions and reliabilities.

7. Uncertainties and inconsistencies not addressed. 

8. Selection, validation an(i' justification of data, assumption, and models. 

9. Emphasis on active instead of passive mitigation and control. 

10. Organization of documents and information. 

11. Inconsistencies in hazards identification, SSC categorization, and design classes 

12. Relative lack of quantification. 

13. Presence of large tanks and inventories. 

14. Limited consideration of offnormal and unanticipated events, and interactions. 

In addition, the ISM process itself is used as a risk-based approach, without consideration of a 

basal set of codes, standards.-•, minimum performance requirements, or basic design criteria. 

Numerical results are being u,,sed in absolute terms for comparisons with dose and frequency limits. No consideration of uncertainties by either analysis or statistics (e.g., 90t' percentile) has 

been included in the work to date - the emphasis has been on expected, "best basis," or median 

(50t" percentile). This expresses greater credibility in the capability of current modeling 

techniques, the quality of the available parameters, and the level of detail available in the 

design (currently at a circa 15 percent level) than appears warranted at this time. As a result, it 

raises doubts that defense-in-depth is adequately followed and that the IROFSs are fully identified with this approach. 'The ISM approach as followed at TWRS-P does not appear to be 

consistent with the NRC's risk-informed, performance-based approach. 

NRC analyses indicate significant radioactive and chemical inventories at the proposed facility 

that translate into potentially high levels of risk unless prevention and mitigation methods are 

used to address accident scenarios. This implies that a significant number of IROFSs will be 

needed. Potential IROFSs include those typical for fuel cycle facilities such as HEPA filter 

banks. NRC analyses also show a significant distance effect. Thus, analyses conducted using NRC-like approaches of 100 meters or so to a public receptor (the facility fence line) result in 

higher estimated consequences and more IROFSs as compared to the DOE approach of using the Hanford site's security per meter. The latter corresponds to distances of 11,000 to 15,000 

meters and reduces the consequences by 100-1000 (due to dilution), resulting in fewer 

IROFSs. Given that personnel are currently onsite that would be considered 

as members of the public under NRC guidance and that future site conditions will shrink the 

security perimeter, a distance much shorter than the 11,000-15,000 range would seem to be 

needed for estimating potent al consequences to the public. 

Furthermore, changes in the contracting approach may significantly affect the design and 

hazard analyses. There will now be a minimum of three contractors or contractor teams 

involved (interim, design and construction, and operations). Thus, it is not clear how the 

expertise on design, technolocgy, and safety will be transferred from one contractor organization

NUREG-1747 62
to the other. The new contractors also may not have to follow the existing design and 

approach, and therefore, some or all of the preceding hazards analyses may no longer be 

relevant. The proposed new contract will incorporate significant incentive clauses for capital 

cost reductions. It remains to be seen how these will be balanced vis-a-vis operating 

requirements and safety issues.

63 NUREG-1747
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NUREG-1747 64
3.0 NRC AREAS OF REVIEW 

3.4 RADIATION SAFETY AND DOSE ASSESSMENT METHODOLOGY 

3.4.1 REGULATORY LIMITS FOR OCCUPATIONAL AND PUBLIC DOSES APPLICABLE 

TO TANK WASTE REM EDIATION SYSTEM-PRIVATIZATION 

This chapter presents a comparison of regulatory approach, dose limits, and monitoring 

thresholds stated in 10 CFR Part 20, "Standards for Protection Against Radiation," and 

10 CFR Part 835, "Occupational Radiation Protection." 

3.4.1.1 INTRODUCTION 

The purpose of this chapter is to discuss the similarities and differences between 10 CFR 

Part 20, "Standards for Protection Against Radiation" (which affects U.S. Nuclear Regulatory 

Commission (NRC) licensees), and 10 CFR Part 835, "Occupational Radiation Protection," 

(which affects U.S. Department of Energy (DOE) facilities), and provide a basis for future 

activities should a transition to regulatory oversight by the NRC take place at some future time. 

Such a transition could affect certain aspects of the contractor's Radiation Control Program. 

The chapter also explains some of the regulatory philosophy of the DOE Regulatory Unit (RU) 

that is responsible for oversight of the TWRS-P contractor and how this philosophy may differ 

from that of the NRC. 

It should be understood at the outset that the NRC and DOE numerical dose limits are identical 

for individuals who receive occupational dose, and another set of identical limits applies to 

members of the public. However, differences exist between 10 CFR Part 20 and 10 CFR Part 

835 with respect to definitions of the conditions under which monitoring is required and how 

persons and situations are classified in terms of the type of monitoring performed. 

Misunderstandings may sometimes arise through the usage of terms that are not defined in the 

regulations, but which are similar to, or have the appearance of formal definitions. These 

include worker, radiation worker, radiological worker (defined in Part 835 but not in Part 20), 

and occupationally exposed, among others. Tables are provided that list and compare the 

definitions contained in the two Parts. A discussion is undertaken of Occupational Dose Limits 

and Limits for Members of the Public as stated in Parts 20 and 835; it will be seen that the limits 

are nearly identical. A comparison of Monitoring Thresholds mandated by Parts 20 and 835 is 

also presented. 

The Radiological Exposure Standards Above Normal Background for the TWRS-P Project are 

derived from 10 CFR Part 835 and the DOE/Regulatory Unit (RU) Top Level Standards, and for 

this reason do not have a directly corresponding regulation in Part 20 with which to compare. 

These standards are discussed in Section 3.4.1.3. 

For the TWRS-P facility, in addition to the categories of general employee and radiological 

worker (defined in Part 835) and member of the public (defined in both Parts), a third category 

known as the co-located worker (CLW) has been defined within the 'Top-Level Radiological,

65 NUREG-1747
Nuclear, and Process Safety Standards and Principles for TWRS Privatization Contractors," DOE/RL-96-0006, Revision 1, July 1998. The CLW is defined in Section 3.4.1.6, Glossary, as 

the following: 

An individual within the Hanford Site, beyond the Contractor-controlled area, performing 

work for or in conjunction with DOE or utilizing other Hanford Site facilities. 

3.4.1.2 DEFINITIONS 

3.4.1.2.1 MEMBERS OF THE PUBLIC 

Definitions related to membl)-s of the public are listed in alphabetical sequence in Table 7. 

From the NRC (10 CFR Par: 20) perspective, an individual who is outside a licensed facility's restricted area would normal!Iy be considered a member of the public." In contrast, an 

individual outside a controlled area as defined in 10 CFR Part 835 may be either a member of the public or a CLW, althou(;ih the concept of CLW is not defined for all DOE facilities, and as 

stated above, is not containe;d in 10 CFR Part 835. If such an individual is employed in a 

neighboring facility where he or she is subject to receiving an occupational dose, that individual 

may be considered to be a CLW relative to the Contractor's facility, according to DOE. 

Otherwise, that individual may fall into one of three other categories discussed below under 

"Individuals Subject to Occupational Dose." 

Although both 10 CFR Part .20 and 10 CFR Part 835 contain regulations governing dose limits 

to members of the public, and although the definitions of "member of the public" are similar in 

both, the regulatory language differs in several respects. Part 20 of 10 CFR provides limits for 

members of the public in general, while 10 CFR Part 835 provides limits only for those 

members of the public entering a controlled area. Note that "controlled area" is defined in 10 

CFR Part 835 differently from 10 CFR Part 20, and that the definition of a "restricted area" in 10 

CFR Part 20 is very close to that of a "controlled area" in 10 CFR Part 835, although the 10 

CFR Part 20 definition of a "restricted area" has no direct parallel in 10 CFR Part 835. The 

dose limits associated with these definitions are provided in Section 3.4.1.3. 

3.4.1.2.2 INDIVIDUALS SIJUBJECT TO OCCUPATIONAL DOSE LIMITS 

While the definitions of "merriber of the public" are similar in both 10 CFR Part 20 and 10 CFR 

Part 835, a comparison of the regulations pertaining to occupational dose is not as clear. The 

definitions in each Part differ with respect to individuals employed as workers in a facility that 

utilizes radioactive material. This section discusses the contrasting definitions and regulatory approaches of 10 CFR Parts 20 and 835 with respect to occupational dose. Table 8 lists the definitions in alphabetical ordler and clearly shows where a definition in 10 CFR Part 20 has no 

equivalent in 10 CFR Part 835, and vice-versa. 

Because terms related to "radiation area" are contained in the definitions of Table 8, Section 

3.4.1.2.3 provides definitions of such terms in Table 9. 

35 An exception would be if an individual normally permitted to receive an occupational dose utilizes radioactive 

materials from the licensed facility outside its restricted area, but under the control of the licensee.

NUREG-1747 66
Table 7: Definitions Related to "Member of the Public"

TERM f Part 20 Definitions I Part 835 Definitions 

Controlled Area Controlled area means an area, outside of a Controlled area means any area to which access 

restricted area but inside the site boundary, is managed by or for DOE to protect individuals 

access to which can be limited by the licensee from exposure to radiation and/or radioactive 

for any reason. material. 

Member of the Member of the public means any individual Member of the public means an individual who is 

Public except when that individual is receiving an not a general employee. An individual is not a 

occupational dose. "member of the public" during any period in which 

the individual receives an occupational dose. 

Public Dose Public dose means the dose received by a (no corresponding definition) 

member of the public from exposure to radiation 

or radioactive material released by a licensee, 

or to any other source of radiation under the 

control of a licensee. Public dose does not 

include occupational dose or doses received 

from background radiation, from any medical 

administration the individual has received, from 

exposure to individuals administered radioactive 

material and released in accordance with 

§35.75, or from voluntary participation in 

medical research programs. 

Restricted Area Restricted area means an area, access to which (no corresponding definition) 

is limited by the licensee for the purpose of 

protecting individuals against undue risks from 

exposure to radiation and radioactive materials. 

Restricted area does not include areas used as 

residential quarters, but separate rooms in a 

residential building may be set apart as a 

restricted area. 

Site Boundary Site boundary means that line beyond which the (no corresponding definition) 

land or property is not owned, leased, or 

otherwise controlled by the licensee. 

Unrestricted Area Unrestricted area means an area, access to (no corresponding definition) 

which is neither limited nor controlled by the 

licensee. 

To summarize, the 10 CFR Part 20 regulations distinguish only between a "member of the 

public" (discussed above) and an individual who receives an "occupational dose." An individual 

may receive an occupational dose from radioactive materials present in the facility where the 

work is performed without being directly employed by that facility. 

In contrast, under the regulations in 10 CFR Part 835, an individual may be a "member of the 

public," an individual receiving an "occupational dose," a "general employee," or a "radiological 

worker." An individual need not be a general employee or a radiological worker in order to 

receive an occupational dose. As with the 10 CFR Part 20 regulations, an individual may 

receive an occupational dose from radioactive materials present in the facility where the work is 

performed without being directly employed by that facility. A person may receive an 

occupational dose even though he or she is not a radiological worker. 

The occupational dose limits are the same in both Parts. However, for the TWRS-P facility, a 

set of dose limits has been established for a series of "event probability ranges," as described 

in RLREG-98-18, Revision 0, "Regulatory Unit Position on Radiological Safety for Hanford Co67 NUREG-1747
Located Workers," Septemhber 16, 1998, Exhibit 1: Table 2-1, "Radiological Exposure Standards Above Normal BE3Eckground." Numerical limits are discussed in Section 3.4.1.3. 

In accordance with DOE/RL-96-0006, individuals designated as CLWs are permitted to accrue 

occupational doses that are greater than that permitted for members of the public. 

3.4.1.2.3 RADIATION AREA, HIGH RADIATION AREA, VERY HIGH RADIATION AREA 

The previous section provided definitions of terms related to "occupational dose." Some of the 

terms contained in the definitions in Table 8 require explanation. Table 9 in this section 

provides definitions of terms related to "radiation area" as defined in Table 8. 

3.4.1.2.4 CONTAMINATIONIr AND RADIOACTIVE MATERIAL 

Table 10 in this section captures additional definitions relevant to radiological protection contained in 10 CFR Part 2( and 10 CFR Part 835 regulations. These terms are listed because 

they are related to the means by which occupational radiation doses might be encountered in 

facilities utilizing radioactive materials. It also serves to point out some of the other differences 

in the two sets of regulations;. Both Parts provide definitions of "Airborne Radioactivity Area," 

but the remaining terms are defined only in Part 835. 

3.4.1.2.5 DEFINITIONS OF TERMS RELATED TO DOSE AND DOSE EQUIVALENT 

The definitions contained in this section are provided to remind the reader of the technical 

terminology that describes hI'w dose is measured. Specific terms have been assigned to 

differentiate the type of measurement to be quantified. A thorough understanding of these 

terms is not required in order to appreciate the differences in 10 CFR Part 20 versus 10 CFR 

Part 835 radiological protection standards. In Table 11, the definitions are given in a 

hierarchical format with each definition building on its antecedent. Note that in 10 CFR Parts 20 

and 835 there are small differences in the way these definitions are phrased, but that they are 

equivalent in all other respects. 

3.4.1.3 LIMITS 

Table 12 provides a comparison of the limits set by 10 CFR Part 20 and 10 CFR Part 835. 

Note that the two sets of limi!,S are effectively the same. 

As indicated in Table 12, the, annual dose limit for members of the public is the same in both 

Parts, namely 0.1 rem (0.001 sievert). As stated previously, 10 CFR Part 835 is concerned only with those members of the public who enter a controlled area, while Part 20 covers any member of the public. Part ,.0 of 10 CFR also provides a limit of 0.002 rem (0.02 millisievert) in 

any one hour for the dose in an unrestricted area. There is no corresponding definition in 10 

CFR Part 835. Note that while 10 CFR Part 835 is concerned with members of the public who 

enter a controlled area (as defined in 10 CFR Part 835), the situation is not analogous to 10 

CFR Part 20, since 10 CFR -art 20 concerns members of the public outside any restricted area 

(as defined in 10 CFR Part 20).

NUREG-1747 68
Table 8: Definitions Related to "Occupational Dose"

TERM J Part 20 Definitions I Part 835 Definitions 

Controlled Controlled area means an area, outside of a Controlled area means any area to which access is 

Area restricted area but inside the site boundary, managed by or for DOE to protect individuals from access to which can be limited by the licensee for exposure to radiation and/or radioactive material. _any reason. 

General (no corresponding definition) General employee means an individual who is either 

Employee a DOE or DOE contractor employee; an employee of 

a subcontractor to a DOE contractor; or an individual 

who performs work for or in conjunction with DOE or utilizes DOE facilities. 

Occupational Occupational dose means the dose received by an Occupational dose means an individual's ionizing Dose individual in the course of employment in which radiation dose (external and internal) as a result of 

the individual's assigned duties involve exposure that individual's work assignment. Occupational dose 

to radiation or to radioactive material from licensed does not include doses received as a medical patient 

and unlicensed sources of radiation, whether in the or doses resulting from background radiation or possession of the licensee or other person. participation as a subject in medical research 

Occupational dose does not include dose received programs. 

from background radiation, from any medical 

administration the individual has received, from 

exposure to individuals administered radioactive 

material and released in accordance with §35.75, 

from voluntary participation in medical research 

programs, or as a member of the public. 

Radiological (no corresponding definition) Radiological area means any area within a controlled 

Area area defined in this section as a "radiation area," 

"high radiation area," "very high radiation area," "contamination area," "high contamination area," or "airborne radioactivity area." 

Radiological (no corresponding definition) Radiological worker means a general employee Worker whose job assignment involves operation of radiation 

producing devices or working with radioactive 

materials, or who is likely to be routinely 

occupationally exposed above 0.1 rem (0.001 sievert) 

per year total effective dose equivalent. 

Restricted Area Restricted area means an area, access to which is (no corresponding definition) 

limited by the licensee for the purpose of 

protecting individuals against undue risks from 

exposure to radiation and radioactive materials. 

Restricted area does not include areas used as residential quarters, but separate rooms in a residential building may be set apart as a restricted area. 

Site Boundary Site boundary means that line beyond which the (no corresponding definition) land or property is not owned, leased, or otherwise 

controlled by the licensee. 

Unrestricted Unrestricted area means an area, access to which (no corresponding definition) 

Area is neither limited nor controlled by the licensee.

69 NUREG-1747
Table 9: Definitions Related to "Radiation Area" 

TERM J _ Part 20 Definitions T Part 835 Definitions 

Radiation Area Radiation area. mneans an area, accessible to Radiation area means any area accessible to 

individuals, in Miich radiation levels could result individuals in which radiation levels could result in 

in an individual receiving a dose equivalent in an individual receiving a deep dose equivalent in excess of 0.0I5 rem (0.05 mSv) in 1 hour at 30 excess of 0.005 rem (0.05 millisievert) in 1 hour at 

centimeters from the radiation source or from any 30 centimeters from the source or from any surface 

surface that the radiation penetrates, that the radiation penetrates. 

High Radiation High radiation area means an area, accessible to High radiation area means any area, accessible to Area individuals, in 'hich radiation levels from individuals, in which radiation levels could result in 

radiation sourc:Es external to the body could result an individual receiving a deep dose equivalent in in an individual receiving a dose equivalent in excess of 0.1 rem (0.001 sievert) in 1 hour at 30 excess of 0.1 rem (1 mSv) in 1 hour at 30 centimeters from the radiation source or from any centimeters frcorn the radiation source or 30 surface that the radiation penetrates. 

centimeters frmirn any surface that the radiation 

penetrates. 

Very High Very high radiation area means an area, Very high radiation area means any area Radiation Area accessible to ndividuals, in which radiation levels accessible to individuals in which radiation levels 

from radiation sources external to the body could could result in an individual receiving an absorbed 

result in an individual receiving an absorbed dose dose in excess of 500 rads (5 grays) in one hour at 

in excess of 5,10 rads (5 grays) in 1 hour at 1 1 meter from a radiation source or from any surface 

meter from a radiation source or 1 meter from that the radiation penetrates. 

any surface thal the radiation penetrates. (Note: 

At very high dcses received at high dose rates, 

units of absorbed dose (e.g., rads and grays) are appropriate, ratlher than units of dose equivalent 

(e.g., rems and sieverts)). 

Table 10: Definitions :Related to "Contamination" and "Radioactive Material" 

TERM j Part 20 Definitions Part 835 Definitions 

Airborne Airborne radioactivity area means a room, enclosure, or Airborne radioactivity area means any area, Radioactivity area in which airbomrn radioactive accessible to individuals, where: 

Area materials, composed wholly or partly of licensed material, (1) The concentration of airborne radioactivity, above 

exist in concentrationE; -- (1) In excess of the derived air natural background, exceeds or is likely to exceed the 

concentrations (DACe) specified in Appendix B, to derived air concentration (DAC) values listed in §§20.1001 - 20.2401, or (2) To such a degree that an Appendix A or Appendix C of this part; or (2) An individual present in tie area without respiratory individual present in the area without respiratory protective equipment could exceed, during the hours an protection could receive an intake exceeding 12 

individual is present in a week, an intake of 0.6 percent of DAC-hours in a week. 

the annual limit on intake (ALl) or 12 DAC-hours. 

Contamination (no corresponding definition) Contamination area means any area, accessible to Area individuals, where removable surface contamination 

levels exceed or are likely to exceed the removable 

surface contamination values specified in Appendix D 

of this part, but do not exceed 100 times those 

values. 

High (no corresponding de.irition) High contamination area means any area, accessible 

Contamination to individuals, where removable surface Area contamination levels exceed or are likely to exceed 

100 times the removable surface contamination 

values specified in Appendix D of this part. 

Radioactive (no corresponding deirition) Radioactive material area means any area within a Material Area controlled area, accessible to individuals, in which 

items or containers of radioactive material exist and 

the total activity of radioactive material exceeds the applicable values provided in Appendix E of this part.

NUREG-1747 70
Table 11: Definitions Related to "Dose"

TERM J Part 20 Definition I Part 835 Definitions 

Dose Dose or radiation dose is a generic term that Dose is a general term for absorbed dose, dose 

means absorbed dose, dose equivalent, equivalent, effective dose equivalent, committed dose 

effective equivalent, committed effective dose equivalent, or dose equivalent, committed dose equivalent, total effective dose equivalent as defined in this part. 

committed effective dose equivalent, or total 

effective dose equivalent, as defined in other 

paragraphs of this section. 

Dose Equivalent Dose equivalent (HT) means the product of Dose equivalent (H) means the product of absorbed 

the absorbed dose in tissue, quality factor, dose (D) in rad (or gray) in tissue, a quality factor (Q), 

H1- and all other necessary modifying factors at and other modifying factors (N). Dose equivalent is 

the location of interest. The units of dose expressed in units of rem (or sievert) (1 rem = 0.01 

equivalent are the rem and sievert (Sv). sievert). 

Effective Dose Effective dose equivalent (HE) is the sum of Effective dose equivalent (HE) means the summation 

Equivalent the products of the dose equivalent to the of the products of the dose equivalent received by 

organ or tissue (HT) and the weighting factors specified tissues of the body (HT) and the appropriate 

HE (wT) applicable to each of the body organs or weighting factor (wT)-that is, HE = I WT HT. It includes 

tissues that are irradiated (HE = I wT HT). the dose from radiation sources internal and/or 

external to the body. For purposes of compliance with 

this part, deep dose equivalent to the whole body may 

be used as effective dose equivalent for external 

exposures. The effective dose equivalent is expressed 

in units of rem (or sievert). 

Committed Dose Committed dose equivalent (HT,11) means the Committed dose equivalent (HT.-I) means the dose 

Equivalent dose equivalent to organs or tissues of equivalent calculated to be received by a tissue or reference (T) that will be received from an organ over a 50-year period after the intake of a 

HT,50 intake of radioactive material by an individual radionuclide into the body. It does not include 

during the 50-year period following the intake, contributions from radiation sources external to the 

body. Committed dose equivalent is expressed in units 

of rem (or sievert). 

Committed Committed effective dose equivalent (HE.W) is Committed effective dose equivalent (HE.50) means the 

Effective Dose the sum of the products of the weighting sum of the committed dose equivalents to various 

Equivalent factors applicable to each of the body organs tissues in the body (HT,ý), each multiplied by the or tissues that are irradiated and the appropriate weighting factor (wT)--that is, HE. 2 = 1 WT 

HE,5o committed dose equivalent to these organs or HT.1 . Committed effective dose equivalent is 

tissues. expressed in units of rem (or sievert). 

Total Effective Total Effective Dose Equivalent (TEDE) Total effective dose equivalent (TEDE) means the 

Dose Equivalent means the sum of the deep-dose equivalent sum of the effective dose equivalent (for external 

(for external exposures) and the committed exposures) and the committed effective dose 

TEDE effective dose equivalent (for internal equivalent (for internal exposures). 

exposures).

71 NUREG-1747
Table 12: Occupal:ional Dose Limits and Limits for Members of the Public 

Category _ IPart 20 Dose Limits Part 835 Dose Limits 

Occupational (1) An annual limnit, which is the more limiting of: (1) A total effective dose equivalent of 5 rems (0.05 sievert); (Total Effective (i) The total effective dose equivalent being equal Dose Equivalent to 5 rems (0.05 Sv); or (2) The sum of the deep dose equivalent for 

external exposures and the committed dose and (ii) The sum ci :he deep-dose equivalent and the equivalent to any organ or tissue other than the 

committed do,;e equivalent to any individual lens of the eye of 50 rems (0.5 sievert); Organ or Tissue) organ or tissue other than the lens of the eye 

being equal to 50 rems (0.5 Sv). 

Occupational (2) The annuad limits to the lens of the eye, to the (3) A lens of the eye dose equivalent of 15 rems skin, and to the extremities, which are: (0.15 sievert); and 

(Lens of Eye, Skin, Extremities) (i) A lens dose equivalent of 15 rems (0.15 Sv), (4) A shallow dose equivalent of 50 rems (0.5 and sievert) to the skin or to any extremity. 

(ii) A shallow-iose equivalent of 50 rems (0.50 

Sv) to the skin or to any extremity. 

Public 0.1 rem (1 millisievert) 0.1 rem (0.001 sievert) 

0.002 rem (0.02 millisievert) in any one hour for (no corresponding definition) 

the dose in anr Unrestricted area 

A more detailed explanation of the dose limits applied to members of the public and to 

individuals subject to occupational dose, which includes CLWs, needs to include a review of the 

four event probability range.S, as described in RL/REG-98-18, Revision 0, "Regulatory Unit 

Position on Radiological Safety for Hanford Co-Located Workers." These are reproduced here 

as Table 13, Radiological Exposure Standards Above Normal Background. 

Table 13 specifies four event probability ranges addressing normal operation and credible 

accident conditions. General guidelines and frequencies listed for the four event probability 

ranges are: 

1. Normal events are typical of normal facility operations expected to occur regularly in the 

course of facility operations; the associated frequency of occurrence during the lifetime of 

the facility is 1 or more per year. As defined in Table 13, a general guideline for this event 

probability is that normal modes of operating the facility systems should provide adequate 

protection of health and safety. 

2. Anticipated events are characterized as minor incidents and upsets of moderate frequency that may occur once or more during the lifetime of the facility; the associated probability 

range is lx1 0-2 to <1 per year. As defined in Table 13, a general guideline for this event 

probability range is that the facility should be capable of returning to operation without 

extensive corrective action or repair. 

3. Unlikely events are characterized as more severe incidents that are not expected, but may 

occur, during the lifetime of the facility; the associated probability range is lx1i04 to lx1 0.2

NUREG-1747 72
per year. As defined in Table 13, a general guideline for this event probability range is that 

the facility should be capable of returning to operation following potentially extensive 

corrective action or repair, as necessary. 

4. Extremely unlikely events are characterized as events that are not expected to occur during 

the lifetime of the facility, but are postulated because their consequences would include the 

potential for the release of significant amounts of radioactive material; the associated 

probability range is lx10i6 to lx10' per year. As defined in Table 13, a general guideline for 

this event probability range is that facility damage may preclude returning to operation. 

Note that a probability of occurrence of 1x10 2 per year is equivalent to a frequency of one 

occurrence in 100 years; 1x10 4 per year equates to one in 10,000 years; and 1x10' per year 

equates to one in 1,000,000 years. These time periods are also known as recurrence intervals. 

The additional category of worker has implications with regard to the design and construction of 

structures, systems, and components (SSCs) of the TWRS-P facility, but does not establish a 

different set of radiation exposure standards (limits) or a different threshold for monitoring these 

workers. As may be observed by review of Table 13, for all categories of events, the limits are 

the same for facility workers and for co-located workers (with the exception that facility workers 

are subject to the additional limits of _<50 rem/yr to any organ, the skin, or extremity, and 

•15 rem/yr to the lens of the eye). 

The benefit of using the co-located worker category is that it allows the TWRS-P facility to treat 

these workers as individuals subject to occupational dose, rather than as members of the public 

under accident conditions. Therefore, in the design and construction of TWRS-P facility SSCs, 

the "offsite" consequences (i.e., the radiation doses to individuals beyond the boundaries of the 

TWRS-P facility) are permitted to be greater for the nearby co-located workers than would be 

the case if these personnel were considered to be members of the public 

For the category of normal events, which have a frequency of occurrence of one or more per 

year, and are considered to have the lowest consequence in terms of harm, co-located workers 

will seldom, if ever, receive any additional radiation exposure from the TWRS-P facility. 

However, operation under non-normal conditions, i.e., with releases above ALARA (as low as 

reasonably achievable) levels, may add to the exposure received by co-located workers. It 

should be noted that co-located workers will continue to be subject to the same limits as the 

worker, so that any exposure received due to operation of TWRS-P (under non-normal 

conditions) is added to that received from the duties performed in the facility in which their 

occupational exposure is normally received. 

DOE regulations pertaining to occupational radiation protection (Part 835) provide some 

specificity beyond that found in NRC regulations (Part 20), due to the inclusion of certain 

definitions in Part 835, including those for "general employee" and "radiological worker."

73 NUREG-1747
Table 13: Raoiiclogical Exposure Standards Above Normal Background

EstimatedI 

Description Frequency Occurrence cf General Guidelines Worker CWoLocated Public Worker 

f(yr"1 ) 

Normal Events: >0.1 Normal modes of <5 rem/yr < 5 rem/yr < 10 mrem/yr Events that occur operating facility _< 50 rem/yr any organ, < 1.0 rem/yr (airborne regularly in the course systems should skin, or extremity ALARA design pathway) of facility operation provide adequate < 15 rem/yr lens of eye objective per < 100 mrem/yr (e.g., normal -facility protection of health < 1.0 rem/yr ALARA 10 CFR (all sources) operations); including and safety. design objective per 835.1002(b)(1 ) < 100 mrem/yr routine and preventive 10CFR835.1002 (b) T (public in the 

maintenance controlled area) activities. < 25 mrem/yr 

Tradioactive 

waste) 

Anticipated Events: 10 2 <f < 10 The facility should be < 5 rem/event 2.2 < 5 rem/event 12.3 _< 100 

Events of moderate capable of returning to 1.0 rem/event design 1.0 rem/event mrem/event # 

frequency that may operation without action threshold (4) design action occur once or more extensive corrective threshold (4) 

during the life of a action or repair. 

facility (e.g., minor 

incidents and upsets). 

Unlikely Events: 10-4 <f < 10" The facility should be <25 rem/ event(23) <25 rem/event (23) <5rem/event(3) 

Events that are not capable of returning to 

expected, but may operation following occur during the potentially extensive 

lifetime of a facility corrective action or (e.g., more severe repair, as necessary. 

incidents). 

Extremely Unlikely 10 -6 <f _< 10- Facility damage may <25 rem/ event (2.3) < 25 rem/event (2,3) < 25 rem/event 

Events: preclude returning to < 5 rem/event 

Events that are not operation. target c) 

expected to occur < 300 rem/event 

*during the life of the to thyroid 

facility but are 

postulated because 

their consequences 

would include the 

potential for the 

release of significant 

amounts of

vvWihrin Iie 01DI-L 

TWRS-P Controlled 

Area Boundary, 

including 241-AP-106

Tne most imiting 

location at or 

beyond the BNFL 

TWRS-P Controlled 

Area Boundary

I ne most limiting 

location along the 

near river 

bank/Hwy 240 

/southem 

boundary

(1) In addition to meeting the listed design objective of 10 CFR 835.1002(b), the inhalation of radioactive material by workers and co-located workers under normal coiditions is kept ALARA through the control of airborne radioactivity as described in 10 

CFR 835.1002(c). 

(2) In addition to meeting the listed worker and co-located worker exposure standards for accidents, the Worker Accident Risk 

Goal is satisfied through the calculal ion of the risk from accidents with accident prevention and mitigation features added as necessary to meet the Goal. 

(3) In addition to meeting the listed exposure standards for accidents, BNFL Inc.'s approach to accident mitigation is to evaluate 

accident consequences to ensure that the calculated exposures are far enough below standards to account for uncertainties in 

the analysis and to provide for suffiaient design margin and operational flexibility. 

(4) When a calculated accident exposure exceeds this threshold, then appropriate actions are taken. These include carrying out a less bounding (i.e., more realistic) e"aluation to show that the accident consequences will be below the threshold or evaluating additional safeguards for cost-effectiveness and/or feasibility. This threshold is not a limit; it does not require the implementation of additional preventive or mitigative features if they are not both cost-effective and feasible.

NUREG-1747 74

I

V I
The importance of these additional definitions in the regulation of the TWRS-P facility is 

explained in RL/REG-98-18, Revision 0, "Regulatory Unit Position on Radiological Safety for 

Hanford Co-Located Workers." This document points out that a person can be occupationally 

exposed even though he or she is not a radiological worker. 

A document which explains the use of the categories of "Facility Worker," "Co-Located 

Worker," and "Public" is BNFL-5193-RES-01, Rev. 0, August 28,1997, "Radiological and 

Nuclear Exposure Standards for Facility and Co-Located Workers." These terms are defined 

in Section 3.4.1.3, "Development of the BNFL Approach to Compliance with Table 1 of 

DOE/RL-96-0006." 

3.4.1.4 MONITORING THRESHOLDS 

Generally speaking, the NRC requires monitoring of adults likely to receive in excess of 

10 percent of the occupational dose limits, while the DOE requires monitoring of adults likely to 

receive in excess of 2 percent of the occupational dose limits. For whole body exposure, these 

equate to 500 mrem and 100 mrem, respectively. 

Table 14, "Monitoring Thresholds," provides information from 10 CFR Part 20 and 10 CFR 

Part 835 relevant to the monitoring of individuals subject to occupational dose and members of 

the public. These Parts also contain information about monitoring exposure to minors and 

declared pregnant women (in 10 CFR Part 20) or declared pregnant workers (in 10 CFR 

Part 835). 

3.4.2 RADIATION PROTECTION PROGRAM 

This section presents a comparison of regulatory requirements and guidance of the NRC and 

the DOE with regard to the criteria for Radiation Protection Programs as described in 10 CFR 

Part 20, Standards for Protection Against Radiation and 10 CFR Part 835, Occupational 

Radiation Protection. 

3.4.2.1 INTRODUCTION 

Should a transition to regulatory oversight by the NRC take place at some future time, certain 

aspects of the contractor's Radiation Control Program could be affected. The objectives of this 

section are to: 

1. Describe the contractor's Radiation Protection Program (RPP) and the regulatory basis for 

requiring its use. 

2. Provide an explanation of differences in the regulatory approach of the two agencies with 

respect to this programmatic requirement. 

3. Identify, insofar as possible, the extent to which the Contractor's RPP would be affected 

by a transition to NRC regulatory authority.

75 NUREG-1747
Table 14: Monitoring Thresholds

Section in P I:art 20 1 Part 835 

10 CFR: I 

10 CFR 20.1502 Conditions 10 CFR 835.402 Individual monitoring. 

requiring individual monitoring of 

external and internal occupational 

dose. 

Each licensee shall monitor (no corresponding text) 

exposures to radiation and radioactive 

material at levels sufficient to 

demonstrate cornpliance with the 

occupational dose limits of this part. 

As a minimum 

External (a) Each licensee shall monitor (a) For the purpose of monitoring individual 

occupational exposure to radiation exposures to external radiation, personnel from licensed ard unlicensed dosimeters shall be provided to and used by: 

radiation sources; under the control of 

the licensee and shall supply and (1) Radiological workers who, under typical require the use of individual conditions, are likely to receive one or more of 

monitoring devices by-- the following: 

(i) An effective dose equivalent to the whole 

(1) Adults likely lo receive, in 1 year body of 0.1 rem (0.001 sievert) or more in a 

from sources exl:ernal to the body, a year; 

dose in excess of 10 percent of the (ii) A shallow dose equivalent to the skin or to 

limits in §20.120"1{a), any extremity of 5 rems (0.05 sievert) or.more 

in a year; 

(2) Minors likely to receive, in 1 year, (iii) A lens of the eye dose equivalent of 1.5 

from radiation sources external to the rems (0.015 sievert) or more in a year; 

body, a deep dose equivalent in 

excess of 0.1 remn (1 mSv), a lens (2) Declared pregnant workers who are likely dose equivalent ir excess of 0.15 rem to receive from external sources a dose 

(1.5 mSv), or a shallow dose equivalent to the embryo/fetus in excess of 10 

equivalent to the skin or to the percent of the limit at Sec. 835.206(a); 

extremities in exc:ess of 0.5 rem (5 

mSv); (3) Occupationally exposed minors likely to 

receive a dose in excess of 50 percent of the 

(3) Declared pregnant women likely to applicable limits at Sec. 835.207 in a year from 

receive during the entire pregnancy, external sources; 

from radiation scurces external to the 

body, a deep dose equivalent in (4) Members of the public entering a controlled 

excess of 0.1 tern (1 mSv);2 and area likely to receive a dose in excess of 50 

percent of the limit at Sec. 835.208 in a 

(4) Individuals entering a high or very year from external sources; and 

high radiation area. 

(5) Individuals entering a high or very high 2 All of the occupational doses in radiation area. 

§20.1201 continue to be applicable to 

the declared preg nant worker as long 

as the embryo/fetus dose limit is not

NUREG-1747 76
Section in Part 20 1 Part 835 

10 CFR:_I . 

exceeded. 

Internal (b) Each licensee shall monitor the (c) For the purpose of monitoring individual 

occupational intake of radioactive exposures to internal radiation, internal 

material by and assess the committed dosimetry programs (including routine 

effective dose equivalent to -- bioassay programs) shall be conducted for: 

(1) Adults likely to receive, in 1 year, (1) Radiological workers who, under typical 

an intake in excess of 10 percent of conditions, are likely to receive a committed 

the applicable ALl(s) in Table 1, effective dose equivalent of 0.1 rem (0.001 

Columns 1 and 2, of Appendix B to sievert) or more from all occupational 

§§20.1001-20.2400; radionuclide intakes in a year; 

[see note in last row of this table] 

(2) Declared pregnant workers likely to receive 

(2) Minors likely to receive, in 1 year, an intake or intakes resulting in a dose 

a committed effective dose equivalent equivalent to the embryo/fetus in excess of 10 

in excess of 0.1 rem (1 mSv); and percent of the limit stated at Sec. 835.206(a); 

(3) Occupationally exposed minors who are 

(3) Declared pregnant women likely to likely to receive a dose in excess of 50 percent 

receive, during the entire pregnancy, a of the applicable limit stated at Sec. 835.207 

committed effective dose equivalent in from all radionuclide intakes in a year; or 

excess of 0.1 rem (1 mSv). 

(4) Members of the public entering a controlled 

(no corresponding text) area likely to receive a dose in excess of 50 

percent of the limit stated at Sec. 835.208 from 

all radionuclide intakes in a year. 

[Note: 1 ALl = 5 rem, or 5000 mrem 

for whole body exposure; therefore 

10% of the limit (1 ALl) = 500 mrem; 

also 1 ALl = 50 rem, or 50 000 mrem 

for the maximally exposed organ; 

therefore 10% of the limit (1 ALl) = 5 

rem = 5000 mrem] 

Table 14: Monitoring Thresholds 

(Continued) 

The acronym RPP is sometimes used to refer to both "River Protection Project" and to 

"Radiation Protection Program;" however, in this section, it will be used only in the latter sense, 

i.e., to refer to the Radiation Protection Program. 

3.4.2.2 DESCRIPTION OF THE RADIATION PROTECTION PROGRAM 

A document prepared by the former contractor (BNFL Inc.) entitled "Radiation Protection 

Program for Design" (BNFL-TWP-SER-003) is the radiation program document for achieving 

compliance with the Code of Federal Regulations (CFR) Part 835, "Occupational Radiation

77 NUREG-1747
Protection." As of November 15, 1999, BNFL Inc. had submitted the third revision to this 

document to the DOE-RU. In a letter from BNFL Inc. to the RU dated January 31, 2000, BNFL 

Inc. indicated its intention to submit Revision 4 of this document, to be known as the RPP for 

Design and Limited Construction, on or before May 1, 2000. This was subsequently delayed to 

June 30, 2000. Upon submittal, it was subjected to an acceptance review by the RU, which 

rejected it for a number of inadequacies. 

The Introduction to BNFL-TWP-SER-003 provides the following description of the RPP: 

The RPP is developed and submitted for regulatory approval in stages corresponding to 

the status of the [TWRS..P] Project. This initial RPP submittal describes the plans and 

measures for achieving c'ompliance with the requirements of 10 CFR [Part] 835 that are 

applicable to the [TWRS-P] Facility design phase. No radiological source term will exist 

to cause personnel exposures during the design phase; however, decisions made 

during the design phase will affect exposures during facility operations and deactivation. 

3.4.2.3 REGULATORY BASIS 

In contrast to 10 CFR Part 20, the regulations in 10 CFR 835.101 are very prescriptive with 

respect to the requirement for, a radiation protection program. While both agencies require the 

implementation of an RPP, the DOE includes more stringent requirements such as the 

following: 

1. 10 CFR 835.101(a) requires approval by DOE. 

2. 10 CFR 835.101(b) gives authority to DOE to direct or make modifications to a RPP. 

3. 10 CFR 835.101(e) requires the RPP to address, but not necessarily be limited to, each 

requirement in Part 835. 

The NRC requires the develhpment, documentation, and implementation of a radiation 

protection program by its licensees (10 CFR 20.1101(a)), but the regulation does not require the licensee to submit this document for approval by the Agency, and does not give the NRC 

the direct authority to require modifications of the RPP. 

The relevant sections of 10 CIFR 20.1101 and 10 CFR 835.101 are provided in Table 15. This 

table indicates that there are few corresponding sections in the two Parts. There are similarities 

between 10 CFR 20.1101 (a) and 10 CFR 835.101 (a) and also between 10 CFR 20.1101 (c) and 

10 CFR 835.101 (g), but, otherwise, the two regulations are quite different. The related 

recordkeeping requirements of 10 CFR 20.2102 have some similarities to portions of 10 CFR 

835.701 and 10 CFR 835.704, as shown in Table 16.

NUREG-1747 78
Table 15: Radiation Protection Plan Requirements

Part 20 f Part 835 

Subpart B - Radiation Protection Programs Subpart B - Management and Administrative Requirements 

10 CFR 20.1101 Radiation protection programs 10 CFR 835.101 Radiation protection programs 

(a) Each licensee shall develop, document, and (a) A DOE activity shall be conducted in compliance with a documented 

implement a radiation protection program radiation protection program (RPP) as approved by the DOE. 

commensurate with the scope and extent of 

licensed activities and sufficient to ensure 

compliance with the provisions of this part. (See 10 

CFR20.2102 for record- keeping requirements 

relating to these programs.) 

(b) The licensee shall use, to the extent practical, (no corresponding regulation) 

procedures and engineering controls based upon 

sound radiation protection principles to achieve 

occupational doses and doses to members of the 

public that are as low as is reasonably achievable 

(ALARA). 

(no corresponding regulation) (b) The DOE may direct or make modifications to a RPP. 

(no corresponding regulation) (c) The content of each RPP shall be commensurate with the nature of 

the activities performed and shall include formal plans and measures for 

applying the as low as reasonably achievable (ALARA) process to 

occupational exposure. 

(no corresponding regulation) (d) The RPP shall specify the existing and/or anticipated operational 

tasks that are intended to be within the scope of the RPP. Except as 

provided in Sec. 835.101 (h), any task outside the scope of a RPP shall 

not be initiated until an update of the RPP is approved by DOE. 

(no corresponding regulation) (e) The content of the RPP shall address, but shall not necessarily be 

limited to, each requirement in this part. 

(no corresponding regulation) (f) The RPP shall include plans, schedules, and other measures for 

achieving compliance with regulations of this part. Unless otherwise 

specified in this part, compliance with amendments to this part shall be 

achieved no later than 180 days following approval of the revised RPP 

by DOE. Compliance with the requirements of Sec. 835.402(d) for 

radiobio- assay program accreditation shall be achieved no later than 

January 1,2002. 

(c) The licensee shall periodically (at least annually) (g) An update of the RPP shall be submitted to DOE: 

review the radiation protection program content and (1) Whenever a change or an addition to the RPP is made; 

implementation. (2) Prior to the initiation of a task not within the scope of the RPP; or 

(3) Within 180 days of the effective date of any modifications to this part. 

(no corresponding regulation) (h) Changes, additions, or updates to the RPP may become effective 

without prior Department approval only if the changes do not decrease 

the effectiveness of the RPP and the RPP, as changed, continues to 

meet the requirements of this part. Proposed changes that decrease the 

effectiveness of the RPP shall not be implemented without submittal to 

and approval by the Department. 

(no corresponding regulation) (i) An initial RPP or an update shall be considered approved 180 days 

after its submission unless rejected by DOE at an earlier date.

79 NUREG-1747
"Table 16: Recordkeeping Requirements

Part220 - Part 83S 

Subpart L - Records Subpart H - Records 

10 CFR 20.2102 Records of Radiation Protection 10 CFR 835.701 General Provisions 

programs 

(a) Each licensee shall maintain reco'ds of the radiation (a) Records shall be maintained to document compliance with protection program, including: this part and with radiation protection programs required by 10 

(1) The provisions of the program; and CFR 835.101. 

(2) Audits and other reviews of progr.irn content and 

implementation. (b) Unless otherwise specified in this subpart, records shall be (b) The licensee shall retain the records required by retained until final disposition is authorized by DOE. 

paragraph (a)(1) of this section until the Commission 

terminates each pertinent license reclu ring the record. The 

licensee shall retain the records required by paragraph (a)(2) 

of this section for 3 years after the record is made. 

10 CFR 835.704 Administrative records 

(no corresponding regulation) (b) Actions taken to maintain occupational exposures as low as reasonably achievable, including the actions required for this 

purpose by Sec. 835.101, as well as facility design and control 

actions required by Secs. 835.1001, 835.1002, and 835.1003, 

shall be documented. 

See 10 CFR20.2102(a)(2), above. (c) Records shall be maintained to document the results of 

internal audits and other reviews of program content and 

implementation. 

As stated previously, the 10 CFR Part 20 (NRC) requirements for an applicant's or licensee's 

RPP are less prescriptive thain the 10 CFR Part 835 (DOE) requirements. If regulatory authority for the TWRS-P Proj 'ect were to transition from DOE to NRC, in the process of 

replacing references to DOE. and 10 CFR Part 835 with references to NRC and 10 CFR 

Parts 20 and 70, the contracl:or may want to use the opportunity to remove requirements that 

are not part of the NRC reg ilations. This could be problematical, given that changes that are 

judged to reduce the effectivweness of the RPP may not be acceptable. 

If the NRC were to assume regulatory authority at an early stage, i.e., prior to construction 

and/or operation, the content and format of the RPP would be reviewed in the context of a 

Preliminary/Final Safety Analysis Report (or equivalent), in accordance with the NRC's 

Standard Review Plan (SRP - see Section 3.4.2.6). In any case, the contractor's RPP would be 

reviewed within the scope of the NRC's inspection program to determine whether it is in 

compliance with 10 CFR 20.1 101 a, "Radiation protection programs," which states: 

Each licensee shall develop, document, and implement a radiation protection program 

commensurate with the scope and extent of licensed activities and sufficient to ensure 

compliance with the provisions of this part. 

For the TWRS-P facility, in addition to the categories of general employee and radiological worker (defined in Part 835) and member of the public (defined both in Part 835 and in Part 20), 

a third category known as the co-located worker (CLW), has been defined within the 'Top-Level Radiological, Nuclear, and P'ocess Safety Standards and Principles for TWRS Privatization

NUREG-1747 80
Contractors," DOE/RL-96-0006, Revision 1, July 1998. Within this document, the co-located 

worker is defined in Section 6.0, Glossary, as the following: 

An individual within the Hanford Site, beyond the Contractor-controlled area, performing 

work for or in conjunction with DOE or utilizing other Hanford Site facilities. 

The co-located worker is referenced in the Introduction to Section IV of the DOE 

Implementation Guide G-10 CFR 835/B2, "Occupational ALARA Program" (DOE 1994), which 

states: 

This section gives the basic guidelines for conducting an operational ALARA program. 

It includes the requirements and guidance for developing, implementing, and 

documenting the program to reduce doses to workers and "co-located workers" to levels 

that are ALARA. 

Currently no plans exist regarding the change in status, if any, of the CLW category if a 

transition to NRC regulatory authority were to occur. 

3.4.2.4. ORGANIZATION OF THE RPP FOR DESIGN 

The first four sections of the RPP are the "Introduction," "RPP Document Organization," 

"Purpose," and "Applicability." Most of the relevant information in these sections has been 

summarized in the introduction to this section (3.4.2). The major content of the RPP is 

contained in RPP Section 5, which addresses "... each functional element of 10 CFR Part 835 

applicable during the RPP-WTP design phase" (Section 5.0, Paragraph 4). RPP Section 5.1 

covers "Maintenance of the RPP." Sections 5.2, 5.3, and 5.4 discuss Audits, Management of 

Records, and Training, respectively, which are required to demonstrate compliance with 10 

CFR Part 835. 

RPP Section 5.5, "BNFL Inc. Application of ALARA to the RPP-WTP Facility Design," is 

comprised of 14 subsections that are intended to describe how the contractor will apply ALARA 

to the facility design. Two of the fourteen RPP subsections were deemed inapplicable to the 

design phase: 5.5.4, "Radiological Performance Goals/Indicators," and 5.5.13, "Radiological 

Work/Experiment Planning." 

RPP Subsections 5.5.1 through 5.5.14 are summarized in Table 17. These are based on the 

essential elements in Section IV of the DOE Implementation Guide G-10 CFR 835/B2, 

"Occupational ALARA Program" (DOE 1994). The essential element numbers are shown in 

boldface type below the corresponding subsection numbers. 

The first paragraph of RPP Section 5.5 states "the form and content of this section of the RPP 

is consistent with" the DOE Implementation Guide. The information in this Implementation 

Guide is similar to that found in NRC Regulatory Guides 8.8 and 8.10. Although a 

comprehensive intercomparison of regulatory guidance is not the intention of this summary, the 

following chapter provides an indication of similarities and differences of approach between the 

NRC and DOE.

81 NUREG-1747
The RPP for Design concludes with a Bibliography (Section 6.0) and an Appendix of nearly 40 pages that lists in a tabular format 

each section of Part 835 and its requirements, accompanied by the BNFL Inc. Plans and Measures for Achieving Compliance. If a particular section of Part 835 is not applicable at the design stage, or if no actual requirement is stated, this is so indicated in the 

table. Most plans and measures are described by reference to the appropriate part of the text of the RPP. 

Table 17: Summary of Application of ALARA to the Facility Design 

RPPl 

RPP Section Section Title Summary ..... of Section Contents 

5,5. BNFL Inc. Application of ALARA to the Facility Design 

.. ,I 

tI)- management general punbic. 

Commitment 

5.5.2. Organization and Project General Manager 

(2) Responsibilities Project Manager 

Project Engineering Manager 

Project Quality Assurance Manager 

Project Operations Manager (Design Phase) 

Project Safety and Regulatory Programs Manager 

Project Safety Committee 

ALARA Subcommittee (ASC) 

5.5.3. Administrative "Administrative Control Levels," are described in DOE Implementation Guide G-10 CFR Part 835/B2, "Occupational ALARA Program" (3) Control Levels (DOE 1994) in relation to occupational exposure and are, therefore, not applicable to this RPP. 

5.5.4. Radiological This RPP reflects the design phase of the Project only. Radiological Performance Goals/Indicators are not applicable to this RPP. 

(4) Performance 

Goals/ Indicators 

5.5.5. ALARA Training Specific technical training (including ALARA training) shall be planned, scheduled, provided, documented, and maintained for personnel (5) in their respective technical disciplines as defined by position descriptions and specific work assignments. A training matrix shall be 

maintained for defining and tracking training requirements, 

5.5.6. Plans and Plans and procedures for ALARA shall be commensurate with the activities authorized to be performed under this RPP. 

(6) Procedures 

5.5.7. Internal Perform internal audits of all functional elements of the radiological control program. Such audits shall ensure that each functional 

(7) Assessments/ element, including ALARA, Is audited formally no less frequently than every 3 years. 

Audits 

*Numbers in boldface type refer to the corresponding essential elements in DOE IG G-1 0 CFR Part 835/12, Occupational ALARA Program

NUREG-1747 82
Table 17: Summary of Application of ALARA to the Facility Design (continued) 

R P P . . . ..•... Section Section Title Summary of Section Contents 

5.5.8. Optimization Methodology 

(8) 

5.5.8,1. Process It is not expected that cost benefit analysis (CBA) will be used as the primary driver in every ALARA decision. CBA is viewed as one of the 

Description inputs in an ALARA analysis, and depending on the particulars of the situation (the formality and degree of quantitative analysis should 

reflect the scale and type of problem under consideration), CBA may not be required in order to arrive at an appropriate decision. 

5.5,8.2. Applied Value of A key input parameter for CBA is the value of the person-rem detriment. BNFL Inc. policy is based on the United Kingdom National Radiation 

Protection for Protection Board (NRPB) guidance, together with an adjustment for United States commercial and DOE practices. 

Optimization 

5.5.9. ALARA Design Process 

(9) 

5.5.9.1. Overview The ALARA design process applied in all stages of design. Experience has shown that the greatest potential for significant dose savings at 

the lowest cost is achieved at the earliest stages of design. 

5,5.9.2. Hierarchy of ALARA design process uses a hierarchy of controls giving priority to those controls that are most effective. 

Protection 

5.5.10. ALARA Design ALARA design criteria applied throughout the design of the facility: 

Criteria 5.5.10.1. Primary method - physical design features (e.g., confinement, ventilation, remote handling, and shielding). 

5.5.10.2. Administrative controls - employed only as supplemental methods to control radiation exposure. 

5.5.10.3. Physical design features for specific activities demonstrated to be impractical - use administrative controls, 

5.5.10.4. Optimization methods (i.e., cost benefit analyses) - to assure that occupational exposure is maintained ALARA in developing and 

justifying facility design and physical controls. (Not used, or documented in all ALARA decisions,) 

5.5.10.5. Design objective, under normal condition - avoid releases of airborne radioactive materials to the workplace atmosphere. 

5.5.10.6. Confinement and ventilation - control the Inhalation of airborne radioactive material by workers to levels that are ALARA. 

5.5.10.7. Design or modification of a facility and the selection of materials - include features that facilitate operations, maintenance, 

decontamination, and decommissioning. 

5.5.10.8. Combination of physical design features and administrative controls - provide that anticipated occupational dose not exceed limits 

in 10 CFR 835.202; ALARA process is utilized for personnel exposures to ionizing radiation. 

5.5.10.9. Design objective for external sources of radiation In areas of continuous occupancy (2,000 hours per year) - maintain exposure 

levels below average of 0.5 mrem per hour and as far below this average as Is reasonably achievable. 

5.5.10.10 Design objectives where occupancy differs from the above (e.g., less than 2,000 hours per year) - ALARA and less than 20 

percent of the applicable standards in 10 CFR 835.20. 

*Numbers in boldface type refer to the corresponding essential elements in DOE IG G-10 CFR Part 835/B2, Occupational ALARA Program

83 NUREG-1747
Table 17: Summary of Application of ALARA to the Facility Design (continued) 

RPP S " 

Section Section Title Summaryof Section Contents 

5.5.11. ALARA Design ALARA design process described in procedures and codes of practice approved by Project management. Process ALARA design process consists of: completion of a baseline design proposal; Identification, evaluation of alternatives for baseline case Components doses; 

ALARA assessments; formal ALARA reviews; final decision process; incorporate changes into design; ALARA documentation. 

5.5.11.1. Baseline Design Baseline design proposal Is a proposed facility, or portion of the facility, that meets the criteria outlined in Section 5.5.10, ALARA Design I Proposal Criteria. 

5.5.11.2. Identification of For each radiation exposure scenario evaluated during ALARA process, alternatives are generated for later evaluation - ensures alternatives 

Alterna.ives ar osdrelsset-l and cossety 

i Substitution or rnirirrrizafinn Of -nurriq Tnrmr nffnetinn nnrAnnnnl dose 

Increased reliability of processes and equipment 

* Increasing distance and shielding to the source term 

Increasing effectiveness of engineered controls 

Decreasing the need for exposure 

Decreasing exposure time 

* Modification of the facility layout or process flow, 

5.5.11.3. ALARA Design ALARA design assessments are conducted and documented for each part of the design. The following phases or components of the design Assessments should be assessed by the designer, either individually or in combination, with involvement by a radiological engineer, as appropriate: 

* Process 

Operation and maintenance philosophy Plant layout (to include adequate provisions for access and egress to controlled areas, and adequacy of plant monitoring) 

Cell layouts 

Source minimization 

Contamination control 

Individual shield items (e.g., glovebox shielding, shield doors, shield windows) 

Bulk shielding (walls, ceilings, and floors) 

Construction/installation 

• Design aspects of operation 

• Design aspects of decommissioning. 

ALARA assessment normally conducted during site selection, but site has been pre-selected by DOE. Estimate of the dose resulting from each design alternative and the associated cost are needed for ALARA assessment. 

More than one alternative applied to an exposure situation may provide equivalent ALARA benefit - consider operational experience of 

existing plants if reasonable to do so. In the general case, in ALARA assessments consider any design modification where: 

Reducing dose might result in an increase of a conventional hazard (e.g., risk of injury from collision with equipment). Result is greater design, construction, operating, or decommissioning costs. 

Difficulties in building, operating, or decommissioning the plant are increased. 

The creation of an additional hazard does not necessarily eliminate selection of an alternative under consideration.

NUREG-1747 84
Table 17: Summary of Application of ALARA to the Facility Design (continued) 

,Section RPP Section Title Summary of Section Contents . .. 1. .... 

5.5.11.4. ALARA Design Formal reviews to look for improvements required to demonstrate ALARA compliance and to record ALARA decisions. 

Reviews ALARA reviews also can be used to record where dose reduction has been achieved by the use of "good engineering practices." 

Reviews should use appropriate checklists to ensure consistency; shall be conducted by personnel not involved directly in producing the 

design, The outcome of the reviews will record the key ALARA decisions made In each design stage. 

5.5.11.5. Consensus More than one alternative may be proposed that achieves ALARA objective. ALARA Subcomittee (ASC) will select the optimum 

Approval alternative. ASC provides recommendation(s) through Plant Safety Committee to Project Manager for consideration and approval. 

Contested issues should be clearly identified, characterized, and negotiated between ASC and Project Manager. May require General 

Manager to make final decision. 

5.5.11.6. Incorporate Following a decision to incorporate the ALARA changes into the design; the changes will be implemented using authorized design change 

Changes into control procedures. 

Design 

5.5.12. ALARA All records pertaining to the ALARA design review process including formal ALARA design reviews, cost/benefit reviews, design process 

Documentation audits, and assessments that include ALARA shall be retained in accordance with BNFL Inc. records retention procedures. 

5.5.13. Radiological Because this RPP submittal is for design work only, no element of radiological work or experiment planning need be addressed at this 

(10) Work/Experiment time. 

Planning 

5.5.14. Records BNFL Inc. shall generate and retain all records necessary to demonstrate compliance with 10 CFR 835.1001. These records include 

(11) ALARA training records, formal ALARA design reviews, cost/benefit reviews, design process audits, and assessments that include 

ALARA. 

"*Numbers in boldface type refer to the corresponding essential elements in DOE IG G-10 CFR 10 CFR 835/B2, Occupational ALARA Program

85 NUREG-1747
3.4.2.5. COMPARISON OF: NRC AND DOE ALARA GUIDANCE

Section 5.5, Paragraph 1 of the RPP for Design, Rev. 3, states "the form and content of this 

section of the PP [protection program (i.e., RPP)] is consistent with DOE Implementation Guide 

(IG) G-1 0 CFR Part 835/B2...," which was issued in November 1994. It was based on the then

current (1993) version of 1 C, CFR Part 835. Following the 1998 revisions to 10 CFR Part 835, 

the IG was updated and renumbered as DOE G 441.1-2, which was released on March 17, 

1999. However, the contra(tor may chose to use the latest revision or to continue using the old 

version. If the contractor chcooses to use the old version, it must ascertain that continued use of 

that version does not conflict with the current (1998) version of 10 CFR Part 835; in other 

words, compliance with the current law is required. It is also necessary for the contractor to 

ensure that use of the older version does not reduce the effectiveness of the RPP. At the time 

this summary was being prepared, the contractor had not conveyed any decision to DOE 

regarding the use of the revised IG. 36 

There are many differences between the NRC Regulatory Guides (RG) and the 1994 version of 

the DOE Implementation Guide (IG). The approaches are similar, as would be expected given the similarities in 10 CFR Parts 20 and 835, and given that the development of the ALARA 

philosophy, under the auspic:es of the International Commission on Radiological Protection, has 

served as a common resource for guidance developed by both the NRC and DOE. The 

Introduction to NRC RG 8.8 states that "[tlhis guide provides information relevant to attaining 

goals and objectives for planning, designing, constructing, operating, and decommissioning a 

light-water reactor nuclear power station ... " to achieve ALARA. In contrast, RG 8.10, besides 

applying to all specific licensees, "...describes an operating philosophy that the NRC staff 

believes all specific licensees should follow to keep occupational exposures to radiation ... [ALARA]." The DOE IG on ALARA applies to all radiological activities within DOE and its 

various facilities and also takes into account other hazards (e.g., chemical) that workers may 

encounter. Because RG 8.10 applies to all specific licensees, it is the most directly comparable 

to the DOE IG. Some topics not covered specifically in RG 8.10 are presented in RG 8.8; 

however, some material in the DOE IG is not specifically dealt with in either NRC RG. Although RG 8.8 applies specifically to: light-water reactor nuclear power stations, the ALARA principles 

described in it are not exclus,,ive to that type of facility. 

Section IV of Implementation Guide G-1 0 CFR Part 835/B2 provides "the basic guidelines for 

conducting an operational ALARA program." It lists 11 "...'essential elements' that shall be 

incorporated into an acceptaLble occupational ALARA program." There are parallels to this 

guidance in NRC RGs 8.8 and 8.10. Table 18 lists these to the extent that comparable 

concepts can be identified. While a direct comparison of the DOE-and NRC guidance documents is somewhat limited by the considerable difference in styles in which they are 

written, the 11 essential elements are as follows. The information in parentheses indicates the 

primary source of the requirements: DOE's "Radiological Control Manual" (RCM), with the 

associated numbers denoting the article numbers, and 10 CFR Part 835. 

1. Management Commitment: Establish commitment and participation of all line management 

and all levels of the work force to the ALARA policy (RCM 121); 

36 Per e-mail from Jeanie Polehri, DOE-RU, 3/16/00.

NUREG-1747 86
2. Assignment of Responsibilities: Assign specific responsibilities to line management and 

workers involved in implementing the ALARA program; 

3. Administrative Control Levels (RCM 211.2 and 211.4): Establish a challenging level of 

administrative control that is more restrictive than the DOE RCM Administrative Control 

Level of 2,000 mrem per year; 

4. Radiological Performance Goals (RCM 121.9, 132.1 and 133): Establish, approve, and 

review quarterly a program of radiological performance goals; 

5. ALARA Training: Require the attendance of managers and workers involved with any 

aspect of radiological operations (RCM 651-654); 

6. Plans and Procedures: Set up formal plans and procedures to attain and maintain 

occupational exposures ALARA (10 CFR 835.101 (c)); 

7. Internal Audits/Assessments: Conduct comprehensive internal reviews, audits, and 

evaluations periodically and report the results to the highest levels of site management (10 

CFR 835.102); 

8. Optimization Methodology: Use methods of optimization to assure that occupational 

exposure is maintained ALARA when developing and justifying the facility design and 

physical controls during the design of new facilities or major modification of old facilities (10 

CFR 835.1002(a)); 

9. Radiological Design Review: Ensure the integration of appropriate methods and 

considerations during the design phase to maintain occupational exposures ALARA during 

subsequent construction, modification, and operation of the equipment or facility 

(10 CFR 835.1001 (a)); 

10. Radiological Work/Experiment Planning: Integrate measures and controls to maintain 

occupational exposures ALARA for specific operations and experiments; and 

11. Records: Maintain documents that demonstrate compliance and that the program is 

adequately carried out (10 CFR 835.704(b) and RCM 711-713). 

3.4.2.6. RADIATION PROTECTION ASPECTS FROM THE NRC STANDARD REVIEW 

PLAN 

NUREG-1 702, "Standard Review Plan for the Review of a License Application for the Tank 

Waste Remediation System-Privatization (TWRS-P) Project," (Main Reference 15) provides 

guidance for the review and evaluation of health, safety, and environmental protection in 

applications for licenses to possess and use special nuclear material (SNM) during the 

remediation of radioactive tank waste at Hanford. The guidance is also applicable to the review 

and evaluation of proposed amendments and license renewal applications. Specific filing 

requirements for license applications and for issuance of such licenses are in 10 CFR Part 70, 

"Domestic Licensing of Special Nuclear Material."

87 NUREG-1747
Table 118I: Comparison of DOE Implementation Guide 

G-10 CFIR Part 835/B-2 and NRC Regulatory Guide 8.10 

DOE IG G-10 CFR Part 835/B-2 NRC RG 

Essential Element 8.10 or 8.8 

Regulatory NRC Regulatory Position Description (where available) (Number) Title I Position: Note: most are paraphrased 

Narrative Description • _____ 

(1) Management Commitment 8.10 C.l.a. Plant personnel should be made aware of management's Establish commitment and commitment to keep occupational exposures ALARA. 

participation of all line 

management and all levels of the 8.8 C.i.a. Program for Maintaining Station Personnel Radiation Doses ALARA. 

work force to the ALARA policy 

(RCM 121) 

(2) Assignment of 8.10 C.1 .c. Ensure well-supervised radiation protection capability with well

Responsibilities defined responsibilities. 

Assign specific responsibilities to 

line management and workers 8.8 C.i.b. Organization, Personnel, and Responsibilities. 

involved in implementing the 

ALARA program 

(3) Administrative Control Levels No direct A similar goal is sought under 8.10.C.1.f. (Modifications to 

Establish a challenging level of comparison procedures/equipment/facilities should be made where they will 

administrative control that is more substantially reduce exposures at a reasonable cost) and under 

restrictive than the DOE RCM 8.1 0.C.2.b. (Periodically review procedures to identify how exposures Administrative Control Level of can be reduced). 

2,000 mrem per year 

(4) Radiological Performance No direct The DOE IG requires a radiological performance goal program, to be 

Goals comparison reviewed quarterly. RG 8.8 C.1.b(1)(a)-(e) discusses a corporate Establish, approve, and review program with specific objectives and periodic review. 

quarterly a program of radiological 

performance goals; 

(5) ALARA Training 8.10 C.1 .d. Plant workers receive sufficient training. 

Require the attendance of 

managers and workers involved 

with any aspect of radiological 8.8 C.1 .c. Discusses who should receive training, and to what extent; also 

operations retraining; minimum content. 

(6) Plans and Procedures 8.10 C.1.e., RSO should be given sufficient authority; Set up formal plans and C.1 .f., Modifications to reduce exposures should be made; 

procedures to attain and maintain C.2.b. RSO & staff look for ways to reduce exposure. 

occupational exposures ALARA 

(7) Internal Audits/Assessments 8.10 C.t .b. Periodically perform a formal audit to determine how exposure might Conduct comprehensive internal be lowered. 

reviews, audits, and evaluations 

periodically and report the results 

to the highest levels of site 

management

37 R.G. 8.10 is listed first because it is the more general and, therefore, more directly comparable to the DOE IG.

NUREG-1747 88
Table 18: Comparison of DOE Implementation Guide 

G-10 CFR Part 835/B-2 and NRC Regulatory Guide 8.10 (continued) 

DOE IG G-10 CFR 8351B-2 NRC RG 

Essential Element 8.10 or 8.8 

Regulatory NRC Regulatory Position Description (where available) 

(Number) Title : Position7 Note: most are paraphrased 

Narrative Description _____ '__ 

(8) Optimization Methodology 

Use methods of optimization to 

assure that occupational exposure 

is maintained ALARA when 

developing and justifying the facility No direct The DOE IG indicates optimization methodology shall be used per 10 

design and physical controls during comparison CFR 835.1002(a) and refers to a formally documented optimization 

the design of new facilities or major methodology for certain sites. No identical NRC requirement. 

modification of old facilities 

(9) Radiological Design Review 8.10 C.1f. Modifications to reduce exposures should be made. 

Ensure the integration of 

appropriate methods and 

considerations during the design 

phase to maintain occupational 

exposures ALARA during 8.8 C.1.d Review of New or Modified Designs and Equipment Selection. 

subsequent construction, 

modification, and operation of the 

equipment or facility 

(10) Radiological Work/Experiment 8.8.C.3.a. The DOE IG specifies a formal ALARA review should be carried out, 

Planning & b. while the NRC RG 8.8. C.3.a. & b. provide similar guidance, though 

Integrate measures and controls to less prescriptive. 

maintain occupational exposures 

ALARA for specific operations and 

experiments 

(11) Records No direct RG 8.8 C.3 discusses recordkeeping within a radiation protection 

Maintain documents that comparison program, but is less prescriptive than the DOE IG. 

demonstrate compliance and that 

the program is adequately carried 

out

The purposes of the SRP are to: 

1. Ensure the quality and uniformity of staff reviews. 

2. Present a well-defined base from which to evaluate proposed changes in the scope, level of 

detail, and acceptance criteria of reviews. 

3. Serve as the basis for the review of requests by licensees for changes in their licenses. 

Thus, the SRP, at any point in time, can provide a basis for the review of proposed new or 

renewal applications and amendments to existing licenses, as well as modifications to the SRP 

resulting from new NRC requirements and licensee initiatives. 

37 R.G. 8.10 is listed first because it is the more general and, therefore, more directly comparable to the DOE IG.

89 NUREG-1747
The responsibility of the stafl in review of license application, renewal application, or license 

amendment for a TWRS-FP lacility is to determine that: 

1. There is reasonable assu rance that the facility can and will be operated in a manner that will 

not be inimical to the common defense and security. 

2. The facility will provide reasonable protection of the health and safety of workers, the public, 

and the environment. 

To carry out this responsibility, the staff: 

1. Evaluates information provided by an applicant. 

2. Determines, through incependent assessments, that the applicant has demonstrated a 

reasonable safety program that is in accordance with regulatory requirements. 

To facilitate carrying out this responsibility, the SRP clearly states and identifies those 

standards, criteria, and bases that the staff should use in reaching licensing decisions. 

Although 10 CFR Part 70, as revised8, does not specifically include a TWRS-P facility in its list 

of activities requiring the inclusion of requirements found in Subpart H of 10 CFR Part 70, the 

staff believes that a TWRS-P facility is an activity that could significantly affect public health and 

safety and, therefore, plans to invoke the requirements found in Subpart H for this type of 

facility. As such, NRC requirements in 10 CFR Part 70, as revised, require that an applicant 

submit a complete description of the safety program for the possession and use of SNM to 

show how compliance with the applicable requirements will be accomplished. 

The requirements in 10 CFR: Part 70 specify, in general terms, the information to be supplied in 

a Safety Program Description. The specific information to be submitted by an applicant and 

evaluated by staff is identified in the SRP. A license application should contain a Safety 

Program Description that adbresses all the topics in the 'Table of Contents" of the SRP. 

The applicant's, or licensee's, Radiation Safety Program and Radiation Safety Design Features 

are covered in Chapters 4.1 and 4.2 of NUREG-1702, respectively. These chapters contain 

specific information concerning what constitutes an acceptable Radiation Protection Program, from the perspective of the NRC. The 11 essential elements of the DOE G-10 CFR Part 

835/B2, shown in Table 18, are not expressly cited in the SRP; however, the guidance contains 

provisions for the review of ALARA considerations (at the design stage, during operations, and 

for modifications), organizational relationships and personnel qualifications, and training. These 

are also covered in both the DOE IG and the NRC RGs. The remaining items in Table 3.4.2.4 

would all be considered parts of a good radiation protection program, as evidenced by the 

similarities to guidance found in the NRC RGs 8.8 and 8.10. These items would not be 

explicitly covered by the SRPI-mandated review, but this should not be considered as a 

justification for eliminating any existing elements of the RPP and, therefore, possibly reducing 

its effectiveness. 

38 This reference is to the draft revision to 10 CFR Part 70, subject to on-going dialogue. The SRP uses sidebars to 

indicate additional references to the draft version of 10 CFR Part 70.

NUREG-1747 90
3.4.2.7. SUMMARY

This summary has provided a discussion of the RPP for the Tank Waste Remediation System

Privatization Project, concentrating on differences in regulatory requirements and guidance of 

DOE and the NRC. At the time this summary was being prepared, the contractor had not yet 

produced an acceptable version of the RPP for Design and Limited Construction; therefore, this 

discussion has focused on Rev. 3, the RPP for Design. 

It was noted that the NRC requires the development, documentation, and implementation of an 

RPP by its licensees (10 CFR 20.1101 (a)), but it does not require the licensee to submit this 

document for approval by the Agency and does not give it the direct authority to require 

modifications of the RPP. By way of contrast, the regulations in 10 CFR 835.101 are very 

prescriptive with respect to the requirement for an RPP. While both agencies require the 

implementation of an RPP, the DOE includes more stringent requirements such as approval by 

DOE, authority of DOE to direct or make modifications to an RPP, and the requirement that the 

RPP must address each requirement in 10 CFR Part 835. 

This summary also included comparisons of the guidance documents issued by the two 

agencies that are intended to assist the regulated entities in understanding and complying with 

its regulations (10 CFR Parts 20 and 835). As noted in Section 3.4.2.5, direct comparison of 

the DOE and NRC guidance documents is somewhat limited by the difference in styles in which 

they are written, as well as the difference in orientation (e.g., types of facilities, regulatory basis) 

and emphasis (as shown in Table 18). The DOE IG G-1 0 CFR Part 835/B2 lists 11 "essential 

elements" that are required for an acceptable occupational ALARA program. The NRC RGs 

are not so specific. 

The TWRS-P facility would be subject to review in accordance with the NRC's SRP, NUREG

1702, if there were a transition of regulatory authority from DOE to the NRC. Section 3.4.2.6 

indicates how the SRP would relate to the contractor's RPP. It was stated there that the SRP, 

at any point in time, can provide a basis for the review of proposed new or renewal applications, 

and amendments to existing licenses, as well as modifications to the SRP resulting from new 

NRC requirements and licensee initiatives. This does not necessarily imply that an existing 

facility would be required to make modifications as transition to NRC regulatory authority takes 

place. However, if regulated by the NRC, future modifications and license renewal applications 

would likely be reviewed using the SRP.

91 NUREG-1 747
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NUREG-1747 92
3.0 AREAS OF REVIEW 

3.5 CRITICALITY SAFETY 

3.5.1 CRITICALITY SAFETY PROGRAM DOCUMENT AND REQUIREMENTS 

The umbrella Nuclear Criticality Safety program document, "Criticality Safety Program for the 

River Protection Project-Waste Treatment Plant, PL-W375-NS00001, Revision 1" provides an 

overall description of the Nuclear Criticality Safety program. The intent of this program 

document is to implement the U.S. Department of Energy (DOE) Top-Level requirement in 

DOE/RL-96-0006: 'The facility shall be designed and operated in a manner that prevents 

nuclear criticality." While the document adheres to good practices generally accepted in the 

criticality safety community as well as the guidance provided by American National Standard 

Institute/American Nuclear Society (ANSI/ANS) 8.1," Nuclear Criticality Safety in Operations 

with Fissionable Materials outside Reactors," and ANSI/ANS-8.19, "Administrative Practices for 

Nuclear Criticality Safety," it lacks sufficient detail with respect to implementation of the Nuclear 

Criticality Safety Program and the controls to be used at the proposed new waste treatment 

facility. 

There is a commitment to adhere to the Double Contingency Principle. Section 5.1 of 

PL-W375-NS00001 states, "Process designs shall incorporate sufficient factors of safety to 

require at least two unlikely, independent, and concurrent changes in process conditions before 

a criticality accident is possible." The document goes on to say that Double Contingency 

protection can be provided by either the control of two independent process parameters (which 

is the preferred approach, when practical, to prevent common-mode failure) OR by a system of 

multiple controls on a single parameter. Depending on the specifics of the situation, multiple 

controls on a single parameter may not be sufficient to demonstrate double contingency to the 

U.S. Nuclear Regulatory Commission (NRC). 

Appendix 1 identifies work which will need to be completed to resolve open technical issues. 

These issues include parametric studies of keff for worst case waste mixtures as opposed to the 

Pu-water mixture discussed in the Initial Safety Analysis Report, study fissile material over

concentration accidents in the melter, study criticality parameters for the low activity waste 

process train, study solids settling in process vessels, and study boiling and drying-out of waste 

due to excess temperature. The River Protection Project-Waste Treatment Plant (RPP-WTP) 

has committed to initiate these activities prior to the submittal of the Preliminary Safety Analysis 

Report and complete these activities prior to submission of the Final Safety Analysis Report. 

However, some of these technical issues may be difficult to resolve. It is essential to identify 

any technical issues that could potentially delay start-up of the facility as early in the process as 

possible to allow for adequate time to study and resolve such issues. 

3.5.2 INTERIM CRITICALITY SAFETY EVALUATIONS 

Based on the interim criticality evaluations submitted, BNFL Inc. developed its nuclear criticality 

safety program by bounding keff for nearly dry waste solids with a maximum plutonium content. 

This approach relies upon providing controls on the feed and analyzing the process to ensure 

that no normal operations or process upsets could result in formation of a critical mass. As in 

the Criticality Safety Program document, double contingency is required but is permitted by 

either the control of two independent process parameters (which is the preferred approach,

93 NUREG-1747
when practical, to prevent cormmon-mode failure) OR by a system of multiple controls on a 

single parameter. As indicated above, multiple controls on a single parameter may not be 

sufficient to demonstrate double contingency to the NRC. Additionally, there is not sufficient 

detail to determine how double contingency will be applied to the processes nor is it clear that 

all potential concentration ITmechanisms (e.g. preferential settling of specific isotopes) have been 

evaluated. 

In the evaluations reviewed, BNFL Inc. has relied upon concentration limits for special nuclear 

material as provided for in their contract with DOE as an upper boundary for criticality analyses, despite the fact that there are waste forms within the Hanford tank waste that may be above 

those limits. The criticality values are from the Carter Handbook that is widely used in the DOE 

complex. However, these have not undergone rigorous validation. The NRC uses ANSI/ANS 

8.1 and other ANS standards that have undergone rigorous validation as detailed in Regulatory Guide 3.71. Also, BNFL Inc. has committed to only one criticality safety alarm in each area 

requiring alarms which is inconsistent with 10 CFR Part 70 requirements which requires that "coverage of all areas shall be provided by two detectors." This is one situation where the NRC 

guidance is more restrictive than the ANSI standard which allows coverage by one detector. 

There are processes or potential processes and operations that could occur at the facility that 

have not been discussed in any criticality safety evaluations such as solvent extraction and 

precipitation of solids which could have significant safety concerns, especially with the 

uncertainties in the waste stream. 

3.5.3 UNCERTAINTIES AND UNRESOLVED NUCLEAR CRITICALITY SAFETY ISSUES 

In addition to the unresolved technical issues that are identified in the nuclear criticality safety (NCS) program document, there are additional technical items identified in the limited number 

of criticality safety evaluations submitted to the NRC, such as studying the possibility of solids 

precipitation from the mixing of fissile streams, verifying the maximum density of solids and 

verifying the maximum fissile concentration in the glass product. A major point of concern is 

the large amount of uncertainty in the waste tank feed material. It seems extremely difficult to 

determine the "worst-case "conditions for analysis when there is so much question as to the 

make-up of the input waste st;:ream. Especially when considering the possibility of build-up on 

resin (such as ion exchange) filters and processes like solvent extraction and precipitation 

where fissile material could reach potentially high concentrations, great variability in the input 

waste stream could lead to large differences in process conditions. 

3.5.4 CONFORMANCE WITl-H NRC REGULATIONS 

Based on the criticality safety program document and criticality safety analyses provided, more 

information would be required to determine the overall conformance with NRC regulations. 

With the exception of the issues addressed above involving meeting the double contingency principle and the issue of dual coverage with respect to criticality alarms (this would require 

either an exemption or facility changes should the NRC ever assume regulatory authority), it is 

not possible to determine the-' state of conformance with NRC regulations for the proposed new 

vitrification facility. Much more detail about the design of the facility and processes would be 

required by the NRC to make, this assessment. Use of the ANSI/ANS 8.1 subcritical limits is 

endorsed by the NRC as stat:ed in Regulatory Guide 3.71 and is preferable the approach

NUREG-1747 94
chosen by BNFL Inc. to use safe concentration limits as established by the Carter Handbook. 

Additionally, good practices like using engineered controls like safe geometry process vessels 

and defense-in-depth should be practiced. Based on the information provided, conformance 

with NRC regulations would certainly be possible, especially if NRC regulations were 

considered in the early stages of design of the facility and processes. More detail about 

implementation of the criticality program and specifics with regard to items that would be relied 

upon for criticality safety and risk estimates would be required to fully assess the state of safety 

and compliance with respect to applicable NRC criticality safety regulations.

95 NUREG-1747
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NUREG-1747 96
3.0 AREAS OF REVIEW 

3.6 PROCESS AND CHEMICAL SAFETY 

3.6.1 STANDARDS APPROVAL PACKAGE (SAFETY REQUIREMENTS DOCUMENT AND 

HAZARDS ANALYSIS REPORT) AND INITIAL SAFETY ANALYSIS REPORT 

The BNFL Inc. Safety Requirements Document (SRD) and Hazards Analysis Report (HAR) 

arrived in September 1997 and identified numerous process and chemical safety scenarios 

(many with both radiological and chemical consequences) using a preliminary hazard and 

operability (HAZOP) method. The Contractor used reviewers with expertise in similar types of 

plants and processes to bridge areas of incomplete information because the preliminary design 

was still undergoing development and, as such, many open areas existed. The HAZOP used 

an American Institute of Chemical Engineers (AIChE) risk matrix (a binning approach) to assist 

with identifying those potential scenarios that posed a sufficient risk (i.e., consideration of both 

frequency and consequence) to warrant quantitative risk analysis and the consideration of 

safety controls. In general, the list included many entries of a benign nature that obscured 

some of the more hazardous conditions at the proposed facility. In addition, the expert review 

panel inadvertently included the effects of mitigation in some of their assessments, and, 

consequently, the frequency and consequence bins for some events were underestimated. 

Thus, potential items relied on for safety (IROFS) were overlooked. From these reviews, the 

following issues were identified in the process and chemical safety area: 

1. Incomplete process and hazard descriptions. 

2. Inconsistency of unmitigated and mitigated analyses in the HAR and fault schedules. 

3. Emphasis on active instead of passive mitigation and control. 

4. Little information on the chemical storage areas. 

5. No consideration of chemical effects due to ammonia, several resin reactions, and nitrogen 

oxides. 

6. Inconsistencies in hazards identification, structures, systems, and components (SSC) 

categorization, and design classes. 

7. Relative lack of quantification. 

8. Presence of large tanks and inventories. 

9. Limited consideration of off normal and unanticipated events, and interactions. 

The actual comments may be found in the SRD/HAR comment transmittal letter (see Footnote 

18 on page 39). 

The Initial Safety Analysis Report (ISAR) was dated January 12, 1998. The ISAR represented 

a more detailed design as compared to the SRD/HAR package. The ISAR advanced the 

design and presented quantitative analyses on several process and chemical safety scenarios,

97 NUREG-1747
including tank failures, crystalline silicotitanate powder dispersion, ion exchange resin fires, 

glass spill, anhydrous amrnmnia leaks, and a nitric acid spill. However, the level of the design corresponded to a preliminary conceptual to adequately address many of the concerns and 

issues. Furthermore, the LU.E;. Nuclear Regulatory Commission (NRC) staff's analyses indicated the input values arid methods used appeared to be nonrepresentative and could 

underestimate the potential consequences. For example, the temperature for the vapor calculations and the chemical inventories (particularly nitric acid) at risk appeared to be low. 

The following comments and issues were raised in the process and chemical safety arena as a 

result of the ISAR review: 

1. Lack of detailed process information and chemical interaction data. 

2. Lack of conservatism. 

3. Inconsistency of unmitigated and mitigated analyses. 

4. Emphasis on active instead of passive mitigation and control. 

5. Little information on chemical storage areas. 

6. No IROFS's identified for' chemical concerns. 

7. Inconsistencies in hazards identification, SSC categorization, and design classes. 

8. Relative lack of quantification. 

9. Presence of large tanks and inventories. 

10. Limited consideration of offnormal and unanticipated events, and interactions. 

The actual comments may be found in the ISAR comment transmittal letter (see Footnote 19 on 

page 40). 

3.6.2 DESIGN SAFETY FE:ATURES SUBMITTAL 

The BNFL Inc. Design Safety Features (DSF) submittal consisted of two parts. Part 1 

discussed general features used to address general hazards and concerns. In essence, this 

part presented the practical irterpretation of the codes and standards from the SRD. 

"Standard" chemical process industry approaches would be used unless radiological limits were 

shown to be exceeded. Par,: 2 discussed and analyzed ten examples in more detail, including 

loss of confinement events with both radiological and chemical consequences. The following 

comments and questions wera raised in the process and chemical safety arena from the DSF 

review: 

1. Lack of detailed process information and chemical interaction data. 

2. Lack of conservatism and consideration of uncertainties. 

3. Consideration of corrosion and materials. 

4. Juxtaposition of mitigated and unmitigated analyses. 

5. Inconsistencies in the analyses. 

The actual comments may be found in the DSF comment transmittal letter (see Footnote 34 on 

page 64). The DSF constituted an example of approaches for safety analyses and did not 

represent an actual design or safety document.

NUREG-1747 98
3.6.3 RELATED ACTIVITIES AND TOPICAL MEETINGS

The DSF submittal and related subjects were discussed at the Topical Meeting in Richland on 

March 27, 1999. A follow-up item from the meeting was the U.S. Department of 

Energy/Regulatory Unit (DOE/RU) recommendation that the Contractor develop an acceptable 

approach to the treatment of chemical hazards that is consistent with the treatment of 

radiological hazards, as required by the contract. During a teleconference call, the Contractor 

stated its intent to use standard commercial practice and process industry standards for 

chemical safety. It was also indicated that a subcontractor would completely construct the cold 

chemical storage area(s). The Contractor subsequently submitted a position paper to the RU 

proposing to relax some of the SRD criteria that would have required the quantitative evaluation 

of essentially all chemicals at the proposed facility in order to achieve cost savings. The RU did 

not accept that position39. The Contractor maintained that there were essentially no IROFSs 

related to chemical safety. 

The April 2000 Topical Meeting also discussed chemical safety. This meeting generated many 

issues which are discussed in the next section. Summarizing, the Contractor stated they were 

following the standard, chemical process industry approach to design and safety, and from their 

analyses for TWRS-P, this could correspond to worker risk levels of circa 1 E-3/yr. It was noted 

by NRC staff present at the meeting that this could be two orders of magnitude greater than the 

estimated worker radiological risk at the proposed facility. DOE appeared to be cognizant of 

the disparity. However, because the DOE/RU regulates by the contract using the approved 

SRD and the SRD incorporates chemical process industry standards, the RU appeared to be 

willing to acquiesce to a much higher level of risk from chemical events than from radioactive 

events. 

3.6.4 NRC ASSESSMENT OF THE PRESENT STATUS 

The NRC staff developed a point paper on the process safety issues associated with the 

proposed facility (see Section 3.3 of Chapter 3.0 and Chapter 7.0, Main References 18 and 19). 

This paper indicated IROFS would likely be needed to reduce the chemical risks to levels 

commensurate with NRC regulations in 10 CFR Part 70 and to meet the SRP guidance. For 

example, controls would be needed to maintain plant habitability during potential releases of 

volatile species (e.g., NOx and ammonia). Other issues from the point paper, the Topical 

Meeting, and the Topical Meeting documentation include the following general comments and 

observations: 

1. Level of Detail in the Analysis: Several times in the analyses and at the April 2000 

Topical Meeting, it was stated that a robust design consistent with chemical industry 

standards would be used. However, few specific details were presented to substantiate 

this approach. Limited information was presented on the pneumatic transport system 

(for glass formers) and its potential hazards. Furthermore, it was mentioned at the 

Topical Meeting that piping and instrumentation drawings (P&IDs) were not available for 

the chemical handling area (note, however, that a limited number of process flow 

diagrams were included in the FFP [Firm Fixed Price] submittal of 4/24/00). It would be 

39 Gibbs, D.C., U.S. Department of Energy, letter to M. Bullock, BNFL Inc., December 7, 1999. 

Gibbs, D.C., U.S. Department of Energy, letter to M. Bullock, BNFL Inc., December 22, 1999.

99 NUREG-1747
anticipated that the Contractor would provide and use more specific information based 

upon actual designs and diagrams for the safety analyses. 

2. Incomplete and Uneven Application of HAZOP Process: The Contractor does not 

appear to have conducted a complete HAZOP study of the chemical safety area. In 

addition, the presence of working components and mitigating controls appears to have 

been assumed (e.g., the static mixers and different sized connectors) for operational control purposes, which has lead to a lack of identification of potential controls for safety 

purposes. For example, successful operation of the inline mixers depends upon the 

pumping mechanisms, flowmeters, other sensors, motor-operated valve/air valves, and 

control systems. 

3. Inappropriate Focus on Routine Operations: The majority of the conditions focus on 

routine or near routinrli conditions - sometimes there are areas where SSCs are 

assumed to work but are omitted from further analysis(e.g., see the previous comment 

on the inline mixers). Offnormal and accident conditions are not well covered; for 

example, the plugging of pipes and misfeeds do not appear to have been given adequate consideration. Storage facilities such as these involve considerable operator involvement and actiVities, with frequent deliveries, and potential misfeeds would likely 

be anticipated events (e.g., potentially annually). The Contractor's proposed approach of reliance upon different types of hose connectors is likely to become a manual, 

administrative control because most supply trucks carry multiple connectors of different 

types. The relative lack of sampling and instrumentation on the chemical compositions allows these misfeeds to propagate through the facility. At least three possible misfeed 

combinations stand OUt as having potentially significant hazardous outcomes, and it 

would be expected that these would be analyzed and discussed in more detail in the 

report: 

a. Sugar (a white powder) is mistakenly added to the silo holding silica (also a white 

powder). This misfeed could double or triple the amount of sugar added to the 

melter, resulting in a loss of redox control and large discharges of NOx and 

hydrocarbons to the offgas systems. 

b. The full strength nritric acid is sent into the plant and contacts the ion exchange 

resins, resulting in a resin fire and release. 

c. The concentrated i tric acid is added to the concentrated caustic tank - or vice versa. 

This would likely overheat the tank contents from the vigorous acid/base 

neutralization reactions, leading to fuming (NOx) and possibly rupture of the tank 

(i.e., more NOx if the nitric acid tank is involved). 

4. Lack of Consideration of Improved Safety Margins: The Contractor does not appear to 

have considered simple controls that have the potential to cost effectively address and 

reduce the risk of some chemical concerns and greatly improve safety margins. For 

example, inline pH prooes would provide immediate feedback of the presence or 

absence of acids and bases and could be used - in conjunction with density or 

conductivity measurements - to prevent the unplanned mixing of reactive chemicals.

NUREG-1747 100
5. Approach of Chemical Industry Standards: The Contractor indicates the design 

approach is based upon chemical industry standards. However, as presented, the 

Contractor's approach does not appear to be consistent with AlChE guidelines for 

chemical process safety, which generally follow an as low as reasonably achievable 

(ALARA) like approach for chemicals. For example, the AIChE/Center for Chemical 

Process Safety (CCPS) Guidelines for Engineering Design for Process Safety (New 

York, 1993) discusses in detail inherently safer plants for continuing the improvement of 

the good safety record of the chemical and petrochemical industries (see Chapter 2, for 

example). These same guidelines emphasize multiple safety layers (redundancy). 

Neither "inherently safer" nor "multiple safety layers" are evident in the proposed design. 

Often, the process industry uses a risk matrix that identifies (either qualitatively or 

quantitatively) the consequences and frequencies of potential events, with the 

identification of controls to reduce risk as appropriate. This was actually initiated by the 

Contractor in the original HAR (in 1997), but does not appear to have been pursued 

further. In addition, best handling practices from the chemical industry, which generally 

emphasize precautions and redundancy (e.g., for nitric acid), do not appear to have 

been consulted. 

6. Potential High Level of Risk: As presented, the proposed approach has a potential, 

chemical/process risk level two or more orders of magnitude greater than the 

radiological risk and one order of magnitude greater than standard worker risk. This 

appears excessively high. For example, there exists a potential for a multiple fatality 

accident with a frequency of circa 1 E-3/yr from the failure of the LAW off-gas line during 

a seismic event. This is a potentially significant change from the HAR which dismissed 

this type of accident as "incredible" (i.e., implied to be less than 1 E-6/yr). 

7. Potential Inconsistency with NRC Draft Regulations and Guidance: The NRC has issued 

a standard review plan (SRP) for tank waste remediation system (TWRS) facilities as 

NUREG-1702, in final form. This identifies a target frequency of 1 E-5/yr for accidents of 

high consequences, and a target frequency range of 1 E-2Iyr to 1 E-5/yr for accidents 

with intermediate consequences. The application of ALARA (for chemical in addition to 

radiological events) is implied. Draft NUREG-1520, "Standard Review Plan for the 

Review of a License Application for a Fuel Cycle Facility," and draft NUREG-1 701, 

"Standard Review Plan for the Review of a License Application for the Atomic Vapor 

Laser Isotope (AVLIS) Facility," have similar guidance. The revision of 10 CFR Part 70 

states that accidents with potential high consequences should be rendered highly 

unlikely by the application of controls (i.e., I ROFS), while accidents with potential 

intermediate consequences should be reduced to unlikely frequencies by the application 

of controls. Again, the application of ALARA is implied for both radiological and 

chemical scenarios. High consequences and intermediate consequences refer to 

potential accidents involving radioactive materials, chemicals evolved by radioactive 

materials, and chemicals that impact the safe handling of radioactive materials. It would 

appear that the Contractors handling of some of these potential accidents at a TWRS

Privatization (TWRS-P) facility may not be consistent with the NRC approach. As an 

example, the release mentioned in Comment #6 would be categorized as a high 

consequence event and controls would be required to reduce its likelihood if the facility 

were licensed by NRC.

101 NUREG-1747
8. Specific Chemical Characteristics: Specific characteristics of certain chemicals and 

their associated potential hazards do not appear to have been considered. Chemical 

compatibility, commingling, storage, handling and other potentially hazardous issues 

need to be addressed more completely. As cited below, appropriately conservative 

scenarios and their preventative measures need further assessment: 

a. Interactions between concentrated nitric acid and concentrated sodium hydroxide. 

b. Process safety and control - example in report of inadvertent transfer of concentrated 

nitric acid into cesium ion exchange columns. 

c. Sucrose-related dust explosions. 

d. Ammonia release 

e. NOx emissions f rorn the LAW melter offgas treatment system. 

f. Seismic analysis results. 

g. Design basis event definition. 

h. Construction, laboratory, and maintenance chemicals. 

i. Start-up simulants. 

In addition, industry has codes and practices for handling many of these chemicals which 

would be expected to influence the design of chemical handling areas; for example: 

Piping - American Society of Mechanical Engineers (ASME) B31.3 Category M Fluid 

Service. 

Acid Storage - ASME Boiler and Pressure Vessel Code (B&PVC) Section VIII. 

Material - "Materials of Construction for Nitric Acid" ed. by Monitz and Pollock (National 

Association of Corrosion Engineers, 1986) is a good reference. 

Safety Instrumentation Systems (SIS.) - American National Standards 

Institute/Instrument Society of American (ANSI/ISA) S84.01-1996 and Institute of 

Electrical and Electrcnics Engineers-603. 

US NRC, "Chemical Process Safety at Fuel Cycle Facilities," NUREG-1 601. 

US NRC, "Guidance ofn Management Controls/Quality Assurance Requirements for 

Operations, Chemical Safety, and Fire Protection for Fuel Cycle Facilities," Federal 

Register, Vol. 54, No. 53, March 21, 1989, pages 11590-11598. 

9. Management Measures a.nd Corrosion: It is not clear if the Contractor has adequately 

considered management measures (including surveillance and maintenance), particularly 

for chronic effects such as corrosion/erosion. The Contractor stated in the April 2000

NUREG-1747 102
Topical Meeting documentation that standard process industry approaches will be followed. 

However, the AIChE "Guidelines" (referenced previously; page 162 et seq) note that 

corrosion is a major problem in the process industries, with 31 percent of failures due to 

general corrosion, 24 percent due to stress corrosion cracking, and 10 percent due to 

pitting. Thus, the stated, process industry-like approach is unlikely to avoid corrosion 

failures, and it would seem that the Mechanical Integrity (MI) (ref: 2.3.2.8, page 6) is a little 

bit sketchy. A complete MI covers more than just procedures, maintenance, quality 

assurance and training. The contractor might want to review the 29 CFR 1910.119 (j) to 

see what other elements need to be included in the MI program. 

10. Environmental Conditions: The analyses for the Topical Meeting used various 

environmental conditions in some of the considerations. It is not clear if these are 

adequately representative of the Hanford site. For example, the nitric acid dispersion 

calculations use a temperature of 950F, while the building designs in the related 20-month 

FFP submittal use 1 130F (and a -23°F for the minimum) for equipment exposed to the 

weather. It is not even clear if the 1 130F adequately represents the solar effect at the 

Hanford site. However, the use of the lower temperature (95°F) is likely to underestimate 

chemical evaporation rates and effects. 

11. Inline Mixers: It is unusual to use inline mixing for dilution and potential mixing reactions with 

significant enthalpies and in a batch mode (e.g., acid-base interactions). Also, inline mixers 

place considerable emphasis upon the successful operation of the pumping mechanisms, 

flowmeters, other sensors, MOV/air valves, and control systems for adequate metering and 

mixing. These other systems do not appear to have been adequately identified, considered, 

and analyzed by the Contractor. 

12. Reliance upon Administrative Controls: The design approach appears to rely extensively on 

operator actions to prevent or mitigate the effects of chemical hazards. Examples include 

the sequential dilutions and the evacuation procedures. The normally accepted practice 

and NRC regulatory emphasis are minimization of the reliance upon administrative controls. 

13. Distributed Control System (DCS): The DCS is used to control the filling of numerous tanks 

and other important functions such as mixing. Conceivably, the DCS can be considered as 

part of a safety system (safety instrumented system - SIS). It would seem that ANSI/ISA 

S84.01-1996 "Application of Safety Instrumented Systems for the Process Industries," 

IEEE-603, or another standard for safety system controls should be considered as 

applicable to the design of the chemical storage areas. 

14. Delivery Schedule and Traffic: It was mentioned during the Topical Meeting that deliveries 

would be almost continuous during the daylight shift (almost a tank truck every hour). From 

the presentation, off loading would occur at least 75 percent of the dayshift time for the 

chemical storage area near the Pretreatment Building and at least 90 percent of the time for 

the glass formers storage facility. It is not clear if the safety analyses and control strategies 

have adequately evaluated the potential impact from these frequent deliveries, such as the 

increased frequencies and inventories for spills and major accidents, and the necessary 

design and management measures. For example, a tank truck might contain 5,000 gallons 

of nitric acid in addition to the plant's inventory, and part of the delivery time would be 

outside the diked areas. This is not currently considered. A truck accident does not appear 

to have been analyzed. Standard design features for off loading reactive chemicals such as

103 NUREG-1747
diked/capture basins and an enclosure with water sprays do not appear to be included in 

the design. It would be anticipated that a future revision will address these oversights. 

The staff has also developed specific comments on the chemical safety approach at the proposed facility based upon the April 2000 Topical meeting and its supporting documentation. 

Document citations refer to: 

Department of Energy (US.) (DOE). RTP-W375-SA-0001 0, Rev. A, 'Waste Treatment 

Plant Chemical Hazards," Draft Report Review Comments by Janet Ledbetter Ferrill and 

Jim Scott/Div. 01. DOE: Richland, Washington. April 19, 2000. 

The specific comments and abservations are as follows: 

15. Adequacy of Chemical-Only Controls in a Radiochemical Facility: According to Section 2. 

on "Chemical Hazard Methodology," the "chemical-only" control strategies and standards 

are based on best industry practice and not on calculated risk. It is not clear if any particular industry or other vitrification facility has been used as precedence for determining these practices. As noted in the text, DOE Regulatory Guide 1.78 for control room 

habitability during a hazardous chemical release was published 26 years ago (June 1974) and the NUREGs are nuclear based. Other current practice is to prepare Risk Management Program (RMP) per U.S Environmental Protection Agency (EPA) guidance. It would be 

anticipated that control strategies would have to address some radiological concerns at the 

plant caused by chemicals such as plumes/habitability, fires, and misfeeds based upon 

some form of integrated safety analysis and perception issues. A high-profile radiochemical 

facility like TWRS-P would be unlikely to accept the accident risk of the process industries 

even in its chemical handling areas. 

16. Properties and Concerns of Ammonia: Ammonia is the number one chemical involved in 

accidents reported under" the EPA's risk management program for high risk industries. The 

National Institute for Occupational Safety and Health (NIOSH) immediate danger to life and 

health (IDLH) and Emergency Response Planning Guideline-3 (ERPG) values for ammonia 

are 300 ppm and 1000 ppm, respectively. In Section 3.3 of the report (page 23) it is stated 

that 43 lb/hr of ammonia will be going to the selective catalytic reduction unit for NO. abatement. A break in the ammonia supply line to NO, abatement system could, therefore, 

result in a release of around 16.5 cfm of ammonia gas (at room temperature), and would 

require about 55,000 cf[TI of air to adequately dilute the ammonia to below the IDLH value of 

300 ppm. This accident scenario, which could constitute a significant toxicological hazard to 

the facility worker, has not been considered in the Waste Treatment Plant Chemical 

Hazards report. 

17. Properties and Concerns of Cerium (IV) Nitrate: This chemical arrives as a liquid and is 

extremely corrosive to st-ainless steels (it chemically machines the surfaces at rates 

approaching several mils per hour). It is also a strong oxidizer of organic materials. It is not 

clear if the Contractor haEs considered all of these properties, particularly the impact of 

residual concentrations oa cerium (IV) as it is transferred through stainless steel piping to 

waste management. Also, the means for introducing the cerium (IV) solution into the cell 

area is not clearly explained, and temporary connections may be necessary.

NUREG-1747 104
18. Properties and Concerns of Permanganate: It is not clear if the Contractor appreciates the 

properties of this chemical. In particular, all permanganate solutions decompose at slow but 

significant rates, and leave a fine coating of manganese dioxide that is difficult to remove. 

Permanganate solutions can violently react with organic materials and accelerate the aging 

and embrittlement of elastomers and seals (e.g., in pumps and seals). 

19. Properties of Peroxide: The Contractor recognizes peroxide as a strong oxidizer. However, 

explosive catalytic decomposition can occur if typical peroxide concentrations (30-50 vol%) 

are introduced to surfaces or solutions containing transition metals (e.g., rust, iron, nickel, 

manganese). Industrial experience has shown this to be capable of rupturing pipes. 

Residual concentrations will also attack and embrittle elastomers and seals (e.g., in pump 

areas). The means for introducing the peroxide solution into the cell area is not clearly 

explained, and temporary connections may be necessary. 

20. Carbon Dioxide: The proposed facility would use large quantities of carbon dioxide (for 

surface decontamination), and has a relatively large tank containing liquid C02. However, 

potential incidents involving this material, such as asphyxiation, do not appear to have been 

seriously considered in the analyses. It would seem that C02 or breathing air monitors of 

some type might be needed and could be identified as safety controls. 

21. Concerns about Silica & Zircon Sand: There are several inconsistencies with regards to the 

"Acute Toxic" and "Chronic Toxic" data for Silica and Zircon Sand, as shown in Tables 4 and 

8 of the document. 

Table 4 

a) Silica acute and chronic toxic data needs to be the same as that of zircon sand as 

presented in Table 8. The rationale is that it is a known significant health hazard per 

NIOSH and Occupational Safety and Health Administration (OSHA). 

b) Zircon sand acute and chronic toxic data is listed as NA in Table 4. This needs to be 

replaced with the data presented in Table 8 for zircon sand. 

Table 8 

a) Silica acute toxic data needs to read same as that for zircon sand. 

22. Concerns about a 20,000 Gallon Leak of 12.2 M Nitric Acid or 19 M Sodium Hydroxide: 

Table 1 of the Waste Treatment Plant Chemical Hazards report states that there will be 

20,000 gallons each of concentrated nitric acid (12.2 M) and sodium hydroxide (19 M) on 

site. Given that both chemicals are hazardous and very reactive, the report needs to 

discuss in detail: (I) the rationale for using these very large onsite inventories in the 

proposed BNFL Inc. design, and (11) the emergency plans for a 20,000 gallon leak of either 

of these chemicals due to a leak in the discharge piping for these chemicals. Prudent 

design in accordance with AIChE guidance (essentially a chemical version of the ALARA 

principle) would dictate that provisions should be made to promptly empty the contents of 

the tank into temporary holding vessels and, as far as possible, the piping should be 

designed to eliminate the possibility of a large leak, for example, by using a top entering

105 NUREG-1747
tank discharge route andi eliminating the bottom discharge from-the tank as depicted in the 

Figure in Appendix A. Cther process industry standards (AIChE, American Petroleum 

Institute, ANSI) would likely require vacuum and over-pressure protection, venting, NOx 

abatement on the vent(s), and management measures (coding and inspections). These 

standards would likely apply to other tanks in the facility as well. 

23. Ambiguities in the Nitric Acid Scenario: There are inconsistencies and ambiguities. For 

modeling the nitric acid s;pill, a 5,000 gallon quantity of 12.2 M nitric acid is used. However, Table 1 cites up to 20,000 gallons present onsite and only one tank is identified. The basis 

for the selection of the worst case temperature of 95°F and meteorological conditions for 

dispersion of 1 m/s is not stated. Other points of needed clarification include the EPICODE 

input data and other modal parameters selected for the equation on page 43, including the 

partial vapor pressure o0: nitric acid at a weight percent of 61.1 percent and at 950F. Per the last sentence, paragraph 5, page 43 - "The partial pressure of nitric acid only must be used 

in the above equation, or the source term will be overestimated." It is not clear if this 

overestimation is stipulated in or caused by the use of EPICODE. 

24. Model Selection, Validation, and Comparisons: No copy of EPICODE (commercially available) was supplied :o verify calculations for modeling worst-case scenario. There was 

insufficient model input (dIata provided in the report to confirm or compare the EPICODE 

model results using a similar model, ARCHIE, which is a program initially developed in 1989 and available on the EPA public domain website. This program uses hydrazine instead of 

water as the reference liquid. It would be beneficial to include the rationale for using the 

EPICODE model instead of or in addition to one of the other EPA chemical release 

programs such as ALOHA, SLAB, ARCHIE or TSCREEN for various release scenarios. For 

example, EPICODE is a modeling and simulation emergency release program acceptable to DOE and EPA; ARCHIE also considers fire and -explosion hazards, which are not fully presented in this report. It is not clear that the Gaussian dispersion formulas used to 

calculate the dispersed volatilized chemical concentration downwind from the spill are a 

better representation for tiis particular site and chemical (nitric acid). Some model 

references reviewed include: 

a. ALOHA and ARCHIE: A Comparison, Mary Evans, Report No. HAZMAT 93-2, April 1993, Modeling and Simulation Studies Branch, Hazardous Materials Response and 

Assessment Division, Office of Ocean Resources Conservation and Assessment, National Oceanic and Atmospheric Administration, Seattle, Washington 98115. 

b. ARCHIE Model, U.S. EPA Region 7 Website. 

c. Atmospheric Dispersion Modeling Resources, Second Edition, Emergency Management Advisory Committee, Subcommittee on Consequence Assessment and Protective 

Action, U.S. DOE. March 1995. 

25. Rationale for Only Analyjing Nitric Acid Spills: Nitric acid is the only chemical modeled in 

an emergency release scenario. It would seem that other chemicals, such as sodium 

hydroxide and ammonia, might also pose concerns. It would be helpful to have the basis 

stated for concluding that 12.2 M nitric acid was determined to be the only chemical with 

appreciable volatility.

NUREG-1747 106
26. Heat Tracing of Nitric Acid Equipment and Lines: All the tanks/associated piping in the 

Process Description section (3.1.1 through 3.1.5), except the 12.2 M nitric acid tank, have 

heat tracing and insulation for freeze protection. Some basis would be anticipated to 

support this. 

27. Nitric Acid Coaxial Piping - (Ref: 3.1.1, page 8): The report states that "piping outside the 

berm will be coaxial piping" to protect personnel from being exposed to nitric acid. It is 

advisable to look at the pressure side of acid piping inside the berm to ensure that no 

personnel will be exposed to the acid accidentally. Sometimes the piping inside the berm 

may be close enough to personnel outside the berm to present a potential hazard in the 

event of an acid leak. 

28. Completeness of Interaction Matrix: The matrix of possible interactions of between different 

chemicals (page 29) does not include the interactions between utilities (steam and water, 

especially steam) and chemicals. Steam is used to provide heat to the urea unit (3.3.1, 

page 24) and the Cs HNO 3 recovery system (pages 33-34). It may not be a bad idea to 

consider the possible interactions between steam and the chemicals. For the sake of 

completeness, the interaction matrix may be expanded to include chemicals, materials of 

construction, utilities, operator(s), energy source, and others (air/water/land). 

29. Stack Release Concentrations: It would seem that the postulated NOx release coming out 

of the stack at approximately 2000-3000 ppm exceeds the 100 ppm IDLH at the stack and 

quite possibly at the ground. Not knowing the height of the stack, flow rate, etc., of the 

system, it would not be reasonable to give a rough order estimate on the ground level NOx. 

However, BNFL Inc. definitely needs to address this issue by modeling, surrogate testing, 

and also possibly looking at other similar facilities to determine the ground level 

concentrations and emergency response procedures. Simply cutting off the feed does not 

seem to be an adequate response or control strategy. 

30. Pneumatic Transfer Systems for Glass Former Powders: The Contractor intends to use a 

dense-phase pneumatic transfer system for the transport of the glass former powders. Due 

to the higher operating pressures involved (typically around 50 psig), a leak in the piping, 

such as at a coupling or a diverter valve, can result in an intense spray of dust into the 

immediate environment. Furthermore, the dust from some of the powders is hazardous. 

For example, silica (a major ingredient of the glass formers) is carcinogenic and has an 

OSHA Permissable Exposure Levels (PEL) (respirable fraction) and American Council of 

Government and Industrial Hygienists (ACGIH) threshold limit value (respirable fraction) of 

100 micrograms/m 3 (this is less than 0.1 ppm). Consequently, it would be anticipated that a 

leak scenario involving this system would be explored to ensure that worker safety is not 

compromised. 

31. Other Pneumatic Transfer System Features: There are other design safety considerations 

for these systems, including the following: 

a. Pipe couplings and hangers need to be designed to withstand pressure surges of two to 

three times the normal operating pressure. Such surges are normally encountered in 

dense-phase pneumatic transport systems and can induce significant, hammer-like 

stresses.

107 NUREG-1747
b. Pneumatic transpodr systems are prolific generators of static electricity and the design must incorporate measures that prevent or mitigate potential accidents initiated by static 

charges. 

c. There are likely to be, ASME code requirements and stamps for many of the 

components in the system (e.g., vessels). 

32. Multiple Uses for Some Cbomponents: Some lines and components will be used to mix 

and/or feed different chemicals at different times. It is not clear if an adequate analysis has 

been performed to addre:ss the potential, inadvertent mixing of incompatible chemicals in 

these lines and components. For example, on different occasions, the MCCMT 

(Miscellaneous Cold Che,-micals Mix Tank) will be used for the preparation and addition of 

sugar shimming solutions and decontamination chemicals (e.g., usually nitric acid based in 

TWRS-P - see Section 3,,.3.3 of the report). Since the preparation of chemicals in the 

MCCMT will be a manually controlled operation subject to human factors in addition to 

component failures, the; Contractor's design may be vulnerable to the inadvertent generation of copious quantities of \JOx from the unplanned mixing of these chemicals. It would be 

anticipated that such scenarios would be analyzed and the appropriate controls identified. 

As an aside, this was the'• reason that the addition of nitric acid to the shim tank was 

restricted at West Valley Demonstration Project even though the tank was vented to a 

venturi scrubber. 

33. Diesel Fuel Concerns: Large quantities of diesel fuel will be stored and handled at the 

facility. It would be anti:'ipated that there would be a discussion of the safe handling of this 

fuel.

NUREG-1747 108
3.0 AREAS OF REVIEW 

3.7 FIRE PROTECTION 

3.7.1 STANDARDS APPROVAL PACKAGE (SAFETY REQUIREMENTS DOCUMENT AND 

HAZARDS ANALYSIS REPORT) AND INITIAL SAFETY ANALYSIS REPORT 

Although there were fire protection comments generated during the Standards Approval 

Package (SAP) and the Initial Safety Analysis Report (ISAR) reviews, most of these comments 

were concerned with clarifications or omission of details. The design had not yet progressed to 

a point were an actual description of fire protection systems could be developed and reviewed 

against our existing review criteria. 

From these reviews, comments and questions were raised concerning: 

1. Discounting of fire in the hazard analysis due to non-conservative assumptions regarding 

combustible loading. 

2. Failure to reference specific applicable subsections in National Fire Protection Association 

(NFPA) Codes in the "tailoring process." 

3. Need to determine a minimum fire resistance for fire barriers. 

4. Need to designate building construction type as per NFPA 220. 

5. Manual fire fighting requirements and coordination with the Hanford Fire Department. 

6. Need to properly consider heating, ventilation, and air conditioning (HVAC) filter fire in 

hazards analysis. 

7. Need to commit to specific NFPA standards. 

The actual comments may be found in the comment transmittal letters4". 

Most of these concerns were resolved through the upcoming Topical and Level 1 Meetings on 

fire protection and the documentation provided to support those meetings. 

3.7.2 TOPICAL MEETING ON FIRE PROTECTION (FEBRUARY 1999) 

The first detailed information provided by BNFL on fire protection was provided in the 

"Transmittal of Information for the February Topical Meeting 41." This letter was written as a 

response to U.S. Department of Energy/Regulatory Unit (DOE/RU) comments in their Initial 

40 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, 

October 17, 1997. 

Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, February 6, 

1998. 

41 Edwards letter to D.C. Gibbs, U.S. Department of Energy, February 09, 2000.

109 NUREG-1747
Safety Evaluation Report and described elements of the Tank Waste Remediation System Privatization (TWRS-P) fire safety program in more detail than was provided in the ISAR and 

SAP submittals. In this submittal information was provided regarding: 

1. A description of the TWIA1RS-P fire protection program. 

2. TWRS-P Facility Design and Layout Configuration. 

3. Life Safety Features. 

4. Fire Protection System. 

5. Fire Protection Personne-l. 

6. Fire Hazards Analysis - Methodology and Assumptions. 

7. Analysis to Achieve a Safe State. 

8. Seismic/Fire Interaction,. 

9. Applicability of Codes and Standards. 

10. Fire Resistance of Fire Area Boundaries. 

11. Structural Steel Fireprooling. 

12. Fire Doors, Ducts, Dam;pers, and Penetration Seals. 

13. Employment of Automatic Suppression Systems. 

14. Filter Plenum Fire Protec:lion. 

15. Compliance with Standards for Flammable Gases. 

16. Fire Safety Features for Mechanical/Electrical Systems Areas. 

17. Baseline Needs Assessment. 

The information provided closed many of the concerns of both NRC and the RU. Concerns that 

remained unresolved or came about as result of the submittal were: 

1. Interfacing with the Hanlord Fire Department and the Need for an Onsite Fire Brigade. 2. Extent of Fireproofing of Structural Steel. 

3. Extent of Employment ol0:Sprinkler Systems. 

4. Filter Plenum Protection. 

A revised version of the February 9 submittal was submitted to the RU on March 4, 1999. This 

version clarified some of the responses to address RU concerns, however, it did not resolve the 

NRC concerns listed above. 

3.7.3 LEVEL 1 MEETING ON FIRE PROTECTION (SEPTEMBER 1999) 

The next meeting on fire protection involving the NRC staff was the first Level 1 Meeting held 

on September 29, 1999 at Bl\IFL Inc. in Richland, Washington. 

In regard to interfacing with the Hanford Fire Department, no definitive decision had been made 

as yet. As presently plannedl, the TWRS-P project will use a baseline needs assessment based 

on the resLilts of the Fire Haz;iards Analysis (FHA) which will be reconciled with the capabilities of the Hanford Fire Department (HFD). A question arose about training of facility staff in use of 

fire extinguishers. BNFL Inc. indicated that only selected operations people would be trained 

and that there was no plan at present to rely on facility personnel for fighting incipient fires. 

The meeting minutes specified that fire-exposed structural steel that supports a fire barrier 

(wall, floor, ceiling) will be firE~proofed to maintain the fire-resistance integrity of the building. BNFL inc. also stated that their approach would be consistent with NFPA 220, the NRC criteria,

NUREG-1747 110
although it might not be consistent with the Uniform Building Codes (UBC) (the DOE criteria). 

During the meeting, however, the possibility of not fire-proofing structural steel in areas of low 

combustible loading was also discussed. This type of approach would not meet either DOE or 

NRC criteria. A letter 42 providing comments on the Explosive Hazards I and II topical meetings 

also contained a short description of the NRC criteria for building construction and expressed 

concern about the stated BNFL Inc position regarding fire proofing of structural steel. 

The use of sprinklers was discussed at length between the RU and BNFL Inc. Most of the 

discussion was centered around the desire of the RU (and NRC) for BNFL Inc. to establish a 

definitive set of criteria for use of sprinklers in a given area. BNFL Inc. replied that they would 

rely on the FHA and judgement because of the uniqueness of every different fire area in terms 

of type and distribution of combustibles, heat release rates, ventilation, etc. Some tentative 

guidance that BNFL Inc. provided included the following consensus reached by the project fire 

protection engineers: 

1. C5 Areas: High potential for contamination areas having low fireloading. It was the 

understanding of the RU and NRC that BNFL Inc. did not want to provide sprinklers or steel 

fire-proofing in some C5 areas. 

2. C2 Areas: Low potential for contamination areas having anticipated in-situ and transient 

combustibles. These areas would likely be sprinklered. 

3. C3 Areas: Moderate contamination areas with varying degrees of combustible loading and 

safety significance. Sprinklers are expected to be used selectively in these areas with 

justifications provided where sprinklers are not used. 

This guidance, in the opinion of the RU and the NRC staff, appeared to leave too much 

uncertainty to alleviate the concern about sprinkler coverage. 

Water spray protection for filter plenums was not discussed in detail during the meeting. It was 

mentioned that the high efficiency particulate air (HEPA) filters would be canister type rather 

than the U.S. rectangular-type filters, and in the opinion of BNFL Inc., not as likely to burn. 

Also, there was discussion of providing automatic water sprays with an optical detector to signal 

the control valve. BNFL Inc. said that the issue was still being considered and did not wish to 

make a commitment. 

BNFL Inc. also used the meeting to back-off somewhat on their earlier commitment to provide a 

seismic supply and standpipe system. BNFL is considering the use of a mobile source such as 

a tanker truck and pumper. The RU noted that the Hanford Fire Department tanker trucks 

are used primarily for brush fires and are not kept filled in the winter. NRC staff expect 

resolution of this issue without significant interaction with BNFL Inc. 

42 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, November 

4, 1999.

ill NUREG-1 747
3.7.4 LEVEL 1 MEETING) ON FIRE PROTECTION (MARCH 2000)

As of the last Level 1 Meetirng in fire protection in March 2000, there were three issues in fire 

protection of primary interest to the RU and NRC staffs. These issues are structural 

classification of the process buildings, structural steel fire proofing, and protection of the final 

HEPA filters from the effects.; of fires. The RU also identified confinement system integrity as a 

fire protection issue. However, since this issue involves basic design of the ventilation system, NRC staff would prefer to make this issue a plant systems/radiation protection issue for the time 

being. 

In regard to the process bui'! ding construction type issue, BNFL proposes a construction type which would not meet the NRC requirements for Type I construction as per NFPA 220 nor the 

fire resistive construction types required by a strict interpretation of the UBC. The RU has 

commented to BNFL Inc.43 that a justification for the selection of a nonfire resistive construction 

type would have to include identification and analysis of building areas with unusual or 

concentrated hazards, bounding fuel packages, and critical structural members whose failure 

due to fires could structurally undermine a significant portion of the facility. The analysis would 

also be expected to address any fire protection administrative controls, such as transient 

combustibles and vehicle access to the proximity of process buildings (e.g., control of exterior 

fire exposure hazards), which are relied on. The NRC stated44 to the RU that the staff was in 

agreement with the RU criteria for resolution of this concern. 

The steel fire proofing issue is related to the construction type issue. I f fire resistance of 

outside walls is not required, fireproofing of structural steel for these walls would not be 

required. In addition, BNFL Inc. wishes to avoid fireproofing of structural steel in areas of low 

combustible loading. The FU has commented to BNFL (Footnote 43) that the RU continues to 

expect that TWRS-P structural fire proofing will meet the requirement of the Uniform Building Code. The RU also expects t:hat any type of equivalency analysis will demonstrate that there 

are no fire hazards that presient a potential threat to structural steel integrity. The NRC stated 

(Footnote 44) to the RU that hlRC staff was in agreement with their comments and added two 

additional points: 

1. Lack of steel protection b'ased on intended use would severely limit any future modifications 

for use of that area, and 

2. Lack of fire suppression capability along with lack of steel protection in the same area 

makes administrative control of combustibles the only defense measure and makes the fire 

protection in the area a potential vulnerability that would be identified in a Fire Hazards 

Analysis (FHA)/Integrated Safety Analysis (ISA). 

BNFL Inc. stated in the fire protection level I meeting that no water sprays would employed in 

filter plenums based on the Superior qualifications of the circular filters to be used. Both NRC 

and DOE fire protection criter'ia call for water spray protection in the final filter bank of the 

43 Gibbs, D.C., U.S. Department o:' Energy, letter to M. Bullock, BNFL Inc., May 03, 2000. 

44 Tokar, M., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, June 05, 2000.

NUREG-1747 112
confinement system. At this time the RU (May 03, 2000 - Footnote 43) is requesting 

clarification on the BNFL Inc. position. 

3.7.5 PRESENT STATUS WITH REFERENCE TO POINT PAPER 

Other fire protection issues that appear to be moving toward resolution but will require more 

detailed information are listed below. Some, but not all, of these issues are discussed in the 

TWRS-P Fire Safety Point Paper4

". 

1. Emergency lighting - Emergency lighting was addressed in the Part b-1 deliverable and 

described as "not important to safety." No engineering evaluations were provided to 

address safety significance and operating requirements for emergency lighting. 

2. Fire detection and alarms - Fire detection and alarms were addressed in the Part b-1 

deliverable and it was stated that they would be designed in accordance with NFPA 72. 

Universal coverage appears to be provided but information on the types of detectors to be 

placed in various areas was not provided. 

3. Fire water pump design - A conceptual drawing of fire pump placement and plumbing was 

provided in the Part b-1 deliverable. Commitments were made to follow NFPA 20 and 24. 

Fire suppression demand calculation and details of pump plumbing will need to be provided. 

4. Seismic water supply - No details provided as yet. A tanker truck and pumper have been 

initially discussed. 

5. Baseline needs for manual fire fighting and coordination with Hanford Fire Department: A 

baseline needs assessment will be required for the Construction Authorization Request 

(CAR) but will need more of the plant design to be completed. Discussions have taken 

place with the Hanford Fire Department but a formal interface is awaiting contract resolution 

between DOE and the project. 

Based on discussion with BNFL Inc. all of these issues were expected to be resolved at the 

time of the CAR submittal. 

Other potential issues that were discussed in the Fire Safety Point Paper but appear to be 

resolved include: 

1. Fire area boundaries. 

2. Treatment of combustibles in FHA. 

3. Spurious actuations and circuit protection. 

4. Penetration seals. 

In the Issues Close Out Topical Meeting of June 27, 2000, BNFL Inc. identified three issues in 

fire protection as being open issues: 

45 Leach, M.N., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, March 02, 

2000.

113 NUREG-1747
1. Use of Automatic Sprinklers. 

2. Structural Steel Fireproofing. 

3. Hanford Fire Department Interface. 

BNFL Inc. claimed closure in the area of automatic sprinklers because of the basic consensus 

between the project and RLI described in the meeting minutes from the March Level 1 Meeting; that is, for areas other than C5 areas, the project may omit sprinklers only with FHA justification and RU approval and a preliminary FHA will be performed on selected parts of the Pretreatment 

Building to demonstrate the projects philosophy on sprinkler coverage. The NRC agrees that 

based on the BNFL Inc. corrmnitments, this issue may be considered closed. 

In regard to structural fireproofing, the project still believes that verbatim compliance with the 

UBC (and due to similar requirements, the NRC standard review plan) is not warranted by the 

hazards. The architects are preparing an equivalency evaluation for review by the RU. Based 

on this, the NRC would not consider the issue closed until the equivalency proposal has been 

reviewed and approved. 

Although technical interface issues were discussed with the Hanford Fire Marshal on May 10, 2000, the interface is on hold pending contract resolution between DOE and the project. The 

NRC would not consider this issue closed until a contract has been completed between the 

HFD and the project. 

The NRC staff considers water sprays in filter plenums to still be an open issue although this 

was not addressed by BNFL. Inc.

NUREG-1747 114
3.0 AREAS OF REVIEW 

3.8 EXPLOSION PROTECTION ISSUES 

3.8.1 STANDARDS APPROVAL PACKAGE (SAFETY REQUIREMENTS DOCUMENT AND 

HAZARDS ANALYSIS REPORT) AND INITIAL SAFETY ANALYSIS REPORT 

Although there were numerous fire protection comments generated during the Standards 

approval Package (SAP) and the Initial Safety Analysis Report (ISAR) reviews, most of these 

comments were concerned with clarifications or omission of details. The design had not yet 

progressed to a point were an actual description of explosion control systems could be 

developed and reviewed against our existing review criteria. 

From these reviews comments and questions were raised concerning: 

1. Flammable Gases in waste receipt tanks. 

2. Presence of ammonium nitrates and organic vapors in melter effluents. 

3. Buildup of ammonium nitrate in off gas system ducts and filters. 

4. Capacity and reliability of passive hydrogen venting system. 

5. Monitoring of hydrogen and other flammable gases. 

The actual comments may be found in the comment transmittal letters (Footnote 40). 

3.8.2 DESIGN SAFETY FEATURES TOPICAL MEETING-HYDROGEN CONTROL 

(JANUARY 1999) 

The first relatively comprehensive treatment of the hydrogen control problem was presented at 

the Topical Meeting on hydrogen control as the first example of design safety features. BNFL 

Inc. presented calculations of hydrogen generation rates that the U.S. Nuclear Regulatory 

Commission (NRC) staff questioned as possibly unconservative4 1 4

6. The proposed means of 

keeping the hydrogen within acceptable limits if active ventilation should fail was a passive 

ventilation system relying on the density differences between air and hydrogen as well as the 

thermal gradient between the vessel and the cell. Hydrogen monitoring was recommended for 

only a short time during initial plant operations. This approach was also questioned in the 

February and March 1999 comment letters. The NRC staff directed its contractor, the Center 

for Nuclear Waste Regulatory Analyses (CNWRA), to provide an independent calculation of 

hydrogen generation rates and indicate areas of significant uncertainty. CNWRA responded 

with the report "Review of BNFL Inc. Design Safety Features Deliverable: Hydrogen Control in 

High Level Waste Storage Tanks 48." In this report, CNWRA showed that a rate of hydrogen 

46 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "Review of 

Hydrogen and Explosion Information Presented at the Topical Meeting on January 29, 1999." February 12, 1999. 

47 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, 

"Comments Concerning BNFL Inc.'s Design Safety Features Submital." March 18, 1999. 

48 Center for Nuclear Waste Regulatory Analyses (CNWRA). CNWRA 99-001, Rev 1, "Review of BNFL Inc. 

Design Safety Features Deliverable: Hydrogen Control in High Level Waste Storage Tanks." CNWRA: 

San Antonio, Texas. April 1999.

115 NUREG-1 747
generation over three times that calculated by BNFL Inc. could be calculated using equally valid 

inventory assumptions as those used by BNFL Inc. Even greater rates could be determined by taking into account uncertainty in parameters and off normal process conditions. Because the 

successful operation of the passive ventilation system was sensitive to hydrogen generation 

rates, the results of the CNVWRA studies raised considerable concern in the NRC staff and the 

RU about the reliability and adequacy of the passive ventilation system. 

3.8.3 EXPLOSIVE HAZARDS TOPICAL MEETING I (AUGUST 1999) 

BNFL Inc. reevaluated the explosion hazard due to radiolytic hydrogen generation as well as 

other possible sources and prepared reports for two topical meetings, Explosive Hazards I and 

I, which were held in August and September 1999, respectively. The August Topical Meeting (Explosive Hazards I) addressed hazards from melter steam explosions, explosions from 

nitrate organic reactions in the melter offgas system, over pressurization of an ion-exchange 

column, ammonium nitrate explosion in the off gas system, and a sugar dust explosion in the 

feed preparation vessel. Although none of these events were considered credible by BNFL 

Inc., all of these event were left as potential issues by NRC and the U.S. Department of 

Energy/Regulatory Unit (DOE/RU) pending further information and analysis 49

. A resubmittal of 

the August 1999 topical report in March 2000 and information from design review meetings alleviated some of the NRC :concerns about steam explosions, nitrate-organic reactions, and 

sugar dust explosions5

". NRC staff present concerns about steam explosions are primarily the 

potential for a refractory failure allowing molten glass to contact water in the cooling jacket. Potential explosions caused by the contact of water and the cold cap were addressed by the 

BNFL Inc. responsesto RU comments. Also, the BNFL Inc. process does not appear to be as 

vulnerable to radiological releases from organic-nitrate and sugar explosions as originally determined from the first submittal, although some questions still remain. Potential explosions from ammonium nitrate formation and over pressurization of the ion exchange column are still 

considered as open issues.. 

3.8.4 EXPLOSIVE HAZARDS TOPICAL MEETING II (SEPTEMBER 1999) 

The September Topical Meeting (Explosive Hazards II) was concerned with the potential for 

explosion of hydrogen gas in approximately 40 process tanks. At this time an active ventilation 

system for hydrogen build-uLp was proposed to take the place of the passive release system that was proposed earlier, probably for a single tank. This active system consisted of an air 

extract system with two 100 percent fans and one 100 percent backup fan. Process air is used 

for dilution during normal ope ration. An air vent system is provided for loss of offsite power (with loss of process air) coriclitions. Although the staff considered this design a significant improvement over the passive system, there were concerns from the presentation and report 

(Footnote 49). These concerns included: 

49 Pierson, R.C., U.S. Nuclear Re:;ulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "NRC and 

CNWRA Comments on the Explosive Hazards Topical Meetings I and II." November 4, 1999. 

50 Tokar, M., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "NRC 

Response to March 23, 2000 Letter Concerning the Resubmittal of the August and September 1999 Topical Meeting 

Reports on Explosive Hazards." April 27, 2000.

NUREG-1747 116
1. Hazard evaluations based on normal operating conditions rather than off normal. 

2. Rupture of vessel vent jumper was not considered in hazard evaluation. 

3. Loss of ability to maintain pressure control in vessel vent system was not considered. 

4. Large surges in offgas were not considered. 

5. The worse case detonation scenario was not properly evaluated. 

A review of the resubmittal in February 2000 also led to a re-evaluation of the open issues. The 

NRC staff concluded that rupture of the vessel vent jumper and loss of ability to maintain 

pressure control can probably be nearly eliminated by an expanded set of performance 

requirements on the system. The other concerns remain however. 

3.8.5 NRC ASSESSMENT OF PRESENT STATUS 

In regard to radiolytic hydrogen explosions, the point paper (Footnote 45) accurately reflects the 

status of the hydrogen control system. An issue that was not resolved at the explosive hazard 

topical meetings was the need for monitoring. This issue was addressed at the June 2000 

Topical Meeting on Close Out of Open Issues. BNFL Inc. has argued that air flow monitoring 

meets the requirement in Section 3-4.1 of the National Fire Protection Association (NFPA) 69, 

"Instrumentation shall be provided to monitor the control of the concentration of combustible 

components." The NRC staff agrees with this argument. BNFL Inc. also argued that the 

concentration of hydrogen would always be below the measurement threshold of normally used 

monitoring instruments except in the case of system failure; which BNFL Inc. has calculated to 

be incredible. Unfortunately, BNFL Inc. did not provide any calculations to show what the range 

and probability distribution of hydrogen concentrations would be under normal operating and 

expected off normal conditions could be. When the project provides these calculations and the 

RU accepts the reliability of the active venting system, this issue should be closed. 

The point paper also discusses sugar dust as another significant potential source of an 

explosion. However, the addition of inerting agents to granular sugar before it's pneumatically 

conveyed to the feed hopper and the use of gravity feed to the melter feed preparation vessel 

are expected to preclude likely explosion scenarios. The project has not yet addressed the 

selection of codes and fire protection measures nor has it evaluated the potential for and 

consequences of a vehicle accident in the sugar storage area (Footnote 50). 

Ammonium nitrate formation and organic vapors were addressed briefly in the point paper 

under fire protection for filter plenums. The technology as presented by BNFL Inc. appears 

capable of preventing explosive buildups in the offgas system and this should probably be the 

strategy as opposed to using fire protection methods as discussed in the point paper. 

Steam explosions in the melter and over pressure in the ion exchange column were not 

considered to be fire protection issues and were not addressed in the point paper.

117 NUREG-1747
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NUREG-1747 118
3.0 AREAS OF REVIEW 

3.9 ENVIRONMENTAL PROTECTION 

3.9.1 ENVIRONMENTAL IMPACT STATEMENT 

The U.S. Department of Energy (DOE) has published an environmental impact statement (EIS) 

on Tank Waste Remediation Systems (TWRS)5 1 . This EIS identified as the preferred 

alternative the phased implementation approach (Phase 1/11), with separations and treatment in 

external facilities. Vitrification is identified for immobilizing both the low activity waste (LAW) 

and high level waste (HLW). The EIS analyzed impacts and accident scenarios for the 

proposed facilities with the intent of enveloping any specific designs and process approaches 

that would be developed by the TWRS-Privatization (TWRS-P) program. Scenarios involving 

spray leaks were identified as having the potential for the most significant effects. The NRC 

staff did not review the EIS as part of its involvement. 

3.9.2 STANDARDS APPROVAL PACKAGE (SAFETY REQUIREMENTS DOCUMENT AND 

HAZARDS ANALYSIS REPORT) AND INITIAL SAFETY ANALYSIS REPORT 

The standards criteria in Section 5.3, "Environmental Radiation Protection," of Volume II of the 

Safety Requirements Document (SRD) outlines the Contractor's approach for the 

environmental protection (EP) program. Although the safety criteria (SC) defines the limits of 

the EP program, this material does not appear to be well supported throughout the remainder of 

the SRD package (Reference 3). For example, environmental monitoring is committed to (SRD 

criteria 9.5-1 and 9.5-2) but not described in relation to the anticipated plant operations. In 

addition, the SRD and Hazards Analysis Report (HAR) inconsistently commit to protection of 

the environment and frequently limit EP in terms of the worker and the public. Other major 

comments included: 

.1. Consistency and integration of environmental submittals with respect to other submittals 

and with the regulatory agencies (DOE/Regulatory Unit (RU), DOE Richland Operations 

Office (RL), DOE Office of Environment, Safety and Health (EH), U.S. Environmental 

Protection Agency (EPA), and the State of Washington). 

2. Application of safety categorizations for EP (i.e., structures, systems, and components 

(SSCs) that prevent or minimize environmental contamination). 

3. Assessment of incidents for discharges to the vadose zone. 

4. No clear identification of effluent and emission points on the site plan and facility 

information. 

5. Description of offgas emissions and controls, including the effect from the emergency 

offgas system and iodine-1 29 removal systems. 

51 Department of Energy (U.S.)(DOE). DOE/EIS-0189, "Tank Waste Remediation System, Hanford Site, Richland, 

Washington, Final Environmental Impact Statement." DOE: Richland, Washington. August 30, 1996.

119 NUREG-1747
The Initial Safety Analysis Report (ISAR) provided little additional detail on the EP approach. It 

does refer to a separate "Ervironmental Report" by the Contractor that presents a broad 

overview of monitoring and rnitigation approaches. The U.S. Nuclear Regulatory Commission 

(NRC) comments (see Footnote 19 on page 40) noted that detail is not provided on such items 

as media to be sampled, sam pling locations, sampling frequencies, methodologies, quality assurance/quality control (Q,/JQC), action levels for elevated measurements or other 

components that might be expected in an EP program (e.g., see the standard review plan Reference 15). Furthermore, the ISAR focused on impacts to the workers, the co-located 

workers, and the public, and did not include an assessment of the effects upon the environment 

(e.g., from accidents). Thus, it was not clear if adequate protection of the environment is 

achieved by SSCs that protect the health and safety of the public (Reference 5 -the 

ISAR-Section 3.3.6, page 3-35). 

Similar NRC comments were submitted later5

" requesting information on how environmental 

protection was supported by the implementing codes and standards listed in the (SRD) and 

updated Integrated Safety Management Plan (ISMP). 

3.9.3 DESIGN SAFETY FEATURES 

The Design Safety Features (DSF) submittal did not discuss EP in any significant detail. 

3.9.4 RELATED ACTIVITIEIS AND TOPICAL MEETINGS 

The Contractor was working with the EPA Regional Office, Washington Department of Ecology, and Washington Department of Health on environmental topics and permitting issues. The 

work required for permit applications started in 1998. Test results and modeling efforts were 

required for such essential items as melter offgas and vessel vent streams. 

Additional NRC comments on Contractor responses were submitted to the DOE/RU5 3 . Many of 

the comments dealt with compliance with the Washington Administrative Code (WAC), and 

there was an initial view that compliance with the WAC was limited to air discharges. The 

facility design was such that no radioactive liquid discharges were to be made because all 

radioactive liquids were to be routed through the existing Effluent Treatment Facility. 

A BNFL Inc. letter forwarded standards identified for safety criteria54. A second set of NRC 

comments on the Contractor Environmental Radiological Protection Program (ERPP) and its 

52 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "NRC 

TWRS Section Comments on BNFl.'s Revised SRD and ISMP." July 23, 1998. 

53 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "U.S. 

Nuclear Regulatory Commission's Comments on BNFL's Tank Waste Remediation Systems Safety Criteria for 

Environmental Radiological Proteclicn Program and for Environmental Radiological Monitoring." December 31, 

1998. 

54 Edwards, BNFL Inc. letter to D.C. Gibbs, Department of Energy, RE ERPP. January 18, 1999

NUREG-1747 120
revision were submitted to the DOE/RU15 . The DOE/RU disposition of the comments were 

forwarded to NRC on March 17, 1999.56 NRC reviewers felt that as low as reasonably 

achievable (ALARA) criteria from the WAC was the correct citation, since 10 CFR Part 835 

refers to ALARA only in the context of an occupational radiation protection program. In 

addition, there are no environmental requirements in 10 CFR Part 835. However, DOE/RU felt 

that 10 CFR Part 835 was adequate. The correct citation for Best Available Radionuclide 

Control Technology (BARCT) for air emissions is WAC 246.247.120, Appendix B. 

For DOE sites, the major Federal law regulating air emissions is the Clean Air Act (CAA). The 

requirements for obtaining permits is to ensure that any emissions of listed hazardous air 

pollutants (HAPs), including radionuclides, comply with the National Emissions Standards for 

Hazardous Air Pollutants (NESHAPS), Section 112. The emission limits are given in State 

Implementation Plans (SIPs). NRC reviewers pointed out that even though DOE intended to 

use the existing monitoring network to collect air samples during construction and operation, the 

Contractor needed to conduct environmental surveillance to demonstrate compliance with the 

ERPP. 

Other clarifications were required regarding the concepts of "groundwater protection" versus 

"groundwater monitoring." The TWRS-P facility was to be operated under the Resource 

Conservative Recovery Act of 1976 (RCRA) as a Treatment Storage and Disposal (TSD) 

facility, and groundwater monitoring was to be addressed as part of RCRA permit requirements. 

NRC reviewers felt that although International Organization for Standardization (ISO) 14001, 

which was selected by the Contractor for groundwater monitoring, provided a sound 

management plan framework, specific codes and standards would need to be identified. 

NRC reviewers pointed out that sections of the WAC related to solid/hazardous wastes should 

have been included in the regulatory basis, since the state of Washington had regulatory 

authority under RCRA. The TWRS-P facility was to be added to an existing site-wide RCRA 

permit. 

Other issues associated with environmental radiation protection include such topics as mixed 

waste (which involves dual regulation), pollution prevention, waste minimization, and 

contamination control. 

DOE/RU responses and NRC comments can be grouped into the following categories: (1) 

compliance demonstration, operating procedures, and administrative procedures; and (2) 

American National Standards Institute (ANSI)/ISO-14001 as an adequate implementing 

standard. 

55 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "U.S. 

Nuclear Regulatory Commission Comments on BNFL's Tank Waste Remediation Systems Safety Criteria for 

Environmental Radiological Protection Program and for Environmental Radiological Monitoring." February 11, 

1999. 

56 Gibbs, D.C., U.S. Department of Energy, letter to R.C. Pierson, U.S. Nuclear Regulatory Commission, March 17, 

1999.

121 NUREG-1747
The TWRS-P Contractor had started submitting permit applications at the time of the 

termination. These were generally based upon the design as it existed in late calendar year 2000. Significantly, the approach intends to store failed melters (i.e., melters removed from 

service due to refractory age, etc., but not necessarily failed as a breach) during the operational phase of the facility, followed by near-surface disposal during the decommissioning period. The 

NRC staff did not review the specific details of this approach. However, based upon meetings in the Spring of 2000, the approach relies upon movement of intact melters in overpacks, with 

total weights of several hunrdred tonnes. Some vitrified waste materials would remain within the 

failed melters. For the failed HLW melter, the great majority of the vitrified waste would have to 

be removed in order to meet near-surface disposal requirements. It was not clear how this 

could be accomplished for a melter without bottom drains. 

3.9.5 NRC ASSESSMENT' OF THE PRESENT STATUS 

The NRC staff concludes th'ii: considerable effort and specificity remain to be done in this 

environmental area. This should be achievable concurrent with the advancement of the design

NUREG-1747 122
3.0 AREAS OF REVIEW 

3.10 QUALITY ASSURANCE AND MANAGEMENT MEASURES 

3.10.1 INTRODUCTION 

The U.S. Department of Energy (DOE) has responsibility for the regulatory oversight of the 

Hanford Tank Waste Remediation System-Privatization (TWRS-P) contract. Should that 

responsibility be transferred to the U.S. Nuclear Regulatory Commission (NRC) at some future 

time, the responsibility for oversight of the TWRS-P quality assurance (QA) program for items 

relied on for safety (IROFS) would be part of the overall responsibility transferred. The 

application of QA to TWRS-P is a fairly complex subject, and the transfer of the QA regulatory 

oversight will also be complex. The complexities stem from the facts that: 

1. DOE has regulations and guidance for QA that govern its facilities, and NRC has its own 

regulations and guidance for QA for NRC-licensed facilities. 

2. The nuclear industry (represented by the American Nuclear Society) standard for QA, 

American Society of Mechanical Engineers (ASME) NQA-1, "Quality Assurance 

Requirements for Nuclear Facility Applications,""7 undergoes continual change. 

3. NRC's regulations and guidance for QA and safety management measures for facilities 

licensed under 10 CFR Part 70 are currently being revised to address IROFS and a safety 

analysis. 

This chapter discusses the DOE and NRC requirements for QA for TWRS-P and clarifies the 

status of the regulatory approach that has been taken. 

3.10.2 DOE QA REQUIREMENTS AND GUIDANCE 

The QA of DOE nuclear facilities such as TWRS-P is currently regulated by the DOE QA rule58. 

By letter of September 8, 1998 (D. Clark Gibbs to Maurice J. Bullock), DOE informed BNFL Inc. 

(the DOE TWRS-P contractor) that: 'The Contractor's processes necessary to implement ISM 

(Integrated Safety Management) are.., subject to, and the RU (DOE's Regulatory Unit) will 

evaluate them against, the requirements of the QA rule." 

DOE has issued a generic guide 59 for implementing its QA rule that provides guidance as to 

what DOE considers an acceptable QA program. This guide includes many references. Before 

the references, the guide states that 'The following references provide acceptable methods for 

implementing many of the requirements of 10 CFR 830.120.... No single reference fully meets 

57 ASME NQA-1 is a consensus American National Standard issued every 3 years (with annual updates) by the 

American Society of Mechanical Engineers. 

'8 Code of Federal Regulations, Title 10, Energy, Part 830, "Nuclear Safety Management," includes the U.S. 

Department of Energy Quality Assurance rule, (10 CFR 830.120, "Quality Assurance Requirements"). 

59 Department of Energy (U.S.)(DOE). DOE G 830.120, Draft, "Implementation Guide for Use with 10 CFR Part 

830.120 and DOE Order 5700.6C, Quality Assurance." September 1997.

123 NUREG-1747
all of the requirements. The principles, recommended approaches, and applications contained 

in these references may be used in conjunction with 10 CFR 830.120...to develop an effective 

management system to achieve quality." 

One of the references listed in the generic DOE guide is ASME NQA-1-1994. ASME 

NQA-1 -1994 has incorporat,,;d the ASME NQA Standard requirements of the previous editions 

of ASME NQA-1 and ASME- NQA-2 60 into a single document: Part I of ASME NQA-1 -1994 

contains the prior ASME NQA-1 Standard requirements and Part II contains the prior ASME 

NQA-2 Standard requirements. Part Ill of ASME NQA-1-1994 contains the ASME NQA 

Standard guidance. 

In addition, DOE has issued a specific guidance document for its QA reviewers to use when 

reviewing TWRS-P QA programs.61 This guidance was based on the principles in the DOE QA rule and its generic guidanc,: document. This TWRS-P specific guidance document was used 

by DOE and DOE contracto' personnel when reviewing the BNFL Inc. QA program description. 

3.10.3 NRC GA REGULATIONS AND GUIDANCE 

Prior to this year, NRC regulations for domestic licensing of special nuclear material (SNM) require QA only for licensees who possess and use SNM in a plutonium processing and fuel 

fabrication plant.62 However,, a near-criticality incident at a low enriched uranium fuel fabrication 

facility prompted the NRC to review its safety regulations for applicants/licensees that possess and process large quantities of SNM.63 As a result of this review the NRC staff recognized the 

need to revise its regulatory base for these licensees, particularly for organizations such as 

those that possess and process a critical mass of SNM. Therefore, the NRC has approved rulemaking 64 to amend 10 CFR Part 70, "Domestic Licensing of Special Nuclear Material." QA and management measures are addressed for wider application in the revision to 10 CFR 

Part 70. 

The NRC QA regulatory requirements for domestic licensing of SNM apply to 10 CFR Part 70 

SNM applicants/licensees authorized to possess and process a critical mass of SNM. 

Additional baseline design cril:eria (requirements) are included for new facilities such as the 

planned TWRS-P facility (if and when applicable) and for new processes at existing facilities. 

60 American Society of Mechanical Engineers (ASME). NQA-2, "Quality Assurance Requirements for Nuclear 

Facility Applications, An Americanil National Standard." ASME: 1994. 

61 Department of Energy (U.S)(DOE). RL/REG-96-01, Revision 0, "Guidance for Review of TWRS Privatization 

Contractor Initial Quality Assurance Program." DOE: Richland, Washington. October 1996. 

62 Code of Federal Regulations, Titje 10, Energy, Part 70, "Domestic Licensing of Special Nuclear Material," (specifically 10 CFR 70.22, "Content of Applications," and its footnote reference to Appendix B of 10 CFR Part 50.) 

63 The results of this review are documented in: Nuclear Regulatory Commission (U.S.)(NRC). NUREG-1324, "Proposed Method for Regulating Major Materials Licensees." NRC: Washington, D.C. February 1992. 

64 Nuclear Regulatory Commission (U.S.)(NRC) . SECY-99-147, "Proposed Rulemaking-Revised Requirements for the Domestic Licensing of Special Nuclear Material." NRC: Washington, D.C. June 2, 1998. This is available on 

the web at htti://www.nrc.aov/NRC,/COMMISSION/SECYS/secyl999-147/1999-147scy.html.

NUREG-1747 124
The revised 10 CFR Part 706" indicates that QA is a part or subset of management measures 

as indicated in the 10 CFR 70.4 definition: 

Management measures means the functions performed by the licensee, generally on a 

continuing basis, that are applied to items relied on for safety, to ensure the itmes (sic) are 

available and reliable to perform their functions when needed. Management measures 

include configuration management, maintenance, training and qualifications, procedures, 

audits and assessments, incident investigations, records management, and other quality 

assurance elements. (Emphasis added.) 

The following requirement regarding management measures in 10 CFR 70.62, "Safety Program 

and Integrated Safety Analysis," is from the revised 10 CFR Part 70: 

(d) Management measures. Each applicant or licensee shall establish management 

measures (including configuration management, maintenance, training and qualifications, 

procedures, audits and assessments, incident investigations, records management, and 

other quality assurance elements) to provide continuing assurance of compliance with the 

performance requirements of §70.61. The measures applied to a particular engineered or 

administrative control or control system may be commensurate with the reduction of the risk 

attributable to that control or control system. The management measures shall ensure that 

engineered and administrative controls and control systems that are identified as items 

relied on for safety pursuant to §70.61 (e) of this Part are designed, implemented, and 

maintained, as necessary, to ensure they are available and reliable to perform their function 

when needed, in the context of compliance with the performance requirements of §70.61 of 

this Part. 

The baseline design criteria in the revised 10 CFR Part 70 address management measures in 

the following manner: 

§70.64 Requirements for new facilities or new processes at existing facilities. 

(a) Baseline design criteria. Each prospective applicant or licensee shall address the 

following baseline design criteria in the design of new facilities. Each existing licensee shall 

address the following baseline design criteria in the design of new processes at existing 

facilities that require a license amendment under §70.72. The baseline design criteria must 

be applied to the design of new facilities and new processes, but do not require retrofits to 

existing facilities or existing processes (e.g., those housing or adjacent to the new process); 

however, all facilities and processes must comply with the performance requirements in 

§70.61. Licensees shall maintain the application of these criteria unless the evaluation 

performed pursuant to paragraph (c) of this section demonstrates that a given item is not 

relied on for safety or does not require adherence to the specified criteria. 

(1) Quality standards and records. The design must be developed and implemented in 

accordance with management measures (including configuration management, 

maintenance, training and qualifications, procedures, audits and assessments, incident 

investigations, records management, and other quality assurance elements) to provide 

adequate assurance that items relied on for safety will be available and reliable to perform 

65 Attachment 1 to the proposed rule noted in the previous footnote.

125 NUREG-1747
their function when needed. Appropriate records of these items must be maintained by or 

under the control of the licensee throughout the life of the facility. 

The requirement that each applicant for a new 10 CFR Part 70 license must include a 

description of its management measures is specified in the following section of the latest draft 

of 10 CFR Part 70: 

V70.65 Additional content of applications. 

(a) In addition to the conrtents required by §70.22, each application (for a new Part 70 

license) must include a description of the applicant's safety program established under 

§70.62, including the intilgrated safety analysis summary and a description of the 

management measures (,ncluding configuration management, maintenance, training and 

qualifications, procedures, audits and assessments, incident investigations, records 

management, and other cquality assurance elements). 

From the above, it is obvious that each applicant for a new 10 CFR Part 70 license must: 

1. Establish management measures (including configuration management, maintenance, 

training and qualifications, procedures, audits and assessments, incident investigations, records management, arnd other quality assurance elements) to provide continuing 

assurance of compliance with the performance requirements; and 

2. Develop and implement the facility design in accordance with management measures 

(including configuration management, maintenance, training and qualifications, procedures, 

audits and assessments, incident investigations, records management, and other quality 

assurance elements) to provide adequate assurance that items relied on for safety will be 

available and reliable to perform their function when needed. 

The NRC QA regulatory requirements in the revised 10 CFR Part 70 regulation (quoted above) 

are very general and nonprescriptive. Therefore, to aid the staff's efforts in any technical review 

of a license application that might be received in the future under the revised 10 CFR Part 70, the NRC staff has drafted or completed several different Standard Review Plans (SRPs) for 

different types of fuel facilities. For example, an SRP with QA regulatory guidance for NRC 

reviewers to use when revie--ving the TWRS-P 10 CFR Part 70 license application has been 

completed and released to the public as NUREG-1702. 66 Chapter 11 of NUREG-1702 

addresses Management Measures, and Section 11.3 addresses QA. 

In line with the NRC concepi:s of maintaining safety, reducing unnecessary regulatory burden, 

improving public confidence, and increasing efficiency and effectiveness of NRC processes, Section 11 3 of NUREG-1 70:2 relies heavily on the industry QA standard, ASME NQA-1 -1994. 

While ASME NQA-1 -1994 has separate sections for "requirements" and "guidance," NRC's 

licensing requirements are existent solely in the agency's regulations. Guidance on how the 

NRC's regulatory requirements are to be implemented is provided in separate documents such 

as SRPs, regulatory guides, etc. From an NRC regulatory perspective, none of the parts of 

66 Nuclear Regulatory Commission (U.S.)(NRC). NUREG-1702, Final Report, "Standard Review Plan for the 

Review of a License Application for :he Tank Waste Remediation System Privatization (TWRS-P) Project." NRC: 

Washington, D.C. March 2000.

NUREG-1747 126
ASME NQA-1 would constitute requirements, i.e., would have the force of law, unless they were 

incorporated specifically into the applicable regulation (10 CFR Part 70). This is not apt to 

happen. 

Subsection 11.3.5.1, "Acceptance Review," of NUREG-1702 shows that the 

applicant's/licensee's commitment to implement and maintain its QA program in conformance 

with the applicable ASME NQA Standard requirements of Parts I and II of ASME NQA-1 -1994 

or equivalent should satisfy the acceptance review criteria. 

Further, Subsection 11.3.5.2, "Safety Evaluation," of NUREG-1702 gives the applicant/licensee 

the option of: 

1. Committing to implement and maintain its QA program in conformance with the applicable 

ASME NQA Standard requirements of Parts I and II of ASME NQA-1 -1994 or equivalent 

and addressing a relatively short list of regulatory review criteria under the headings of (1) 

organization, (2) QA function, and (3) provisions for continuing QA 

- OR

2. Not committing to meet the applicable ASME NQA Standard requirements in Parts I and II 

of ASME NQA-1 -1994 but, rather, addressing the same three regulatory review criteria 

noted above plus a relatively long list (15+ pages) of additional regulatory review criteria as 

well. 

The staff believes that proper implementation of a QA program developed under either Option 

1 or Option 2 as allowed in Subsection 11.3.5.2 of NUREG-1702 would meet NRC's QA 

regulatory requirements in the revised 10 CFR Part 70 and would provide reasonable 

assurance that the TWRS-P items relied on for safety will perform satisfactorily in service 

without undue risk to the health and safety of the public or to the environment. 

3.10.4 COMPARISON OF QA REGULATORY REQUIREMENTS AND GUIDANCE 

The QA regulatory requirements in both NRC's proposed 10 CFR Part 70 and in DOE's QA rule 

provide for flexibility in the approaches that may be taken to meet the regulations. Although 

organized and worded differently, the QA regulatory requirements of NRC's revised 10 CFR 

Part 70 and DOE's QA rule and the corresponding guidance are not inconsistent. 

As noted above, NRC's revised 10 CFR Part 70 simply requires that each applicant or licensee 

establish management measures (including configuration management, maintenance, training 

and qualifications, procedures, audits and assessments, incident investigations, records 

management, and other quality assurance elements) to provide continuing assurance of 

compliance with the performance requirements. 

Correspondingly, in addition to providing many more detailed QA requirements that correspond 

to NRC QA guidance, DOE's QA rule requires that (management of) a contractor responsible 

for a DOE nuclear facility develop, implement, and maintain a QA program acceptable to DOE; 

that the facility design incorporate applicable requirements and design bases; and that the 

adequacy of design products be verified or validated.

127 NUREG-1747
Although there are many similarities between the DOE and NRC QA requirements and 

guidance, a QA program could still be developed that meets the QA requirements and guidance of one organization but does not meet the QA requirements and guidance of the other 

organization. For example, DOE's QA rule requires "assessments" while NRC's guidance refers to "audits." The relationship between assessments and audits is not well defined. Items 

like this are not expected to be significant issues/problems for TWRS-P because the BNFL Inc. 

QA program description thai the RU approves as meeting DOE QA regulatory requirements, and that BNFL, Inc. implemrents, is expected to also meet NRC's proposed QA regulatory 

requirements and guidance. 

3.10.5 13NFL INC. QA PROGRAM FOR TWRS-P 

BNFL Inc. first submitted a description of its QA program to DOE on November 6, 1996.67 The 

NRC provided preliminary cornments to DOE regarding that BNFL Inc. submittal on November 

20, 1996,68 but DOE had provided its comments on that submittal prior to receipt/review of the 

NRC comments.6 9 

BNFL Inc. submitted Revision 2 of the description of its QA program to DOE on February 7, 

1997,70 indicating that the revision provided "the agreed upon disposition" of a number of listed 

items. On February 12, 1997, DOE approved Revision 2 "as the BNFL Quality Assurance 

Program for Part A activities.''71 

In Revision 3 of BNFL-5193.-CAP-01 dated March 27, 1998, BNFL Inc. changed the title of the 

document from "Quality Assurance Program" to "Quality Assurance Program and 

Implementation Plan."72 Revision 3 was a major rewrite of the TWRS-P QA program. It 

described the BNFL Inc. QA program for activities up to the start of construction and, as the title 

indicates, the BNFL Inc. plan for implementing its QA program. The NRC provided detailed 

comments to DOE regarding Revision 3 of the (now-called) Quality Assurance Program 

Implementation Plan (QAPIP) in a letter dated April 3, 1998.73 

67 Bullock, M.J., BNFL Inc., letter tco P. Rasmussen, U.S. Department of Energy, with attachment, 

"BNFL-5193-QAP-01, Revision 0, Quality Assurance Program, November 8, 1996." November 6, 1996. 

68 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to T.R. Sheridan, U.S. Department of Energy, 

November 20,1996. 

69 Sheridan, T.R., U.S. Depatmen1 of Enegy and C. Bell, U.S. Department of Energy, fax to R. Pierson, U.S. Nuclear 

Regulatory Commission and J. Spraul, U.S. Nuclear Regulatory Commission, November 20, 1996. 

70 Bullock, M.J., BNFL Inc., letter to T. R. Sheridan, U.S. Department of Energy, with attachment 

"BNFL-5193-CAP-01, Revision 2, Quality Assurance Program, February 7, 1997." February 7, 1997. 

71 Sheridan, T.R., U.S. Department of Energy, letter to M.J. Bullock, BNFL Inc., February 12, 1997. 

72 Bullock, M.J., BNFL Inc., letter t: T. R. Sheridan, U.S. Department of Energy, with attachment 

"BNFL-5193-QAP-01, Revision 3, Quality Assurance Program, March 27, 1998." March 27, 1998. 

73 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, April 3, 

1998.

NUREG-1747 128
On April 30, 1998, the NRC staff QA reviewer participated in a DOE-BNFL Inc. meeting at 

which an April 22, 1998, draft Revision 4 of the BNFL Inc. QAPIP was discussed. The NRC 

comments regarding Revision 3 of the QAPIP were generally included in the discussion. The 

meeting resulted in a second draft of Revision 4 of the BNFL Inc. QAPIP dated May 4, 1998. 

The final version of Revision 4 was submitted to DOE by BNFL Inc. on May 15, 1998.74 It is 

dated May 1998. 

In Revision 4 of its QAPIP, BNFL Inc. states: "the implementation and maintenance of the QAP 

(Quality Assurance Program) shall comply with the applicable elements of... Quality 

Assurance Requirements for Nuclear Facility Applications (ASME NQA-1 1994a) .... " BNFL 

Inc.'s QAPIP also addresses the regulatory review criteria under the headings of (1) 

organization, (2) QA function, and (3) provisions for continuing QA specified in Section 11.3 of 

NUREG-1702. In addition, Revision 4 of the BNFL Inc. QAPIP also addresses the majority of 

the longer list of additional regulatory review criteria in the Appendix of Section 11.3 of 

NUREG-1 702 as well. 

DOE evaluated the May 1998 final version of Revision 4 of the BNFL Inc. QAPIP for TWRS-P 

using the DOE documents listed in Footnotes 2, 3, and 5. On June 2, 1998, DOE approved the 

May 1998 final version of Revision 4 of the BNFL Inc. QAPIP for TWRS-P subject to the 

following two conditions:75 

1. 'The implementing documents and procedures...required prior to start of preliminary design, 

detailed design, and procurement, shall be issued before the start of those respective 

phases of project activity." 

2. "BNFL shall implement the (QA program)...as approved by the RU,...up to the start of 

construction." 

On May 27, 1999, BNFL Inc. submitted its proposed Revision 5 (as Revision 4a) to its QAPIP, 76 

reflecting its annual update of the document and describing its QA program and commitments 

up to the start of construction. NRC staff comments on the submittal were sent to the RU on 

June 22, 1999.77 They were forwarded by the RU to BNFL Inc. on July 1, 1999,78 as an 

enclosure to the letter that stated: 'These comments (as is customary for such NRC 

transmittals) are in the public record and are being provided to you for information purposes. 

74 Bullock, M.J., BNFL Inc., letter to D.C. Gibbs, U.S. Department of Energy and P. Rasmussen, U.S. Department of 

Energy, with enclosure "BNFL-5193-QAP-01, Revision 4, Quality Assurance Program and Implementation Plan, May 

1998." May 15, 1998. 

75 Gibbs, D.C., U.S. Department of Energy, letter to M.J. Bullock, BNFL Inc., with enclosure "DOE Regulatory Unit 

Evaluation Report of the BNFL Inc. Quality Assurance Program and Implementation Plan, May 1998." June 2, 1998. 

76 Burrows, C., BNFL Inc., letter to D.C. Gibbs, U.S. Department of Energy, with enclosure "BNFL-5193-QAP-01, 

Revision 4A, Quality Assurance Program and Implementation Plan, May 26, 1999." May 27, 1999. 

77 Pierson, R.C., U.S. Nuclear Regulatory Commission, letter to D.C.Gibbs, U.S. Department of Energy, June 22, 

1999. 

78 Gibbs, D.C., U.S. Department of Energy, letter to M.J. Lawrence, BNFL Inc., July 1, 1999.

129 NUREG-1747
The Regulatory Unit's comments on the QAPIP, which will include consideration of the NRC 

comments, will be transmittix;d separately." 

The RU transmitted its comments on the proposed Revision 5 to the BNFL Inc. QAPIP as an 

enclosure to its letter of July 12, 1999.79 The letter states: 'The RU will use the BNFL response 

to the comments to complete an Evaluation Report as the basis for approval or disapproval of 

the QAPIP." 

An enclosure to a letter from the RU to the NRC 80 shows the RU disposition of the NRC 

comments. Of the 29 technical NRC comments, 22 were provided to BNFL as RU comments, 

one was provided to BNFL Inc. after modification following clarification by BNFL Inc., three were 

considered to be equivalent to some other RU comment, and the remaining three were 

transmitted only by DOE's July 1, 1999, letter referred to above. Acceptable justification for not 

including these last three NR:1,. technical QA comments as RU comments was given in DOE's 

July 14 letter. All 12 NRC editorial comments were provided to BNFL Inc. as RU comments. 

Thus, the E3NFL Inc. response to the RU comments also addressed the NRC comments except 

as noted. 

Revision 5 of the BNFL, Inc. QAPIP was approved by the RU and issued on April 4, 2000. 

In June 2000, BNFL, Inc. submitted a proposed revision to address construction as well as 

design. The RU was planning to review this document in July 2000. 

Revision 5 of the BNFL, Inc. QAPIP generally meets NRC QA regulatory requirements for 

safety of activities under 10 CFR 70. However, three issues regarding adequacy of the 

interpretation, application or implementation of the QAPIP should be resolved as early in the 

TWRS-P project as possible. These issues are discussed in detail in Attachment A.19 through 

A.20 to this report. 

1. BNFL Inc. has not adequately identified its commitments or interpretations related to the 

applicability of specific NQA-1 requirements, either as full and explicit commitments to NQA 1, nor have exceptions to and applicability of the provisions of the document been identified. 

Implementing procedures for NQA-1 requirements appear not to have been fully and 

completely developed, and there is no explicit commitment to the records storage and 

software requirements of NQA-1. 

2. The appropriate and adEquate implementation by all major team members and 

subcontractors of QA req uirements should be verified prior to or early in the project design activities in order to preclude extensive avoidable retrofits for design, procurement or other 

project activities. 

3. A complete safety categorization of the structures, systems and components (SSCs) has 

not been fully developed by BNFL Inc. or provided to DOE. Without a list of all SSCs for 

79 Gibbs, D.C., U.S. Department ci Energy, letter to M.J. Lawrence, BNFL Inc., July 12, 1999. 

80 Gibbs, D.C., U.S. Department oa Energy, letter to R.C. Pierson, U.S. Nuclear Regulatory Commission, July 14, 

1999.

NUREG-1747 130
each of the safety categories, a logical and systematic graded QA program cannot be 

developed. 

3.10.6 NRC ASSESSMENT OF CURRENT STATUS AND CONCLUSIONS 

BNFL Inc. QA commitments that the RU approved as meeting DOE QA regulatory 

requirements up to the start of construction, generally meet NRC's QA regulatory requirements 

and guidance. Therefore, the NRC staff concludes that the BNFL Inc. QA commitments for 

TWRS-P, when properly implemented, can meet the NRC QA regulatory requirements of the 

revised 10 CFR Part 70 and the NRC QA regulatory guidance in NUREG-1702. Consequently, 

the staff concludes that the transition of TWRS-P from DOE to NRC regulation and oversight 

should be relatively smooth in the area of QA.

131 NUREG-1747
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NUREG-1747 132
4.0 NRC OBSERVATION AND CONCLUSIONS 

ON THE REGULATORY APPROACH 

The basic concept of the U.S. Department of Energy's (DOE's) regulatory approach is that the 

contractor is responsible for achieving adequate safety, complying with applicable laws and 

regulations, and conforming with top-level safety standards and principles stipulated by DOE. 

Consistent with applicable laws and regulations, the contractor is required to tailor the exercise 

of this responsibility to the specific hazards associated with its activities, and is encouraged to 

do this in a cost-effective manner that applies best commercial practices. The Tank Waste 

Remediation System-Privatization (TWRS-P) contractors have the responsibility to identify and 

recommend to DOE the set of standards, regulations, and requirements necessary to ensure 

adequate safety. DOE's responsibility is to execute the regulatory process, including 

authorization of contractor actions and confirmation that the contractor activities are performed 

safely and within approved limits. The authority of the RU to regulate a TWRS-P contractor is 

derived from the terms of the TWRS-P contract ("regulate by the contract"). Section 1.4 

provides more information on the DOE regulatory approach and supporting documents. 

The U.S. Nuclear Regulatory Commission (NRC) staff has participated with DOE in the TWRS

P program for 3 1⁄2/ years. NRC staff have observed interactions between DOE and the 

contractors and the practical effects from the implementation of the regulatory approach. Many 

of these observations have already been discussed with DOE"1 . An overriding concern has 

been the way all activities have been and continue to be driven by programmatic concerns, 

including cost and schedule. Ironically, this emphasis has allowed some areas of the design 

and safety analyses, such as identification of items relied on for safety (IROFS), to remain 

relatively nonspecific after the 4 years of design efforts by the Contractor. 

4.1 PROGRAMMATIC INFLUENCE UPON DOE REGULATORY ACTIVITIES 

The Regulatory Unit (RU) is part of the DOE Richland Operations *Office (RL) and reports to the 

manager of RL (i.e, at the time of contract termination in June 2000). The RL organization 

reports to DOE/Environmental Restoration and Waste Management (EM) in Washington, D.C. 

Initially, the DOE manager for TWRS-P also reported to RL. In 1998 the DOE Office of River 

Protection (ORP) was created to handle the management of TWRS and related activities, 

including TWRS-P. ORP directly reports to DOE-EM with matrix activities to RL. ORP 

manages and administers the TWRS-P contracts. Thus, the RU and the ORP are both parts of 

the DOE-EM organization and influence is unavoidable. As noted by a DOE external review of 

the RU (see Footnote 12, page 28), true regulatory independence may not be achieved by the 

current approach. Finally, RL is a party to the Tri-Party Agreement (TPA), which emphasizes 

schedules with milestones for remediation of the Hanford site (see Chapter 1.0). This has the 

unintended consequence of emphasizing schedule for the RU-"In the context of contract

81 Leach, M.N., U.S. Nuclear Regulatory Commission, letter to D.C. Gibbs, U.S. Department of Energy, "Potential 

Critical Technical Issues for Construction Authorization Resolution," January 21, 2000.

133 NUREG-1747
based regulation, the regulator has an incentive to conduct the necessary regulatory activities in 

such a manner that the project schedule is not unduly delayed.82

" 

The DOE approach regulates, by incorporating top level standards and requirements into the 

TWRS-P contracts. Thus, the RU regulates "by the contract." The TWRS-P contract (Main 

Reference 21) contained many milestones, interactions, and deliverables for program purposes and reviews, but only three [or regulatory purposes: the Design Safety Features (DSF), the 

Topical Meetings, and the Construction Authorization Request (CAR). In addition, the RU 

made numerous statements that they could only regulate to the contract and frequently referred 

to the schedule of deliverables it contained. The actual program contract officers for DOE are 

in the ORP, and the Contractors appear to focus more on feedback from the program side than 

from the RU/regulatory side. 

DOE has encouraged the Contractor to pursue an approach with a large processing facility containing several melters (the largest proposed melters for radwaste use in the world), close to 

two million gallons of liquids, and potentially tens of megacuries of activity. This emphasis on a 

much larger facility accrues from program issues and TPA commitments. Regulatory and 

safety issues associated with a much larger facility do not appear to have been considered. In 

contrast, the approaches at £)efense Waste Processing Facility (DWPF) and the West Valley Demonstration Project (WVDP) used several years of melter testing and extensive experimental studies and pilot plant testing. The concept of a smaller pilot facility was abandoned during Phase IA at Hanford. Given the many unknowns associated with the different wastes at 

Hanford, it would seem an emphasis on the large facility could encounter numerous problems during the processing of actual waste that could be better addressed in a pilot scale facility. 

Further evidence of programmatic influence upon the regulatory activities are found in the 

review schedules. The reviews of significant deliverables would sometimes overlap or conflict. 

Frequently, the DOE and/or RU would only allocate a very short, limited time for conducting 

significant reviews of major de•liverables. A typical review time would be 2 or 3 weeks for 

multivolume submittals and 's insufficient for adequate review. For example, the RU allowed 

two weeks for the review of the Initial Safety Analysis Report (ISAR) in 1998 and, even though it 

was given adequate notice, diid not appear to actively consider additional NRC comments 

submitted after the 2-week period. The Firm Fixed Price (FFP) submittal totaled seven boxes 

(of interest for design and safety-some 25,000 pages); the active review lasted about 3 

weeks. The review letter from the RU was signed out inside of 1 month and consisted of only 

four pages8 3. 

Sometimes cost and schedule issues also influenced regulatory operations, including reviews. 

Terms such as "cost effective," "return on investment," and "impact to schedule" would 

frequently arise in reports and at meetings discussing safety and regulatory matters. For 

example, there was a discussion at the June, 2000 Topical Meeting about the relative costs of 

82 Gibbs, D. Clark. <d c clark qibbs@rl.qov> to W.J. Pasciak, U.S. Nuclear Regulatory Commission, "Former 

Regulatory Transition Plan," with attachment "RL/REG-99-01, Issues Related to the Potential Regulatory 

Transition of TWRS-P" November 1998. (December 15, 1998). 

83 Gibbs, D.C., U.S. Department oi Energy, memorandum to R.T. French, U.S. Department of Energy, "REG:RAG/00-RU-0395, Review o- 13NFL Inc. (BNFL) Part B-1 Facility and Process Design Deliverables by the 

Office of Safety Regulation (RU)," May 25, 2000.

NUREG-1747 134
ventilation controls and instrumentation. On many occasions, there was an implication that 

regulatory reviews were not allowed to impact cost and schedule. 

Finally, as already noted, the DOE has terminated the TWRS-P contract and is transitioning to 

a Management and Operations (M&O) contract for TWRS. Few of the recent decisions and 

activities on the contracts have actually mentioned safety (see the ORP website via 

www.hanford.gov). The Request for Procurement mentions regulation by DOE but offers 

incentives of 20-30 percent for cost savings; no incentives for safety are mentioned. 

4.2 INADEQUATE MAINTENANCE OF DESIGN AND AUTHORIZATION BASIS 

DOCUMENTS 

Throughout most of the design effort (about 2 years), the design and safety teams of the 

Contractor have worked quasi-independently. The design and authorization basis documents 

have not been updated (prior to contract termination, the plan was to issue updates in late 

2000) and amendment requests have only recently started to be received from the Contractor. 

Changes in fundamental aspects of the design have occurred in this time period without 

regulatory review. This was formally noted after contract termination84 although the changes 

were made about 18 months earlier. DOE subsequently approved these changes after a 

qualitative evaluation". 

The contractor has selected and proposed standards and regulations for TWRS-P, which have 

subsequently been approved by the RU. These, combined with the design, became the "authorization basis." A formal change process has been identified for the standards and 

regulations. However, in practice, it was not clear if these standards and changes are being 

adequately maintained and updated by both the regulator and the Contractor, and if they are 

being adequately complied with by the design teams and reviewers. For example, during Part 

B-i, an inspection of the authorization basis was twice postponed because it was known the 

Contractor was not following procedures and maintaining the design/authorization basis, and 

would not pass the inspection. While a corrective action conference was ultimately conducted 

(see Section 3.1), this occurred much later in March 2000-some 19 months after the start of 

Part B-I. As a further example, in response to an NRC inquiry about revision dates for the 

Safety Requirements Document (SRD), the DOE advised that its version of the SRD was not a 

controlled copy and is updated by revised pages rather than a complete document (see 

Footnote 20 on page 41). In a recent design review on "Pretreatment Ion Exchange Systems" 

in June 2000, the discussion indicated that the design and the design team had not considered 

the standards and requirements in the authorization basis-this was overlooked by the design 

4 Gibbs, D.C., U.S. Department of Energy, letter to P.O. Strawbridge, BNFL Inc., "Contract Number DE-AC27-96

RL13308 - Regulatory Unit (RU) Partial Approval of Authorization Basis Amendment Request (ABAR), ABAR-W375

00-00014, Rev. 0, Part A Hazard Analysis Report (HAR) Significant and Bounding Hazard Evaluation and Initial 

Safety Analysis Report (ISAR) Fundamental Aspects of Design," 00-RU-0455. July 6, 2000. 

'5 Gibbs, D.C., U.S. Department of Energy, letter to P.O. Strawbridge, BNFL Inc., "Contract Number DE-AC27-96

RL13308 - Regulatory Unit (RU) Approval of Authorization Basis Amendment Request, ABAR-W375-00-00014, Rev. 

0, Part A HAR Significant and Bounding Hazard Evaluation and ISAR Fundamental Aspects of Design," 00-RU-0529. 

August 10, 2000.

135 NUREG-1747
review report86. in contrast, the NRC uses a formal process with public participation to codify regulations and establish guidance and licenses. For changes and deviations in the standards 

used or proposed by licensees, the NRC would conduct a comprehensive review of the 

application and documentation incorporating the change. 

The DOE and the RU frequ;ntly accept the standards and regulations proposed by the 

contractor with a limited inde;!pendent review of its applicability to the proposed design. It is not 

clear if DOE is independently assessing the design's compliance with the selected standards 

and authorization basis. The burden is placed upon the TWRS regulator to analyze and 

approve or disapprove the standard. The NRC would conduct a thorough review of any applicant/licensee from NRC: requirements and/or NRC-endorsed industry codes and 

standards. More of the burden is placed upon the license applicant. Significant changes might require NRC reviews considerably longer (up to a year or more) than some of those observed 

during this program and would be based on a more complete level of design. 

The design and authorization basis-the license basis-are spread over several different 

documents of differing vintages and designs. Changes, including modifications in approach, 

are not always well documented or substantiated. For example, the FFP contract is more 

detailed than the ISAR, yet it lacks some features (e.g., some tanks and pumps) and adds 

others (e.g., an extra ion exchange column), without explanation. Various capacities are mentioned in other documents (i.e., lx, 2x, and 4x flow rates) without a clear explanation of the 

bases for some portions of the proposed facility and equipment handling much greater quantities of materials and activity. The authorization basis has not been well defined in 

quantitative parameters, including the flow rate level. This has important ramifications for the safety reviews-obviously a scenario with a release at a 4x flow rate is likely to have a greater 

consequence than one at a 1 x flow rate, and this may influence severity levels and safety 

controls. 

4.3 USE OF A RISK-EIASED APPROACH 

The use of a risk-based approach to the design and risk-based analyses were used as the 

basis for the Integrated Safely( Management (ISM) process, which includes hazards 

identification, consequence estimation, and control mitigation with limited additional 

considerations. This is essentially a completely fluid process without a basal level of requirements and, as presently practiced, does not consider unknowns, uncertainties, errors, 

proven practices, future plans, and experience. As practiced in TWRS-P, there appears to be 

more emphasis on the process and less on the results. 

The ISM approach at TWRS-P is a circular process (see Section 1.4). A clear, central concept of ISM is that the contractor :.should tailor the design and safety requirements to the specific hazards of the activities and operations at a facility. DOE/RU policy endorses tailoring via the 

following process: 

1. Identify applicable requirements. 

2. Define the scope of the work or operations to be analyzed. 

86 Gilbert, R.A., U.S. Department cf Energy, memorandum to Regulatory Unit Staff, U.S. Department of Energy, "Design Review Report-April-June 2000 Design Reviews," August 4, 2000.

NUREG-1747 136
3. Analyze the hazards. 

4. Propose, analyze, select, and implement controls. 

5. Perform the work or operations (does not apply at the design stage). 

6. Assess, feedback, and improve/modify (as appropriate). 

As the ISM reviews cycle, the concept is for the safety controls to become tailored to the 

hazards. However, an unintended consequence of this approach is the potential elimination of 

margins, conservatism, and defense-in-depth in the design and safety bases because minimal 

requirements are not incorporated. This can obscure minimum standards and performance 

requirements commonly found to be effective from experience and lessons-learned. In 

addition, as the ISM "cycles," assumptions about source terms, releases, etc., are challenged 

and may be reduced. In practice, the assumption can sufficiently reduce consequences to 

place a scenario(s) into less severe categories. Since the ISM approach focuses on the higher 

risk scenarios, once a scenario's consequence is reduced, the assumptions may not be 

revisited. Without a minimal level of requirements, this circular logic may result in fewer safety 

controls and more risk from the proposed facility. The preliminary nature of the design and Part 

B-1 further compounds these concerns as more conservatism would seem to be needed when 

less design information is available. DOE has also experienced difficulty communicating the 

ISM approach to the contractors. For comparison, the NRC uses a risk-informed, performance

based approach with defense-in-depth, appropriate levels of conservatism, and a minimum set 

of standards and requirements that are codified in the regulations. 

The TWRS-P regulatory approach has requirements for reviewing conservatism and defense

in- depth (DID - see CAR guidance in Chapter 7.0, Main Reference 14). DOE most recently 

has expressed concerns about too much conservatism, in public statements on the FFP 

submittal87. In regards to the same FFP submittal, DOE has also noticed the level of 

conservatism and design margins are not mentioned and had concerns about spares and 

redundant equipment and instrumentation. These concerns about conservatism and 

redundancies are further elaborated upon in a more detailed letter88. As the ISM has been 

practiced, most high severity scenarios have just two safety control approaches. 

The DOE and RU acknowledge that there are unknowns and uncertainties associated with the 

program and these should be considered in the reviews of submittals. However, as a practical 

matter, little effort has been given into defining this further and only limited feedback has been 

provided to or received from the contractors. At the present time, the focus continues to be 

upon best-basis and average conditions for many of the analyses. The use of average 

conditions without addressing uncertainties and incorporating reasonable conservatism and DID 

is not likely to be acceptable to the NRC. Furthermore, there does not appear to be a 

realization of future effects. These include site changes (e.g., other privatization initiatives, 

relaxation of security) and facility effects (e.g., higher flow rates desired, wear and aging of 

structures, systems, components (SSCs), erosion/corrosion effects, process changes). 

87 Inside Energy, May 15, 2000. 

88 Gibbs, D.C., U.S. Department of Energy, letter to P.O. Strawbridge, BNFL, Inc., "Regulatory Unit (RU) Review of 

the BNFL Inc. Part B-1 Facility and Process Design Deliverables, (FFP Submittal)," 00-RU-418. June 2, 2000.

137 NUREG-1747
The RU expects the facility, design to be consistent with a risk level of circa 1 E-5/yr to the 

worker and circa 1 E-6/yr to the public. At face value, these are approximately comparable to 

the NRC policy statements on risk and lower than the risk levels implied by the new 10 CFR 

Part 70. However, the public doses are estimated at a minimum distance of the site boundary (12 km to the receptor) and, sometimes, further. Nonconservative values are used for input data. The application of AI.kRA (as low as reasonably achievable) to the design appears limited. The RU has allowed the use of a chemical safety standard that increases the potential risk and allows the existence of a relatively high probability, high consequence event. The NRC 

would likely ask for more substantiation of the values and approaches used with appropriate 

levels of conservatism and redundancy. 

The lack of a clear approach contributes to confusion regarding the level of conservatism, 

particularly for a preliminary design. Using the NRC SRP for TWRS (NUREG-1702) as a guide, it is unlikely that the current design would have adequate conservatism and DID if it were 

undergoing a licensing process with the NRC. 

4.4 LIMITED USE OF' NRC REGULATIONS AND GUIDANCE 

The DOE regulation of TWFRS-P primarily uses documents and guidance written by DOE and its 

contractors; specific to Hanford and the tank wastes (see Section 1.3). Limited use is made of 

NRC guidance and regulations apart from the principles of good regulation. For example, the 

NRC has revised its 10 CFR Part 70 regulations which would be used if the TWRS-P were to 

transition to NRC regulation. The NRC has published in final form a standard review plan for 

TWRS-P facilities (Main Reference 15). These can be used for either one-step (combined construction and operating) or two-step (construction first, followed by a separate operating submittal) licensing. Howevear, the TWRS/River Protection Project does not plan to use either 

in the regulation of these facilities. In addition, there is a desire to avoid nuclear reactor-related 

regulations, guidance, codes, and standards, without fully evaluating the applicability and 

appropriateness of these reactor areas of review. Given the highly radioactive nature of the 

materials that would be in the proposed TWRS-P facility, it would seem that some reactor 

regulations and guidance might be applicable. 

There are also differences in ':he NRC approach. While the NRC does not usually participate in 

discussions and meetings with potential licensees at a preliminary design level, it occasionally does (e.g., the proposed mixed oxide fuel facility). The NRC does not develop standards via a 

contractual, licensee developed process. The NRC process for developing regulations invites 

stakeholder participation (e.g., the recent 10 CFR Part 70 revisions) and, once the regulations 

are developed, approved and promulgated, the NRC expects all of the affected licensees to 

abide by the regulatory requirements. The related guidance might include values or methods 

for determining source terms and release fractions, an area where there have been significant variations in TWRS-P-for example, the RU initially suggested the use of lower radionuclide 

concentrations after encouraging the contractor to go with higher ones. The NRC generally does not develop customizec. regulations and standards for each licensee. This approach helps to maintain consistency in regulatory and safety matters. 

Unlike many other DOE programs, there did not appear to be any routine reports (monthly, 

quarterly, or annual) that definitively discussed the current design, safety features, and the 

evolution thereof. The standard approach of issuing a draft followed by a final version (after 

comment resolution) a few months later has not been followed for major reports. For example,

NUREG-1747 138
the FFP submittal represented the first complete design "report" after the ISAR, corresponding 

to a chronological spacing of some 28 months. Revisions and updates of major deliverables 

are not timely; the first SRD revision occurred one year after the initial version, and a revised 

Hazards Analysis Report will occur 3 years after the original version. Similar communication 

issues exist with Safety Analysis Reports-the time period between the ISAR and the 

CAR/Preliminary Safety Analysis Report will be at least 3 years. The RU has closed items from 

previous reviews after discussions with the contractor. However, it is not clear if these are 

actually closed in the usual sense of the word (i.e., fully addressed now) or adequately 

addressed in an SER-like (safety evaluation report) method. Frequently, "closure" is used to 

denote agreement on an approach to address the issue, not necessarily closure itself. 

Sometimes a clear course of action or a commitment by the contractor is not apparent. The 

DOE program managers in ORP did not appear to actively communicate with the RU on a 

routine basis and at a functional level. The NRC has found that clear documentation, 

communication, and tracking are beneficial for regulatory reviews. 

DOE has used subcontractors extensively for reviews. For example, approximately 50 percent 

of the RU budget is due to subcontractors. Relatively few people in the DOE/RU/contractor 

environment appeared to have a background compatible with greenfield design and regulation 

of a new process and facility. The subcontractors have performed significant roles in the 

reviews for large submittals and have developed and written guidance. At times, DOE has 

performed more in an oversight role over their subcontractors than in some of the reviews. In 

contrast, the NRC tends to rely less on subcontractors.

139 NUREG-1747
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NUREG-1747 140
5.0 POTENTIAL ISSUES FOR TRANSITION TO 

U.S. NUCLEAR REGULATORY COMMISSION REGULATION 

The U.S. Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE) 

have previously discussed issues related to the potential regulatory transition of Tank Waste 

Remediation System-Privatization (TWRS-P) to NRC regulation in the near future. Many of 

these issues are summarized in Main Reference 17 and have been discussed between the 

DOE and NRC over the 3 year length of the program. DOE is converting the contracts to an 

Management and Operations (M&O) arrangement for TWRS, which has been renamed the 

River Protection Project-Waste Treatment Plant (RPP-WTP). The NRC staff believes that 

enabling legislation from Congress is required for the NRC to regulate either a privatized 

TWRS-P facility or an RPP-WTP, and that any resulting issues are resolvable. From the 

viewpoint of the NRC staff, most of the issues would be addressed by the legislation that 

enables NRC regulatory authority over the TWRS/WTP facilities or NRC external regulation of 

DOE facilities, and by continued refinement and detailing of the proposed facility designs. The 

remaining issues relate to DOE programmatic activities and not regulation. The following 

sections summarize the issues and the preliminary assessment of the NRC staff; Attachment D 

provides a discussion of the issues in more detail. 

5.1 EMERGENCY PLANNING 

The TWRS-P Contractor defined an approach in accordance with DOE guidance. There were 

several issues related to the integration of the TWRS-P emergency plan (EP) with the Hanford 

site EPs and organizations (e.g., the Hanford Fire Department) which were being worked upon 

at the time of contract termination. It is anticipated that the transition and subsequent new 

M&O contractors will address these integration issues. From the point of view of the NRC staff, 

if NRC were to regulate the RPP-WTP facility, it would be expected that the integration issues 

will be fully resolved, communication and responsibilities will be clearly defined, and the WTP 

facility EP manager will have the authority to direct the Hanford site EP as necessary based 

upon the integrated safety analysis (ISA). 

5.2 REGULATORY AUTHORITY AND CLASSIFICATION OF RADIOACTIVE WASTES 

As designed by the Contractor, the TWRS-P facility would have generated vitrified low activity 

waste (LAW) and facility wastes, such as failed melters and high efficiency particulate air 

(HEPA) filters. Long term storage of these wastes was planned at the Hanford site. If the 

decision were made for disposal at the Hanford site, existing waste management and disposal 

areas would be used, subject to the wastes meeting the applicable disposal unit's waste 

acceptance criteria. These disposal facilities are regulated by DOE for radioactive components 

and by the State of Washington for hazardous constituents. If the NRC were to regulate RPP

WTP, the vitrified LAW and facility wastes would be regulated by the NRC for their radioactive 

components while they are located at the facility. The NRC would likely designate the vitrified 

LAW and facility wastes as incidental wastes, i.e., incidental to high level waste (HLW) 

processing. After the contractor has returned the wastes to DOE, regulation of the radioactive 

component would revert to DOE provided incidental waste criteria are met. The NRC and DOE 

have discussed the handling of incidental wastes and criteria several times on other HLW 

programs; basically, the incidental waste was considered to be suitable for near-surface

141 NUREG-1747
disposal if 10 CFR Part 61 criteria for low level waste (LLW) are met. RPP-WTP incidental 

wastes are likely to meet those criteria, and hence, the issue should be easily resolved. 

5.3 SAFEGUARDS AND SECURITY 

The physical protection requirements for various classifications of nuclear materials are very 

similar between the DOE and the NRC. Both recognize the differences in the quantities and 

attractiveness characteristics of the materials in the application of the requirements. In 

addition, for special nuclear material (SNM) at a TWRS facility, the facility design (multiple confinement layers and cells), the dilute concentrations of SNM, and the highly radioactive 

nature of the matrix (waste) containing the SNM contribute to its physical protection. DOE has 

categorized the tank materials as waste and removed it from the safeguards program. The 

NRC believes that operatiors at the proposed TWRS facilities and wastes from specific tanks 

may result in the accumulation or de facto separation of SNM, and, thus, might require safeguards. Consequently, -the NRC has the position that the issue should be revisited once 

the design is further along (,say at the Construction Authorization Request/Preliminary Safety Analysis Report (CAR/PSAR) stage) and when more operating details are available (for 

example, at the Operations Authorization Request stage). Given the similarity between NRC 

and DOE safeguards requirements, the NRC does not see this as a significant issue. 

5.4 REGULATION OF EMISSIONS 

Under the current regulatory approach, air emissions from the facility are regulated by the State 

of Washington. If the NRC were to assume regulatory authority for TWRS/WTP, the NRC 

would regulate portions of air emissions. Since there is precedence for this approach (both the 

State and NRC regulate emissions at the Siemens facility), the NRC staff does not view it as a 

significant issue. 

5.5 CO-LOCATED WORKER 

Section 3.4.2 presents more information on the co-located worker (CLW). In summary, the 

DOE is using the concept of CLW for the proposed TWRS/WTP facilities; essentially the CLW 

is allowed the same dose and risk limits as the workers and all members of the public, visitors, 

etc., to the site are considered "general employees" and are allowed to accumulate 

occupational exposure. For 'TWRS-P, DOE plans to use a large portion of the Hanford site for 

this purpose, which typically results in minimum distances to members of the public (i.e., with 

lower dose and risk limits) of approximately 10 miles. In contrast, the NRC focuses on the 

concept of "controlled area" by the facility operator; at NRC facilities, this is usually associated 

with a fence although the controlled area can extend beyond the fence and even beyond the 

site boundary. Thus, the NRC approach usually corresponds to a shorter distance (usually 100-200 meters) to the public for accident analysis purposes, which can translate into more 

items relied on for safety (IRFOFS). However-, the key concept is the TWRSIWTP operator's authority to exercise control over the Hanford site emergency plans; if such authority is granted 

by DOE to the TWRS/WTP operator, then the "controlled area" concept may be satisfied, and 

the issue is moot. 

In addition, the dose limits re&quire comparison. Under the TWRS-P/privatization regulatory 

program, the dose limits were 25 rem for the worker, CLW, and the public for highly unlikely 

events (the public had a target dose goal of 5 rem). In DOE, highly unlikely corresponds to the

NUREG-1747 142
frequency range of 1 E-4/yr to 1 E-6/yr. From the perspective of the NRC staff, the revised 10 

CFR Part 70 and standard review plan (SRP) have a worker dose limit of 100 rem and a public 

dose limit of 25 rem for high consequence events, and corresponding limits of 25 rem and 5 

rem for intermediate consequence events. High consequence events are to be rendered highly 

unlikely (1 E-5/yr or less in frequency) by safety controls, and intermediate consequence events 

are to be rendered unlikely (in the 1 E-2/yr to 1 E-5/yr frequency range) by safety controls. Thus, 

at face value, the DOE limits are more restrictive although the fuel cycle SRP 89 does allow 

grading of the frequency limit in inverse proportion to the magnitude of the consequences and 

this could result in the limits overlapping. Again, the NRC staff considers this to be a resolvable 

issue. 

As an aside, the NRC notes the DOE regulatory approach should consider future site changes 

planned that are likely to reduce the distances to the public for accident evaluation purposes 

(see Issue A.13 in Attachment A). 

5.6 WASTE OWNERSHIP 

As part of the privatization contracts, DOE retained ownership of the waste materials. Part 70 

of 10 CFR does not require transfer of ownership. Thus, assuming the new contract(s) still 

require ownership of the wastes by DOE, the NRC staff do not believe there is an issue. 

5.7 PAYMENT FOR TRANSITION TO NRC REGULATION AND NRC REGULATORY 

OVERSIGHT 

NRC involvement in TWRS-P has been funded via a line item in the budget. Future regulation 

of a TWRS/WTP facility by the NRC would require legislation that would also identify the 

funding mechanism, be it by line items, fee collection, DOE payments, or a combination thereof. 

It is anticipated that the NRC regulatory costs would be a small fraction of the actual DOE 

expenditures to the contractors on the tank waste programs. The NRC staff does not see this 

as a significant issue. 

5.8 TRI-PARTY AGREEMENT 

DOE is a party to the Tri-Party Agreement (TPA) and is responsible for TPA commitments. The 

NRC staff would expect the NRC to remain a non-party to the TPA consistent with other 

regulators (e.g., Defense Nuclear Facilities Board (DNFSB), Washington Department of Health 

(WDOH)). The NRC staff does not see this as a significant issue. 

5.9 DOE STOP WORK AUTHORITY 

The DOE currently has stop work authority for safety concerns under the contract. The NRC 

staff anticipates both the NRC and DOE would have stop work authority for safety concerns if 

regulatory transition occurred, and the staff does not consider this to be a significant issue. 

89 Nuclear Regulatory Commission (U.S.)(NRC). NUREG-1 520, "Draft Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility." NRC: Washington, D.C. 2000.

143 NUREG-1747
5.10 DOE SAFETY OVE'RSIGHT AFTER REGULATORY TRANSITION 

The DOE currently is the req:lulator for the proposed TWRS/WTP activities. After transition, the 

NRC would be the regulator. The NRC staff anticipates the DOE would want to perform some 

safety oversight activities in a manner analogous to a corporate headquarters unit providing oversight to an operating facility. The NRC staff does not consider this to be a significant issue, 

5.11 APPLICATION C)F NRC (10 CFR PART 2) HEARING REQUESTS 

The potential regulation of the proposed TWRS/WTP facilities by the NRC would probably 

invoke the NRC Rules of Practice (i.e., 10 CFR Part 2), which allow for public hearings. 

Hearings require scheduling and can take time. However, licensing under 10 CFR Part 70 

would focus hearings on more specific issues and be relatively expeditious. In addition, 

legislation authorizing NRC regulation of TWRS/WTP might dictate a schedule for hearings. Furthermore, if the facility is well along into construction or even starting operations at the time 

of transition to NRC regulation, a certification process (for existing facilities) may be more 

appropriate. Certification wc,uld allow the activities at the plant to continue while the hearings 

and other reviews are comp! ,eted. Consequently, whether the route is licensing, Congressional mandate, or certification, the NRC staff does not foresee a significant schedule impact from the 

hearing process, and, thus, Joes not consider this to be a significant issue. 

5.12 CONTRACTUAL OBLIGATIONS FOR FEED DELIVERY 

DOE will continue to deliver radioactive feed material to the Contractor. Presumably, DOE will 

agree to abide by the direction of the NRC-regulated Contractor as to the rate and 

characteristics of this feed material. The NRC staff expects there will be specifications and 

safety requirements for the wastes and the plant operations. If the wastes cannot be blended, 

then an amendment process could be pursued by the licensee (an amendment process has 

already been implemented al the gaseous diffusion plants (GDPs). This is not an issue for 

regulatory transition. 

5.13 COMMUNICATIOCN PLAN WITH STAKEHOLDERS 

The NRC anticipates that the. existing DOE communication plans (e.g., with the Hanford 

Advisory Board) would continue during and after transition to NRC regulation, as they are 

separate from the regulatory processes. In addition, the NRC also maintains public communication and openness as part of its regulatory practices. Therefore, the NRC staff does 

not consider this a significant issue. 

5.14 COST BENEFIT OF NRC REGULATION 

The NRC staff considers NRC regulation of TWRS/WTP facilities to be a national policy issue 

and not decided by costlbene,'it analyses. The NRC staff does not consider this a transition 

issue.

NUREG-1747 144
ROLE OF THE DEFENSE NUCLEAR FACILITIES SAFETY BOARD

The NRC has determined that legislative authority is required for it to regulate TWRS/WTP 

facilities. As part of such legislation, the NRC staff anticipates that regulatory roles will be 

clearly defined and based upon precedent at the enrichment and spent nuclear fuel (SNF) 

storage facilities that the NRC regulates at DOE sites; the staff does not expect a role for dual 

NRC/DNFSB oversight of TWRS/WTP. The NRC staff does not consider this an issue. 

5.16 PRICE-ANDERSON INDEMNIFICATION 

The NRC anticipates the legislation enabling NRC regulation will define any Price-Anderson 

indemnification requirements. The NRC staff does not consider this an issue. 

5.17 TIMING TO AVOID DELAYS FROM REGULATORY TRANSITION 

DOE wishes to select the timing for transition so as to reduce the potential for disruption to a 

minimum. Some timing considerations tend to conflict. In programmatic terms, it would be 

beneficial for DOE that transition occurs sometime after full and confirmed operation so that it 

could be assured that the plant meets its specifications. On the other hand, to minimize 

potential retrofit changes in meeting NRC's regulatory expectations, it might be preferable to 

make the transition as soon as possible. (This can also be accomplished by resolving all 

regulatory differences prior to construction and transitioning later.) 

The NRC staff believes that there are three alternatives for the timing of regulatory transition: 

1) immediately, 2) overlapping, or parallel, regulatory transition process during design, or 3) 

regulatory transition during operations. Alternatives one and two correspond to licensing of the 

vitrification plant, for which rules and guidance already exist. Alternative 3 represents a 

certification route. 

The NRC staff believes the legislation establishing NRC regulation of TWRS/WTP will define 

the timing for transition. 

5.18 DOE FINANCIAL INTERESTS 

DOE has expressed concerns about protecting its financial interests in the vitrification facility. 

DOE has already invested several hundred million dollars in the Phase I activities. The 

completion of the design and construction activities in Phase I may amount to another $4 

Billion. Thus, DOE has considerable financial interests in the program. The NRC has no 

financial interests to protect. The NRC staff does not consider this a regulatory transition issue. 

5.19 QUALITY ASSURANCE 

The DOE TWRS Quality Assurance (QA) program requirements and the TWRS-P contractor's 

DOE-approved QA programs may need to be modified in order to facilitate the transition to 

NRC regulation. The NRC staff believes the differences in QA requirements are small and the 

adequacy of the QA program really depends upon its implementation. The NRC staff does not 

consider this a significant transition issue.

NUREG-1747

5.15

145
5IMPACTS ON OTHER HANFORD SITE FACILITIES

DOE is concerned that there;. may be impacts upon non-TWRS/WTP facilities at Hanford after regulatory transition to the NIRC. The technical impacts to non-TWRS-P facilities of TWRS-P regulatory transition would have to be identified, evaluated, and addressed on an individual 

basis as such issues arise. Coordinating the transition to external regulation of the tank farm 

storage and vitrification facilities will likely reduce the impacts to the non-vitrification portion of 

the tank farms, which is the non-TWRS/WNTP facility most susceptible to these impacts. The 

NRC staff does not consider this an issue for regulatory transition. 

5.21 RESPONSIBILITY FOR OCCUPATIONAL SAFETY 

For its facilities, DOE has regulatory authority for occupational safety. For NRC licensed facilities, the Occupational Safety and Health Administration (OSHA) has such authority. If the 

TWRS/WTP transitioned to l\IRC regulatory oversight, OSHA may assume regulatory authority for occupational safety. The NRC staff anticipates that the enabling legislation for NRC 

regulation would clearly define the responsibilities. 

5.22 TAILORING OF RE'QUIREMENTS 

The NRC and DOE use diflerent approaches for the tailoring of requirements for safe mission 

performance. This issue concerns the different approaches used by DOE and NRC for 

establishing requirements for achieving adequate safety and reconciliation of any differences in the approaches to facilitate seamless transition from DOE to NRC regulation, should transition 

occur at a future date. As discussed in more detail in Attachment D, even though the DOE 

approach to establishing requirements is somewhat different that the NRC approach, the requirements resulting from either approach will likely achieve adequate safety, and significant reconciliation may not be neicessary. The NRC staff believes this issue is not significant and 

disappears if regulatory tranm;sition occurs in the near-term. 

5.23 ACHIEVEMENT OF ADEQUATE SAFETY 

DOE and NRC conclusions cn achievement of adequate safety may be different and need reconciliation to facilitate possible future transition of the TWRS/WTP project to NRC 

regulation. The DOE regulatory approach for TWRS-P safety (radiological, nuclear, and process) places on the TWRE;-P Contractor the responsibility to achieve (a) adequate safety for the workers and the public (b) comply with applicable laws and legal requirements, and (c) conform to DOE stipulated top-level safety standards and principles.90 Within this regulatory framework, the TWRS-P Contractor identifies the work needed, evaluates the associated 

hazards, selects appropriate standards, and justifies the adequacy of the selected standards." 

The NRC has expressed the following concerns about the process for TWRS-P: 

90 Department of Energy (U.S.)(D)(IE). RL/REG-97-10, Rev. 1, "Radiological, Nuclear, and Process Safety 

Regulation of TWRS Prvatization Contractors Regulatory Plan." January 1998. 

91 Department of Energy (U.S.)(DOE). DOE/RL-96-0004, Rev. 0, "Process for Establishing a Set of Radiological, Nuclear, and Process Safety Standards and Requirements for TWRS Privatization." 

February 1996.

NUREG-1 747

5.20

146
1. Safety requirements were proposed and conditionally approved during the pre-conceptual 

design phase of the facility. This attempt to reach judgment on adequate safety was 

performed prior to the majority of design information being available. 

2. For both regulatory processes, the determination of adequate safety is somewhat subject to 

engineering judgment, and as such, should transition occur, additional requirements and 

potential backfits could be needed. 

The NRC staff believe this is a technical issue that requires additional design effort by the 

contractors. 

5.24 EFFECT OF SCHEDULE DELAYS ON THE OVERALL RISK 

DOE has expressed interest in balancing the risks posed by continued operation of the tanks 

with the risks posed by operation of the vitrification plant, regardless of the regulatory 

framework, as a way of avoiding schedule delays. In theory, the requirements could be 

different (and less) against a baseline of an existing hazard which could be reduced by the 

operation of the facility. From the viewpoint of DOE, early startup, high availability, and high 

capacity operation of the vitrification plant could serve to reduce the risks posed by the tanks, 

even if regulatory requirements on the vitrification plant were relaxed. 

NRC requirements have been established to minimize risks of operation of a facility relative to a 

baseline of no-operation, even though NRC can consider balancing risks within a facility. 

However, NRC operates to codified regulations (10 CFR Part 70) and, while provision for risk 

benefits and balancing can be made via the risk-informed, performance-based approach, it can 

be difficult to manage and balance the different risks on a quantitative basis with limited 

information on the design and safety features of the TWRS/WTP facility. Regulatory 

requirements could be administered with recognition of the real risks from the degrading tanks 

as baseline risks. 

The NRC staff anticipates that the scope of NRC regulation-the TWRS/WTP facilities, tank 

farms, other Hanford facilities-will be defined in the enabling legislation. A single regulatory 

entity offers consistency and the potential to consider total risks from tank waste management. 

However, absent specific design and quantitative analysis, it is not possible to assess balancing 

total risk for the tank waste/HLW related systems at Hanford and achieving adequate 

assurances of safety is likely to be based upon individual facility analyses. 

5.25 RESOURCE ALLOCATION ACROSS TWRS-RELATED FACILITIES 

DOE has expressed concerns about balancing the resources available for the safety features of 

the TWRS/WTP (the vitrification facilities in particular) across the entire TWRS complex. The 

NRC is concerned with adequate assurances of safety for the workers, the public, and the 

environment. Once safety and regulatory requirements are met for TWRS/WTP, resource 

allocation is an internal DOE matter. The NRC staff expects the scope of NRC regulation to be 

defined in the enabling legislation.

147 NUREG-1747
5.26 BACKFIT/RETR01I"IT

Changes may be required as the TWRS/WTP facility transitions to NRC regulation. The 

TWRS/WTP facility could be; subjected to new or modified requirements imposed by NRC if regulatory transition occurs. These new or modified requirements may require that the 

TWRS/WTP facility be modified. Terms such as "backfit" and "retrofit" are commonly used to 

describe modifications that are made in response to new or modified requirements. The NRC anticipates that sorne backfits/retrofits may be necessary upon transition of the 

TWRS/WTP to NRC regulatory authority. However, the NRC staff expects these to be minor 

because the facilities would represent new design and construction with complete documentation and the NRC. has participated with DOE and their contractors in Phase IA and Phase lB-1 of the program. Regulatory transition of other TWRS activities might involve older 

facilities with less documentation and could result in more significant requirements. 

5.27 SCOPE OF NRC REGULATION 

DOE has discussed the scops of NRC regulation, including facilities beyond the immediate 

TWRS/WTP facilities. 

Cases exist or will exist on DOE sites where NRC regulates one facility and DOE regulates other facilities, although these have much simpler interfaces than at TWRS. The TMI-2 

Independent Spent Fuel Storage Installation (at the DOE Idaho Site) is an example of a case 

where NRC will license and regulate a new facility that will receive waste from an existing facility, specifically the Test Area North (TAN) facility. In this case, DOE regulates current TAN facility spent fuel pool storaCge operations and will regulate dry storage cask loading operations performed at the TAN facilibt. NRC regulation will begin when the fuel leaves the TAN facility boundary in an approved transportation cask. 92 In another case, at the Portsmouth and 

Paducah GDPs, certain operations have been certified by NRC while others remain under DOE 

control. This shared site arrangement required that regulatory boundaries be established to 

ensure a clear understanding of responsibilities for regulatory oversight and the rules governing particular portions of the planiis. 93 

The NRC staff anticipates that the scope of NRC regulation-the TWRS/WTP facilities, tank farms, other Hanford facilities-will be defined in the enabling legislation. A single regulatory entity offers consistency and the potential to consider total risks from tank waste management. However, absent specific design and quantitative analysis, it is not possible to assess balancing total risk for the tank waste/F-IIW related systems at Hanford and achieving adequate 

.assurances of safety is likely 1:o be based upon individual facility analyses. 

92 Department of Energy (U.S.)(DOE). Section 1.2, Rev. 0, "License Application for the Idaho National Engineering 

Laboratory Three Mile Island Unit Two Independent Spent Fuel Storage Installation, DOE Idaho Operation Office." DOE: Washington, D.C. October 1,996. 

93 Buhl, A.R., T. Murley, G. Edgar and D. Silverman. "NRC Regulation of DOE Facilities." Nuclear News, p. 32. 

May 1997.

NUREG-1747 148
5.28 LOCATION OF THE REGULATOR

DOE believes the regulator should be located near the Hanford site. The NRC staff anticipates 

that NRC regulation of TWRS/WTP facilities would involve some staff FTEs (full time 

equivalents) at the Hanford site, some staff FTEs at the Region IV Office, and some staff FTEs 

at NRC Headquarters. The actual numbers would be determined by Congress via the enabling 

legislation and the scope of NRC regulation it requires. 

5.29 MULTI-STEP LICENSING 

The program for TWRS/WTP is split into several phases, and a multi-step licensing process 

may be more appropriate. The revised 10 CFR Part 70 allows multi-step licensing. An 

amendment process would allow the constructed TWRS/WTP facility the opportunity to obtain 

approval for changes, different flow capacities, and different radionuclide inventories. This is 

not an issue. 

5.30 DIFFERENCES IN 10 CFR PART 20 AND 10 CFR PART 835 

DOE believes differences in 10 CFR Part 20 and 10 CFR Part 835 need to be understood and 

reconciled to the extent necessary. For the most part, DOE and NRC radiation protection 

regulations are similar, bur a few significant differences exist. The most notable difference is 

that requirements found in 10 CFR Part 20 are more comprehensive that the 10 CFR Part 835 

requirements, but where differences in dose values exist, the DOE requirements are usually 

more conservative. Both agencies also agreed that the key to developing a 10 CFR Part 20 

compliant Radiation Protection Program (RPP) lies in the details of the implementation of the 

Safety Requirements Document (SRD) Safety Criteria for radiation protection. The TWRS-P 

Contractor had not developed the implementation details. It was expected that the level of 

detail needed to better evaluate compliance with 10 CFR Part 20 would become available at the 

PSAR stage when the TWRS-P Contractor would have submitted a revised RPP for 

construction. It is expected that a similar approach will be taken by the new contractors working 

under M&O arrangements. 

The NRC staff does not consider this a transition issue. 

5.31 COST OF CONTRACTOR DOCUMENTATION 

DOE expects there may be some costs associated with changing documentation to meet NRC 

expectations after regulatory transition. Regulation by the NRC may require additional 

documentation, different formats, more specificity, or additional analyses. It would be expected 

that if the TWRS/WTP contractors were in compliance with the DOE regulatory framework at 

the time of transition to NRC regulation, the costs for revision of new plant documents (all of 

which would likely be available in electronic formats) would be small. Given the use of M&O 

contracts, DOE would fund these documentation requirements and changes. 

The NRC staff does not consider this a significant transition issue.

149 NUREG-1747
THE ROLE OF I)C'E ORDERS VERSUS NRC RULES

DOE is concerned that there may be misunderstandings regarding the applicability and 

enforceability of DOE Orders and their relationship with DOE rules. 

In the transition of regulatory oversight from DOE to the NRC for the GDPs, the NRC presumed that the GDPs were required to strictly comply with DOE Orders. However, that was not the 

case. Thus, the role of DOE Orders in the conduct of DOE business requires clarification. The 

term "Order" or more precisa;y, "Safety Order" is a misnomer and leads one to a basic 

misconception in understanding the DOE regulatory process. The terminology problems can be 

exacerbated because the N1RC regulatory structure also includes the term "Order" but it is used 

in the more conventional sense. NRC Orders have the force of law and are issued to licensees 

as mandatory requirements. Contrary to what the term "Order" implies, DOE Safety Orders are 

not, on their own, legal requirements and are not necessarily mandatory. 

Order-related lessons learned from the GDP transition are not generally applicable to TWRS. 

DOE Order compliance was specificalfy excluded from the TWRS-P contract. 94 Both the DOE 

and NRC have been involved in developing the TWRS-P regulatory (Integrated Safety Management Plan [ISMP] and SRD) and design (ISA and Safety Analysis Report [SAR]) bases. 

The regulatory basis for the new contracts has not been defined yet. 

The NRC staff does not consider this an issue. 

5.33 THE ROLE OF NATIONAL ENVIRONMENTAL POLICY ACT IN REGULATORY 

TRANSITION 

DOE believes the potential effects of National Environmental Policy Act (NEPA) need to be 

considered for regulatory transition. Since the TWRS/WTP project is a major Federal action, DOE has prepared an environmental impact statement (EIS). The EIS considers the actual 

cleanup activities and opera:ions of the TWRS-P facilities and includes its construction, 

operation, and decontamina",ion and decommissioning activities. 

The NRC staff does not consider this an issue because of the existing NEPA coverage and the 

similarities between the two ;igencies in their approaches to NEPA 

5.34 ROLE OF THE AC:RS OR ACNW IN REGULATORY OVERSIGHT TRANSITION 

DOE is concerned by the potential effects of the Advisory Committee on Reactor Safeguards (ACRS) and/or the Advisory Committee on Nuclear Waste (ACNW) on the transition of 

TWRS/WTP regulation from DOE to NRC. 

The ACNW would likely take an interest in the transition of regulatory responsibility associated 

with TWRS-P. A review of ACNW reports from July 1996 to the present suggests that a typical ACNW recommendation would be expected within 3 months of an initial presentation of the 

issue to the Committee. Sorne issues, that were not time sensitive, were not concluded for up 

94 Department of Energy (U.S.)(D- E-) Rev. 1 "Radiological, Nuclear, and Process Safety Regulation of TWRS 

Privatization Contractors," Regulatory Plan. DOE: Richland, Washington. January 1998.

5.32

NUREG-1747 150
to a year, while reports on other issues were released within 1 week of the presentation to the 

Committee. Records of recent meetings suggest that ACNW involvement in the TWRS-P 

regulatory oversight transition process would not result in any delays. ACNW participation in 

the process would occur in parallel with other transition/licensing activities. 

The NRC staff does not consider this to be a significant issue because no delays are 

anticipated. 

5.35 IMPACT ON FUTURE GENERATIONS 

DOE is interested if there would be any impacts from regulatory transition upon future 

generations and resources. The premise for the TWRS-P vitrification plant is that it will take 

waste, process it, and return a stable vitrified waste form for stable, long-term storage and 

disposal. By doing so, DOE intends to remediate the Hanford waste tanks and, eventually, in 

the mid-21st Century,9" return the land to beneficial us as it was prior to 1940. Future 

generations in the latter half of the 21 st Century would benefit by greater land resources than 

are available now. External regulation of DOE facilities, including TWRSIWTP facilities, is 

expected to improve the safety and protection of workers, the public, and the environment. 

This can only have a positive effect upon the future. 

The NRC staff does not see this as an issue for regulatory transition. 

5.36 PUBLIC INPUT TO REGULATORY DECISIONS 

DOE is concerned that regulatory decisions may not continue to be open to the public and 

public input after transition of regulation to the NRC. The transfer of TWRS-P to NRC 

regulatory oversight would result in the NRC licensing or certifying the TWRS-P Contractor's 

facility. In any licensing action, NRC is required to obtain public participation under 10 CFR 

Part 2, the Rules of Practice. In this process, the interaction is more formal and judicially 

governed than in the DOE case. NRC generally conducts interactions with the licensee in the 

public domain to the maximum extent possible. The ongoing pre-licensing activities on the 

mixed oxide project are a good example. The NRC staff does not consider this an issue as 

NRC practices and procedures usually result in more openness with the public than DOE 

regulatory actions. 

95 The Tri-Party Agreement specifies complete clean-up of tank wastes by 2028, following which time the tanks 

themselves and all appurtenances would require decontamination and decommissioning and disposal.

151 NUREG-1747
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NUREG-1747 152
6.0 FUTURE ITEMS, EVENTS, AND CONCERNS 

The U.S. Department of Energy (DOE) has stated its intention to maintain the schedule as it 

transitions to the new contractor and Management and Operations (M&O) contracting 

arrangement. The original schedule at the time of this writing was (December 2000): 

RPP: Radiation Protection Program for Design, Construction, and Operation. 

LCAR: Limited Construction Authorization Request - due at the end of June 2000. 

Seismic Issues: Subject of the Topical Meeting, end of July 2000. 

ISMP: Integrated Safety Management Plan - revision due at the end of 

August 2000. 

HAR: Hazards Analysis Report - revision due at the end of September 2000. 

SRD: Standards Requirement Document - revision due at the end of October 2000. 

CAR: Construction Authorization Request - due early February 2001. 

No revised schedule for these regulatory events is currently available.

153 NUREG-1747
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NUREG-1747 154
7.0 Main References

1. Pacific Northwest Laboratory (PNL). PNL-1 0773, "Hanford Tank Cleanup: A Guide to 

Understanding the Technical Issues." Third Printing, (Gephart, R.E. and R.E. Lundgren). 

PNL: Richland, Washington. February 1997. 

2. BNFL Inc. BNFL-5193-HAR-01, Rev. 0, "Hazards Analysis Report (HAR)." BNFL Inc.: 

Richland, Washington. 1997. 

3. BNFL Inc. BNFL-5193-SRD-01, Rev. 0, "Safety Requirements Document (SRD)." BNFL 

Inc.: Richland, Washington. September 26, 1997. 

4. BNFL Inc. BNFL-5193-ISP-01, Rev. 4, "Integrated Safety Management Plan (ISMP)." 

BNFL Inc.: Richland, Washington. 1998. 

5. BNFL Inc. BNFL-5193-ISAR-01, Rev. 0, "Initial Safety Analysis Report (ISAR)." 

BNFL Inc.: Richland, Washington. 1998. 

6. Lockheed Martin Advanced Environmental Systems (LMAES). TWRS-P-CD-08, Rev. 0, 

"Safety Requirements Document (SRD)." LMAES: Richland, Washington. July 22, 

1997. 

7. Lockheed Martin Advanced Environmental Systems (LMAES). TWRS-P-CD-010, 

Rev. 2, "Hazards Analysis Report (HAR)." LMAES: Richland, Washington. January 

23, 1998. 

8. Lockheed Martin Advanced Environmental Systems (LMAES). TWRS-P-CD-065, 

Rev. 1, "LMAES ISMP, in Reference 6 and Safety Requirements Document Report." 

January 23, 1998; and TWRS-P-CD-09, Rev. 2, "Integrated Safety Management Plan 

(ISMP)." January 23, 1998. LMAES: Richland, Washington. 

9. Lockheed Martin Advanced Environmental Systems (LMAES). TWRS-P-CD-061, 

Rev. 0, "Initial Safety Assessment Package" and TWRS-P-CD-01 6, Rev. 0 "Initial Safety 

Assessment Report." LMAES: Richland, Washington. January 1998. 

10. U.S. Department of Energy (U.S.)(DOE). RL/REG-98-09, Revision 0, "DOE Regulatory 

Unit: Initial Safety Evaluation Report of the BNFL Inc. Initial Safety Assessment." DOE: 

Richland, Washington. March 1998. 

11. U.S. Department of Energy (U.S.)(DOE). RLIREG-98-10, Revision 0, "DOE Regulatory 

Unit: Initial Safety Evaluation Report of the LMAES ISA Package." DOE: Richland, 

Washington. March 1998. 

12. K.D. DeTienne and A.D. Edmonson, "Government Fair Cost Estimate River Protection 

Project - Privatization, Phase I, Part B," Rev. 0, April 2000

155 NUREG-1747
13. BNFL Inc. "Firm Fixed Price Submittal: Part B-1 Facility and Process Design Deliverables; Deliveirables B-1 -14, B-1 -15, B-1 -16, and B-1 -17 on Current Facility and 

Process Design." BNFL Inc.: Richland, Washington. April 25, 2000. 

14. U.S. Department of Energy (U.S.)(DOE). RL/REG-99-05, Rev. 2, "Guidance for the Review of the TWRS.P Construction Authorization Request (CAR)." DOE: Richland, 

Washington. January 2000. 

15. Nuclear Regulatory Commission (U.S.)(NRC). NUREG-1702, Final Report, "Standard 

Review Plan for the Review of a License Application for the Tank Waste Remediation 

System Privatization (TWRS-P) Project." NRC: Washington, D.C. March 2000. 

16. BNFL Inc. "Design Safety Features, RPT-W375-RU00001, Rev. 0,, Contract DE-AC06

96RL1 3308." BNFL Inc.: Richland, Washington. February 24, 1999. 

17. U.S. Department of Energy (U.S.)(DOE). RL/REG-99-01, Draft, "Issues Related to the 

Potential Regulatory Transition of TWRS-P." DOE: Richland, Washington. 

November 1998. 

18. Nuclear Regulatory Commission (U.S.)(NRC). "Process Safety of Proposed Tank 

Waste Remediation Systems (TWRS)," (Murray, A.P., M.N. Baker, L.W. Chang, and 

M.S. Srinivasan). NRC: Washington, D.C. May 1999. 

19. Nuclear Regulatory Commission (U.S.)(NRC). "Materials Considerations of Proposed Tank Waste Remediation Systems (TWRS)," (Srinivasan, M., A.P. Murray, and 

M.N. Baker). NRC: Washington, D.C. March 2000. 

20. U.S. Department of Energy (U.S.)(DOE). Contract Number DE-AC06-RL1 3308, ''WRS Privatization." DOE: Richland, Washington. September 9, 1996. 

21. U.S. Department of Energy (U.S.)(DOE). Contract Number DE-RP06-96RL13308, 'TWRS Privatization." DOE: Richland, Washington. August 1998. 

22. Nuclear Regulatory Commission (U.S.)(NRC). NUREG/CR-6410, "Nuclear Fuel Cycle Facility Accident Analysis Handbook." NRC: Washington, D.C. March 1998. 

23. U.S. Department of Energy (U.S.)(DOE). HDBK-3010-94, "Airborne Release 

Fractions/Rates And Flespirable Fractions For Nonreactor Nuclear Facilities," Volumes 

1 and 2. DOE: Richland, Washington. December 1994.

NUREG-1747

156
APPENDIX A 

SUMMARY OF ISSUES 

Unresolved from the NRC perspective as of June 1, 2000. 

Al. LEVEL OF DETAIL 

Statement of the Issue 

The documents describe the proposed facility, buildings, processes, and operations at an early 

stage of design, nominally 14 percent according to the Contractor (the basis for identifying this 

level of design has not been provided). However, the descriptions are frequently general and 

there are many places where the information is labeled as "TBD" (to be determined) or to be 

determined by future program activities. Items relied on for safety (IROFS) are not consistently 

identified and described in these reports. The integrated safety management process appears 

to be at a preliminary draft stage. 

Specific Example(s) 

The Firm Fixed Price (FFP) contract contains numerous equipment lists and data sheets that do 

not specify either the materials, construction requirements, or the safety significance. The FFP 

system descriptions likewise do not identify the IROFS and their performance requirements for 

many systems. 

U.S. Department of Energy Position 

The U.S. Department of Energy (DOE) believes the facility and process design information in 

the FFP submittal falls far short of the information required for the Preliminary Safety Analysis 

Report (PSAR). "Significant additional design information is required" (May 25, 2000 letter to 

DOE/Office of River Protection [DOE/ORP] from the Regulatory Unit [RU]). 

U.S. Nuclear Regulatory Commission Concerns 

The U.S. Nuclear Regulatory Commission (NRC) would expect an application to focus on the 

salient design features and their safety implications, and their required performance 

requirements for IROFS. While this does not necessarily correspond to a set percentage of 

design or design level, sufficient information would be anticipated to allow the verification of the 

calculations and conclusions, and the identification of safety parameters. This would 

subsequently allow the determination of adequate assurances of safety. 

NRC Evaluation and Assessment 

At the present time, it would appear that the information lacks the level of detail that the NRC 

would normally anticipate in an application (e.g., see the standard review plan (SRP) (NUREG

1702)), and, as such, it would be unlikely that the NRC would accept such a document as an 

application from a potential licensee. Given the magnitude of the safety information 

requirements, the frequent design changes, and the ongoing DOE contractual changes, it is 

unlikely that the detailed safety information will be available for sometime.

157 NUREG-1747
A2. REASONABLE CONSERVATISM AND ADEQUATE ASSURANCES OF SAFETY 

Statement of the Issue 

Many of the calculations presented to date appear to be based upon best bases or some form of average/mean values that represent actual process design points. Based upon a review of the FFP submittal, some of these best basis values may be an order of magnitude or more, 

lower than values allowed by the contract. 

Specific Example(s) 

The FFP contains two mass balances that indicate feed concentrations one order of magnitude lower for cesium-137 and other radionuclides, as compared to the upper bound of the contract 

range. Sodium values are in the 3M range as compared to a IOM upper bound in the contract. There are many systems with single point failure modes in the proposed design (e.g., single pumps in the evaporators) that would require shut down and drainage of the lines/vessels prior to maintenance. 

DOE Position 

In the May 25, 2000, letter from the RU to ORP, DOE states, "The level of conservatism and 

design margins are not mentioned for operations or safety parameters." However, DOE has challenged the contractor about conservatism in design in prior Topical Meetings and related 

discussions. 

NRC Concerns 

The usual approach followed in licensing commercial nuclear facilities incorporates conservatism and margins into design activities so that flexibility is maintained and adequate levels of safety are assured. In addition, the usual engineering practice is to compensate for less design detail and more uncertainty by increasing the degree of conservatism in the safety analyses. Furthermore, the emphasis on cost (see item 6) has resulted in the (nonconservative) reduction of building footprints and cell sizes (i.e., the Contractor has assumed that this will reduce the program cost). However, this is likely to have unintended 

consequences that may actually increase costs (e.g., via maintenance); for example, multiple repositioning of manipulators during a task because of the tight space. Standard NRC guidance, such as the SRP and the Risk-Informed, Performance Based (RIPB) approach, would expect some level of conservatism in an application. The NRC would likely require incorporation of a level of conservatism into the design in order to maintain adequate 

assurances of safety. 

NRC Evaluation and Assessment 

It would be anticipated that a reasonable amount of conservatism would need to be 

incorporated into the design and safety analyses.

NUREG-1747

158
A3. RISK BASED DESIGN APPROACH

Statement of the Issue 

BNFL Inc. appears to be incorporating risk and probabilistic approaches into many aspects of 

the design process, and appears to be conducting the equivalent of a Level 3 probabilistic risk 

assessment (PRA) (i.e., estimates of reliability/failure probabilities, release source terms, and 

dose consequences). However, the equivalent of Level 3 PRAs are difficult to do and usually require a completely designed facility supported by historical operating information. Such a 

level of design and operational information is not available in the existing documents and is not 

likely to exist until significant additional design efforts are completed or the plant is constructed 

and operated. The approach also places a higher level of credence in the analytical methods 

and results than appears warranted. Furthermore, this approach appears to introduce 

substantial additional levels of difficulty into the design process and subsequent regulatory 

reviews. 

Specific Example(s) 

The Integrated Safety Management (ISM) 2 report and Seismic PRA reports both use PRA type methods. They are based upon an earlier design than the FFP submittal. Calculations and 

control strategies are based upon the numerical results without consideration of standard 

approaches (e.g., hydrogen monitoring for flammable gas accumulation, weld radiography). Proposed controls vary considerably from one iteration to the next. 

DOE Position 

DOE has endorsed a risk-based approach without consideration of baseline criteria or minimum 

standards. 

NRC Concerns 

Few baseline design criteria appear to be used in the design (see NUREG-1702). In the recent 

NRC white paper entitled, "Risk-Informed and Performance-Based Regulation," it is stated that, "... the use of PRA technology should be increased in all regulatory matters to the extent 

supported by the state of the art in PRA methods and data, and in a manner that complements [not replaces] the NRC's deterministic approach and supports the NRC's traditional defense-in

depth philosophy." The white paper notes that the NRC does not endorse an approach that is 

risk-based or make decisions based solely upon the numerical results of risk assessments. 

Stated succinctly, the combined RIPB approach uses risk insight, engineering analysis, and 

judgement, including the principle of defense-in-depth and the incorporation of safety margins and performance history. 

NRC Evaluation and Assessment 

The risk-based approach being taken by DOE and BNFL Inc. in the Tank Waste Remediation 

System-Privatization (TWRS-P) design process is a significant departure from that endorsed by the NRC. Such an approach is not likely to be acceptable in the NRC regulatory arena. 

A4. REDUNDANCY AND DEFENSE-IN-DEPTH

159 NUREG-1747
Statement of the Issue

In the current facility design, some systems have multiple lines to provide for redundancy, while 

other systems have fewer lines or redundant flow paths or even single lines with single point failure modes. For example, the Process Vessel Ventilation System has three 100 percent fans 

with emergency power-one fan running, one fan on standby, and one fan assumed to be 

unavailable due to maintenance. However, the cell ventilation systems have three 50 percent fans (two running, one in standby). Some pumps have duplicates. However, the evaporators (i.e., in two separate areas--low activity waste (LAW) receipt and Melter Feed Preparation) 

represent single point failure nodes in their areas, with one 100 percent pump each; 

maintenance requires drain down and flushing as there are no isolation valves. The normal 

process industry approach would have three pumps for each intended use. 

Another possible single failure vulnerability in the design is the single 7-day fuel oil tank serving the three emergency diesel generators (EDGs). The current system description for the EDGs 

indicates that there is a 2-hour fuel supply for each EDG. If the duration of the loss-of-offsite 

power extends beyond 2 houirs, the EDGs (as implied by the system description) would only be 

supplied fuel from the single 7-day tank which does not meet the single failure criterion and 

Institute of Electrical and Electronics Engineers (IEEE) Std 308 (referenced in the current 

Safety Requirements Documents [SRD]). Also, the current system description for the 4.16 kV 

power distribution system states that dividing loads into separate load groups is not a design criterion. This is counter to the redundancy requirements needed to support safety functions 

such as 3-100% trains of process vessel exhaust fans per the current ISM Summary. 

Based upon the provided information, some IROFS may not have adequate redundancy (e.g., cooling water coils). Dedicated spare tanks do not appear to have been incorporated into the 

approach-for example, only a single cesium concentrate tank is shown in the current design, 

even though it might contain tens of megacuries of activity. Supporting information is either 

nonexistent or appears insufficient to justify the differences in approaches. It appears that data 

for design and safety analysis originate from different sources and contexts, and apparently are 

not checked for compatibility including basic assumptions. This issue may be the result of an 

incomplete safety classification of the structures, systems, and components (SSC) at the 

current time as described in item 5 of these issues. To assure that the facility will be capable of 

planned operation and that excessive maintenance will not be needed, the NRC would 

anticipate that these deficiencies would be addressed in the near-term. 

A5. DESIGN/AUTHORIZATION BASES, CONCEPT EVOLUTION, DESIGN CHANGES, 

AND INTEGRATION 

Statement of the Issue 

It appears that safety analyses are being performed using designs and/or bases that may be 

incompatible or have changed. For example, while it was anticipated that the design would 

meet the standards identified earlier during the hazard analysis and standards identification 

process, decisions have been made on specific design characteristics/components, with the 

intent of later modifying the standards to match the design. 

Significant design changes have occurred (e.g., extra tanks, changes in the buildings, less 

shielding) and may be continuing to occur, often without documentation of an explanation. In

NUREG-1747 160
other words, the design basis (and "Authorization Basis") have been changed but the 

analyses-and particularly the safety analyses-were using different bases and justifications. 

In addition, the Contractor did not appear to be following the SRD and its change process, and 

was not maintaining the Authorization Basis current. 

Many SSCs have been and are being designed at nominal service conditions and not for design 

basis conditions that envelope anticipated, off-normal and accident conditions. Some 

parameters have limited margin. 

Specific Example(s) 

Currently, many system descriptions for electrical systems and the ISM Summary pertaining to 

instrumentation and control (I&C) components do not reference the appropriate IEEE standards 

listed in related SRD safety criteria. The facility's inventory has been significantly increased due 

to the addition of the LAW tanks (the "six-pack" of 400 kgal per tank) and the storage of 

ultrafiltered high level waste (HLW) solids, yet this has not been reflected in the Authorization 

Basis over the past eight months. There is little documentation linking design changes (such as 

the elimination of the sulfate removal step) and their impacts from 1 month to the next. The 

FFP submittal is the first design update since the Initial Safety Analysis Report (ISAR)-a 2

year time period. The ISM Cycle 2 submitted one month after the FFP submittal is based upon 

an earlier design. 

DOE Position 

DOE and BNFL, Inc. have stated that there is an ongoing "tailoring" effort to take exception to 

specific SRD standards and to delete and/or substitute all or parts of standards. Although this 

"tailoring" is going on, there is little formal documentation which evaluates the adequacy of the 

exceptions. 

As the result of a a special inspection in December 1999, the RU issued a Corrective Action 

Notice (CAN) against the Contractor for failing to adequately maintain the Authorization Basis. 

In response, the Contractor submitted several Authorization Basis Amendment Requests 

(ABAR's) documenting some of the changes. 

NRC Concerns 

The Hazards Analysis Report (HAR), SRD, Integrated Safety Management Plan (ISMP), and 

ISAR are all part of the Authorization Basis. The Contractor failed to make changes in the 

Authorization Basis and to apply to the RU for approval. The design effort continued using 

different standards and bases. A design basis approach is not being followed from either the 

design or regulatory perspective.

161 NUREG-1747
NRC Evaluation and Assessnment 

The NRC would anticipate a more traditional approach to Authorization Basis/Design Basis 

management. For example, using the SRP, it would be expected that the documents would 

display an integrated and consistent level of design for the facility, with consistency in the 

design bases, codes, and standards for the safety structures, systems, and components. 

A6. SAFETY EMPHASIS 

Statement of the Issue 

Cost and schedule are mentioned numerous times as the rationale for considering certain 

design approaches (e.g., the use of passive systems to avoid the cost of emergency/safety designated backups, reducing backup equipment). Thus, in an effort to reduce projected capital costs, design decisionns are being made that may increase safety concerns and 

maintenance issues, and, for that matter, may even ultimately result in increased overall costs. 

Specific Example(s) 

Cost and schedule topics have arisen at essentially every Topical Meeting. In the June 2000 

Topical Meeting, there were several questions about the cost (material and machining) of 

Hastelloy for the melters without a discussion of technical or safety-related 

functions/requirements, etc. 

DOE Position 

DOE/RU frequently broaches the subject of cost and schedule in regulatory and safety meetings. Some acquiescence on cost issues occurs. Frequently, it has been stated by both 

the RU and the Contractor that regulation can only occur by the contract. 

NRC Concerns 

This apparent design emphasis on cost and schedule perspectives dilutes the safety issues and 

may overlook safety requirements, thus requiring additional efforts by the Contractor and the 

regulators that may impact subsequent safety reviews. It is worth noting that the FFP 

submittal and associated decisions have not included reviews of the safety aspects of the 

proposed facility. 

A7. DOSE ASSESSMENT METHODOLOGY INCLUDING DATA SOURCES 

Statement of the Issue 

BNFL Inc. has provided substantial information regarding the methodology (i.e., equations) that 

it intends to use for evaluation of worker, co-located worker, and public doses under both 

normal operating and accident conditions. However, as noted in item 16, there remain 

differences in what are considered acceptable approaches for certain accident scenarios. 

Regardless of the scenario, there are issues concerning (a) the sources of data to be used for 

evaluation of doses and (b) the degree of conservatism to be employed, as well as (c) the 

means by which uncertainty of many parameters would be addressed.

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Specific Examples

1. BNFL Inc. should clearly state the justification for the selection and use of its sources of 

data. A particular concern is the use of data derived from small deviations from normal 

conditions applied to accident conditions, without an assessment of its accuracy, validity, and pedigree. The use of the "Best Basis" inventory in various scenarios should 

be justified as this may not adequately represent all feeds or be adequate for enveloping expected inventories for accident scenarios (i.e., it's nonconservative). Also, as the RU 

has pointed out, BNFL Inc. needs to provide a comparison with DOE-HNDB-3010-94 

and NUREG/CR-6410 if other parameters are used. While the use of the Sellafield 

Database was withdrawn (letter CCN01 1928, Dobson to Gibbs, dated March 10, 2000), the use of BNFL Inc. data for reliability, corrosion, and filling voids in other data has 

been verbally mentioned in the April-June 2000 timeframe. 

2. The dose methodology appears to rely too heavily upon optimistic considerations, 

averages, and probabilities in the consequence estimates that imply but do not identify safety controls; these would be necessary items for review when the Construction 

Authorization Request (CAR) is received. 

3. By letter dated October 25, 1999, BNFL Inc. provided a report RPT-W375-NS00001, 

Rev. 0, "Dose Assessment Methodology Description." The RU responded in a letter 

dated December 20, 1999, which indicated that ". . . the [BNFL Inc.] document did not 

explain how a measure of conservatism in the analysis would be achieved to account for 

the level of uncertainty in the parameters." The TWRS staff agrees with the position stated in this letter and has provided written comments (e.g., internal memo from 

Struckmeyer through Tokar to Leach dated March 8, 2000 and entitled, "Background Information and Comments on TWRS-P Project: 'Dose Assessment Methodology Description,' 'Seismic Probabilistic Risk Analysis Dose Assessment Methodology Report,' and 'Seismic Probabilistic Risk Analysis Methodology Report.'" 

DOE Position 

In letter 00-RU-0292 (Gibbs to Bullock, dated April 6, 2000), the RU accepted the final BNFL 

Inc. dose methodology described in a BNFL Inc. letter dated February 23, 2000. However, 

DOE submitted additional questions for consideration. 

NRC Concerns 

Additional radiological protection issues that were raised during the course of the NRC's 

involvement in the TWRS-P project 

The co-located worker issue: 

1. A difference in regulatory approach between DOE and NRC appears in the concept of a 

co-located worker. 

2. For the TWRS-P facility, in addition to the categories of general employee and 

radiological worker (defined in 10 CFR Part 835) and member of the public (defined in 

both 10 CFR Part 20 and Part 835), a third category known as the co-located worker

163 NUREG-1 747
(CLW), has been defined within the "Top-Level Radiological, Nuclear, and Process 

Safety Standards and Principles for TWRS Privatization Contractors," DOE/RL-96-0006, 

Revision 1, July 1998. The co-located worker is defined in Chapter 6.0, "Glossary," as 

the following: 

An individual within the Hanford Site, beyond the Contractor-controlled 

area, performing work for or in conjunction with DOE or utilizing other 

Hanford Site facilities. 

3. No equivalent concept: exists for NRC licensees. The co-located worker definition 

contains the term "controlled area," which is defined differently for DOE and its 

contractors (10 CFR Part 835) than it is for the NRC and its licensees (10 CFR Part 20). 

10 CFR Part 835: Controlled area means any area to which access is 

managed by or for DOE to protect individuals from exposure to 

radiation and/or radioactive material. 

10 CFR Part 20: Controlled area means an area, outside of a restricted 

area but inside the site boundary, access to which can be limited by the 

licensee for any reason. 

Note that the DOE term "controlled area" is virtually identical to the NRC term "restricted 

area." There is no defined DOE term called "restricted area." 

10 CFR Part 20: Restricted area means an area, access to which is 

limited by the licensee for the purpose of protecting individuals against 

undue risks from exposure to radiation and radioactive materials. 

Restricted area does not include areas used as residential quarters, but 

separate rooms in a residential building may be set apart as a restricted 

area. 

4. The co-located worker concept would be a difficult issue to resolve if regulatory control 

of the TWRS-P contractor were to shift from DOE to NRC. It would require either an 

exception to the NRC regulations or a change in (a reduction of) the exposure permitted 

to persons beyond the controlled area boundary (per 10 CFR Part 835) or restricted 

area boundary (per 101' CFR 20). [To accomplish a reduction in exposure after 

construction of the facility had been completed, it would be necessary to require the 

contractor to invoke one or more of the following remedies: retrofit with additional 

shielding, dilute the radioactive waste feed streams to lower the activity (and thereby 

increase the total volume of vitrified waste), moving co-located workers farther away 

from the contractor's facility, and possibly other methods.] Discussion of this issue 

between the two organizations produced no concession or compromise. When DOE 

concluded that there was little or no likelihood that such a transition would occur, the 

issue was left unresolved from the NRC point of view. 

Distance to site boundary: 

Although the concept of "site boundary" may appear to be self-evident, it is not defined 

in 10 CFR Part 835. A definition is provided by Part 20:

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Site boundary means that line beyond which the land or property is not 

owned, leased, or otherwise controlled by the licensee. 

A consequence of the lack of a definition in 10 CFR Part 835, at least from the NRC 

point of view, is that DOE and/or its contractor can specify a "fence-line" for the 

calculation of dose to the public that is well beyond the boundary that is under direct 

oversight of the contractor. The area between the contractor's oversight and the 

boundary of the DOE-defined controlled area is the realm of the co-located workers. It 

should also be noted that most persons who are granted access within the Hanford site 

boundary will be considered occupationally exposed personnel by DOE, whereas under 

NRC regulations, many such persons would be considered members of the public. 

NRC Evaluation and Assessment 

It is imperative that the necessary justifications, as well as the information lacking in the report, 

be provided at or before the submittal of the CAR, as it has a potentially major bearing on the 

design of the facility. For example, if the dose consequence methodology underestimates the 

consequences and/or frequencies of potential events, IROFS and safety related SSC's may not 

be appropriately identified, and the facility may expose the workers, the public, and the 

environment to potentially excessive risks and hazards. Without this information, the regulators 

will be faced with having to make a decision on whether to approve the construction of a facility, 

based upon the acceptability of the facility design and proposed operations, but in the absence 

of key information needed to demonstrate that acceptability. 

A8. OPTIMISTIC DESIGN ASSUMPTIONS IMPACT ON OPERATIONS AND 

MAINTENANCE 

Statement of the Issue 

The design approach assumes that the process and systems function as designed and 

planned, but some assumptions on operations and maintenance are outside the envelope of 

experimental or literature results. Shielding designs assume a full decontamination result 

without consideration of "bleed-through," carryover, or cross-contamination. It is not clear that 

the design addresses potential safety issues associated with the process not meeting its 

intended parameters (e.g., decontamination factors, concentrations, aging). Consequently, it 

may prove difficult for the facility to carry out planned operations or meet throughput 

expectation, and radiation field accrual due to these effects may increase worker doses and 

safety concerns. Though the regulatory review at a subsequent Safety Analysis Report (SAR) 

stage may show the need for design modifications with subsequent delays, it is also possible 

that these concerns will not be realized until the plant is in operation, resulting in plant derating 

or shutdown until the modifications can be completed.

165 NUREG-1747
Specific Example 

The Contractor uses a "best basis inventory" value for most of the feed concentrations. The 

FFP submittal contains two activity balances with feed concentrations approximately 2 percent 

of that allowed by the contract. From the same submittal and LAW building design reviews, the 

designers assume full cesium decontamination without provision for additional shielding (e.g., 

stronger foundations). The FFP design includes several single point failure nodes (e.g., the 

feed evaporator). 

DOE Position 

DOE/ORP appears to be encouraging this optimistic approach; one way is by emphasizing the 

DOE desire for higher capacities. For example, the FFP submittal includes a short report on 

facility capacity increases thai: does not evaluate any safety or regulatory concerns. This report 

also states the plant can increase the glass production rate of the melters by 100 percent. This 

corresponds to 100 percent availability of the melters and continuous operation beyond testing 

conditions with very high bubbling rates. In addition, increases in carryover into the offgas 

system and maintenance are not evaluated. 

DOE/RU has not assessed this aspect of the Contractor's activities. 

NRC Concerns 

The NRC would anticipate both realism and conservatism in a design submittal. The NRC 

would also expect an integrated approach that evaluates potential safety issues from all 

perspectives, including design, construction, operation, maintenance, replacement, and 

decontamination and decommissioning (D&D). An overly optimistic design may translate into 

significant maintenance and exposure issues. Also, optimism and higher capacity operation 

may overlook safety considerations and change the parameters/bases for the safety analyses 

(e.g., higher carryover of radionuclides to less shielded areas, higher frequency of accident 

precursor events due to more wear), potentially resulting in unanalyzed safety questions. 

NRC Evaluation and Assessment 

Both DOE and the Contractor need to incorporate realistic approaches to the proposed design. 

A9. UNCERTAINTIES 

Statement of the Issue 

All mathematical and computer models have uncertainty. Atmospheric transport, dispersion, 

and consequence models have uncertainties that can be large, approaching several hundred 

percent at times. From the perspective of the proposed TWRS-P facility, these uncertainties 

arise from these main areas: 

1. Uncertainties in the waste analyses. 

2. Variability in the waste feeds and at surge/inventory points in the facility.

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3. Model limitations caused by an incomplete understanding of the phenomena involved in 

the accident scenario or the transport, dispersion, and consequence to the human 

receptors. 

4. Partial modeling of the phenomena or the scenario events and sequences. 

5. Unknowns, uncertainties, or the lack of information in the numerical input required by 

the models. 

6. The impact of assumptions and default input data values. 

7. Design uncertainties (i.e., the facility is not fully designed and is unlikely to be so for a 

considerable time period). 

The Contractor has not considered these uncertainties in the analyses to date. 

Specific Example(s) 

The material at risk calculations primarily use either best basis inventory or specific tank 

inventory data (based upon limited sampling). No uncertainty bounds are carried through the 

calculations. No percentile levels or other statistical bounds are identified. The respirable fraction (RF), airborne release fraction (ARF), and dispersion calculations by the Contractor 

sometimes use distributions in the input data as part of a PRA type of analysis. However, this 

approach confuses time dependence with uncertainty. In addition, the Contractor frequently mentions the mean or average for comparison with regulatory limits (e.g., the seismic PRA 

presentation at the October 1999 Topical Meeting). Inclusion of uncertainties or the 

consideration of uncertainties would exceed the limit, thus implying that adequate assurances 

of safety are not being achieved. 

NRC Concerns 

The NRC is concerned with adequate assurances of safety for the workers, the public, and the 

environment. On an experiential basis, the NRC has used and recommended for use 

conservative parameters and assumptions that would be expected to envelope uncertainties by 

a wide margin and beyond credible questioning of results. This approach might overestimate 

unmitigated doses but it results in the imposition of safety controls that reduce estimated doses 

from nuclear facilities that were significantly removed from any regulatory limits. Thus, 

compliance could be assured and adequate assurances of safety demonstrated relatively easily. However, the emphasis of DOE and the Contractor on the use of realistic values implies much smaller margins that may not accommodate uncertainties and may challenge regulatory limits. Fundamentally, if realistic assumptions and values approach a regulatory limit (say, within an order of magnitude), how is compliance demonstrated, and at what point/margin should as low as reasonably achievable (ALARA) be applied to the design process?

167 NUREG-1747
NRC Evaluation and Assessment

BNFL Inc. and the RU have not identified an approach that addresses these uncertainties. The 

NRC would likely anticipate that uncertainties would be adequately addressed prior to a major 

submittal, in order to avoid additional regulatory reviews and potential delays. 

A1O. LACK OF METHODOLOGY AND "REASONABLE CRITERIA" FOR SELECTING 

APPROACHES AND INPUT VALUES 

Statement of the Issue 

The design and analyses use mathematical models which require the selection of appropriate 

computer codes/algorithms and significant amounts of input data. Different models and input 

produce different results (e.g., dose consequence estimates from analyzing potential 

accidents), which can correspond to different potential severity levels, event frequencies, and 

reliabilities. The results (output) influence safety requirements, including identification of 

IROFS. BNFL Inc. has not clearly established the methodology for selecting models and input 

parameters that are consistent and appropriate for identifying and analyzing safety 

requirements, and that adequately incorporate reasonable conservatism/margin to address 

potential unknowns and unceitainties. 

Specific Example(s) 

The seismic and PRA models require a large amount of input data. When questioned about 

the sources and validity of the! data at the October 1999 Topical Meeting, the Contractor 

responded that the data represented examples for exercising the models. Subsequent 

meetings, reports, and further responses to questions have not elicited any methodology and 

reasonable criteria for selecting models, approaches, and input values. 

DOE Position 

DOE is aware of the problem. 

NRC Concerns 

Standard NRC practice requires validation of models and input data. The NRC would anticipate 

that such validation would have occurred and been documented. Similarly, it would be 

anticipated that a methodology for selecting approaches and models that represent the system 

and safety conditions would exist and be documented. In short, the "GIGO" (garbage-in, 

garbage-out) phenomena needs to be avoided. 

NRC Evaluation and Assessment 

In the absence of such methodology for model selection and validation, and clear criteria for 

data selection, it would not be possible to demonstrate adequate assurances of safety.

NUREG-1747 168
All. DEFINITION OF UNMITIGATED EVENTS

Statement of the Issue 

The current approach does not appear to adequately address the significance of unmitigated 

events and their use in identifying IROFS. Currently, implicit assumptions are made about 

barriers; for example, a cell is assumed to provide certain decontamination factors (DFs) in 

accidents which result in a less severe consequence. A lower consequence usually translates 

into lower reliability and quality assurance/quality control (QNQC) requirements. The DF 

assumption may result in overlooking and not identifying safety controls and their requirements, 

including reliabilities. Without a suitable definition for unmitigated events, safety requirements 

and controls may not be adequately identified and evaluated in safety analyses and submittals 

that might be evaluated under the SRP. 

Specific Examples 

The chemical safety analyses assume the function of diluting, identifying, and containment 

mechanisms, and include their source reduction effect without attributing any safety significance to these mechanisms. The ISM cycle 2 analyses use a cell DF of 10 for leak 

events and an aerosol concentration based upon a restricted volume. The FFP submittal mass 

balances use values for radionuclides that are well below the contract allowables (in some 

cases, 5 percent of the allowable value); mitigating mechanisms are not identified. 

DOE Position 

The RU is aware of the concerns. 

NRC Concerns 

Some of the ISM analyses are now using unmitigated analyses. However, there still is an 

apparent bias by the Contractor ISM review teams to implicitly rate hazards in a mitigated 

manner. In addition, the rationale for the selection of input parameters is not clear. Often, the 

parameters selected appear low for the situation and for regulatory and safety purposes. Thus, 

it is not clear that safety requirements are being adequately identified and categorized. 

NRC Evaluation and Assessment 

The Contractor has previously identified its approach for using unmitigated analyses for safety 

determinations and average parameters for determining mitigated effects and risk. This 

approach has been confused and muddled in many of the submittals received to date. The ISM 

process appears to contribute to this confusion. Future safety-related submittals will have to be 

closely reviewed to verify that the distinction of unmitigated analyses is maintained.

169 NUREG-1747
A12. CRITICALITY ANALYSIS

Statement of the Issue 

Based upon submittal of four interim criticality evaluations, BNFL Inc. has developed its 

criticality safety program by bounding keff for nearly dry waste solids with maximum Pu content 

as specified in the BNFL Inc. Contract, Revision 6. This approach relies upon providing 

controls on the feed and analyzing the process to ensure that no normal operations or process 

upsets could result in formation of a critical mass. BNFL Inc. has not adequately addressed 

how double contingency will be applied to the feed sampling control nor is it clear that BNFL 

Inc. has addressed all potential concentration mechanisms in their interim criticality reports to 

date. A second issue is that B3NFL Inc. has committed to only one criticality detector in each 

area requiring alarms which is inconsistent with the requirements of 10 CFR 70(a)(1) which 

requires "coverage of all areas shall be provided by two detectors." 

Specific Examples 

1. In BNFL Inc.'s "Interim Criticality Evaluation of HLW/LAW Ultrafiltration," dated 

September 29, 1999, in Section 4, "Methodology," it states that "it is assumed that the 

fissile concentration ol the feed when received at the WTP is equal to the maximum 

concentration specified in the contract." 

2. On BNFL Inc.'s SRD., page 3-9 (dated November 2, 1999), safety criterion 3.3-8, states 

"Coverage of all areas requiring detection may be provided through a single detector" in 

respect to criticality alarms. 

DOE Position 

Although the RU considers the overall criticality safety case developed in the interim 

assessments to be likely to succeed, the RU still maintains a list of open issues which were 

expected to be closed through a July 2000 BNFL Inc."s final criticality report. These open 

issues were provided in November 30, 1999, letter to Mr. M.J. Lawrence (BNFL Inc.) from Dr. 

D. Clark Gibbs, titled "Regulal:ory Unit Response to the Interim Criticality Evaluation for Cesium 

Ion Exchange, Technetium Ion Exchange and LAW Melter Feed Evaporation." The RU has not 

objected to BNFL Inc.'s implementation of only one criticality detector. 

NRC Concerns 

In each of the evaluations, BNlFL Inc. has relied upon concentration limits for special nuclear 

material as provided for in their contract with DOE as an upper boundary for criticality analyses, 

despite the fact that there are waste forms within the Hanford tank waste that may be above 

those limits. The criticality values are from the Carter Handbook that is widely used in the DOE 

Complex. However, these have not undergone rigorous validation. The NRC uses American 

National Standards Institute/American Nuclear Society (ANSI/ANS) 8.1 and other ANS 

standards that have undergone rigorous validation (see NRC Regulatory Guide 3.71). A review 

using the SRP would expect BNFL Inc. to indicate their control methods for ensuring that 

materials of higher concentration will not enter the facility and that accumulation will not occur 

due to the processing. The SIRP would also expect BNFL Inc. to complete their criticality

NUREG-1 747 170
evaluations of all process areas and update those already completed to account for any design 

changes. Thus, the main concerns are: 

1. NRC is concerned that waste forms outside of the contract scope may intentionally be 

introduced in the future because of the planned expanded use of the plant. Further, that 

the current controls for introduction of materials into the facility need to be better 

defined. 

2. NRC is also concerned that not all concentration mechanisms (e.g. preferential settling 

of specific isotopes) have been evaluated in the accident evaluations. 

3. BNFL Inc. has committed to only one criticality alarm in each area requiring alarms 

which is inconsistent with the requirements of 10 CFR 70(a)(1) which requires "coverage 

of all areas shall be provided by two detectors." This position is in direct conflict with 

NRC regulations and would require exemption or facility changes should NRC ever 

assume regulatory authority. 

NRC Evaluation and Assessment 

Although it appears that BNFL Inc.'s criticality approach may be acceptable, additional 

information would be required to resolve the concerns in items 1 and 2, above. BNFL Inc. will 

have to revise their SRD to meet the third concern. 

A13. FUTURE SITE CONDITIONS 

Statement of the Issue 

It does not appear that adequate consideration has been given to the potential for future 

changes that may occur at the Hanford site relative to the impact of changing site activities and 

land use changes. For example, there have been discussions regarding the addition of other 

industrial facilities onsite and the possibility of a reduction in the site area. Such changes could 

result in significant reductions in the distance to the offsite receptor and thus impact the dose 

calculations. 

Specific Example 

On page ES-24, DOE/EIS-0222D (Hanford land use plan) displays a map indicating an 

industrial area 1-2 miles east of the proposed location for TWRS-P. 

DOE Position 

DOE and the Contractor have not examined and considered the potential effects of future site 

boundary changes that are likely to occur during the facility's operational period. 

NRC Concerns 

Future site changes will result in receptors being closer to the facility. This is likely to produce 

changes in the safety analyses that require more important-to-safety (ITS) SSCs in order to

171 NUREG-1 747
demonstrate adequate assurances of safety. For an existing facility, shutdown for backfits may 

become necessary. 

NRC Evaluation and Assessment 

It would be expected that the possible site and surrounding area conditions in the future and 

during the projected design/operating life of the facility would be reflected in information 

reviewed under the SRP and used to estimate the consequences from the various design 

scenarios. This would include determination of ITS SSC's that encompass the effects of land 

use changes over the facility's lifetime. 

A14. PROCESS TECHNOLOGY 

Statement of the Issue 

1. Technology Development and Testing 

Technology development and testing for the facility are not yet complete. Without 

testing results, performance of the selected technologies to be used in the TWRS-P 

facility may not be demonstrated, and design/operating parameters may not be properly 

formulated. 

2. Process Technology and Safety 

Very limited information has been presented on the process technology and its 

relationship with safety. An incomplete understanding of the facility operation may lead 

to the use of an unrealistic model in the accident and consequence calculations. 

Specific Examples 

Unknowns continue to exist in the areas of radioisotope removal efficiency through various ion 

exchange operations; for example, an additional column has been added as a polishing step and cooling coils are shown on the columns, but no additional data is provided. The technology 

to be used for sulfur removal keeps changing. These unknowns will impact downstream 

process shielding and, therefore, may potentially impact the safety analysis of the facility. 

In addition, BNFL Inc. has not: evaluated the potential generation and impact of ion exchange resin fines throughout the process due to erosion and degradation. If resin fines are generated throughout the process, they may be radioactive and could move to other process units and 

eventually impact downstream radiation levels. Adequate information on these technologies 

would be expected to be available in licensing-type submittals reviewed subject to the SRP. 

Few tests appear to be planned to verify safety parameters prior to construction. In addition, 

several of the design points (e.g., melter performance) appear to be significantly outside the 

current window of experimental results (the normal design approach would select operating conditions within the window of results). Additional testing may be necessary to identify, adequately parameterize, and substantiate some safety requirements prior to the submittal of 

licensing documents in order Io demonstrate adequate assurances of safety.

NUREG-1747 172
DOE Position

Preliminary screening tests were completed in July 1999 for the pretreatment unit operations. 

At the July 1999 Topical Meeting, the RU expressed disappointment at the lack of safety 

emphasis in the presentation and, by implication, the test plan. However, a follow-up Topical 

Meeting or status reporting was not pursued. DOE subsequently received a number of short 

test reports in April 2000, and noted that additional, operational-type parameters have been 

tested and evaluated. The NRC has not seen these test reports, but DOE indicates testing to 

assess and verify safety parameters has not been performed and is not planned. 

NRC Concerns 

The Contractor has specified a large amount of research and technology development testing, 

including testing to support environmental permits for the treated melter offgasses. The test 

results discussed at the April 1999 Topical Meeting were screening tests to identify potential 

methods and ion exchange media. As already noted, the NRC has not seen the April 2000 

short test reports. The NRC is concerned that safety issues related to the technologies may not 

be adequately addressed by the experimental programs. In addition, the test quality level and 

experimental variability (including a limited number of tests and repeatability) may introduce 

additional uncertainty into the safety analyses. 

NRC Evaluation and Assessment 

Additional information would be expected to verify operational parameters, identify potential 

hazards and their parameters (e.g., the resin fines), estimate safety parameters and, ultimately, 

provide sufficient information to show adequate assurances of safety. 

A15. CONSEQUENCES OF EXTREMELY UNLIKELY SEISMIC EVENTS 

Statement of the Issue 

Earthquakes in the extremely unlikely range as defined by DOE standards (i.e., 1 E-4/yr to 1 E

6/yr) are not supposed to result in unacceptable accident consequences (dose consequences). 

The NRC SRP provides similar guidance. The proposed modification of 10 Part 70 places 

frequency limits on events; as an example, a high consequence event has to be rendered 

highly unlikely by the facility's design and/or controls (IROFS) and this is furthered defined in 

the SRP as 1 E-5/yr. It would be expected that BNFL Inc. would consider earthquakes that are 

less frequent but more severe than the 2000-year interval design basis earthquake, and show 

that the facility has been designed such that earthquakes in the extremely unlikely range will not 

result in unacceptable radiation exposures. In the review of a licensing submittal, the SRP 

would anticipate that the consequence methodology for these earthquakes would be clearly 

delineated and readily available for detailed review. 

Specific Example(s) 

The staff reviewed the BNFL Inc. proposed "Interpretation of the Requirements for Meeting the 

Radiation Exposure Standards in Table 1 of DOE/RL-96-0006," stated in an enclosure to the 

September 20, 1999, letter from A. J. Dobson, BNFL Inc., to D. Clark Gibbs, Department of 

Energy/Regulatory Unit (DOE/RU) (CCN#: 006158).

173 NUREG-1 747
Note: The referenced document, DOE/RL-96-0006, contains Table 1, 

Dose Standards Above Normal Background, which was left incomplete 

at the time of its release. The standards for Unlikely Events and 

Extremely Unlikely Events, for workers and co-located workers, were 

purposely left open, i.e., "to be derived" by the contractor, BNFL Inc. 

These dose standards were derived by the contractor and implemented 

in Revision 1 of Volume II, Table 2-1, Radiological Exposure Standards 

Above Normal Background, of its Safety Requirements Document 

regulatory submittal (BNFL-5193-SRD-01, Volumes I and II, Rev. 1, 

June 1998). The latter table provides the complete set of standards, 

filling in the "to be derived" portions of Table 1 of DOE/RL-96-0006. 

BNFL Inc. proposed that "For a design basis natural phenomena hazard (such as the design basis earthquake), a frequency of occurrence is established and, from that frequency, the 

appropriate public, facility worker, and co-located worker dose standards are then determined 

from the Table. For natural phenomena events, the table is only applicable at frequencies 

down to and including those established as the design basis for the facility. The table is not 

applicable to the evaluation of natural phenomena events, with frequencies lower than the 

frequency of the design basis event." 

DOE Position 

A reply dated October 5, 1999, from D. Clark Gibbs, DOE/RU, to M. J. Lawrence, BNFL Inc. 

(99-RU-0570), indicated the r)OE/RU's rejection of the BNFL Inc. proposal. The staff 

concluded that the BNFL Inc. proposal also would not appear to be acceptable within the NRC 

regulatory framework. 

Using the example of the desiign basis earthquake cited by BNFL Inc., this interpretation would 

establish its frequency, which is defined as having a 2000-year recurrence interval (i.e., a 

frequency of occurrence of 5 X 104 yr -1), as the lowest probability to be considered with regard 

to the dose standards. According to Table 2-1 this frequency of occurrence would be in the 

category of "Unlikely Events.." The frequencies of other design basis natural phenomena hazards are not addressed by this example, but presumably would be handled in a similar 

fashion under the BNFL Inc. proposed interpretation. 

RU Response to BNFL Inc.: A reply dated October 5, 1999, from D. Clark Gibbs, DOE/RU, to 

M. J. Lawrence, BNFL Inc. (99-RU-0570), indicated the RU's rejection of the BNFL Inc. 

proposal. The reason for rejection was summarized in the letter as follows: 

1. BNFL must ensure that for any credible event (one with frequency greater than 10"6 per 

year), the event will not result in radiation doses that exceed the values given in the 

Table [2-1]. 

2. The BNFL Inc. proposal confuses the requirements for "design basis earthquake" with 

the broader requirements specified in the Glossary of DOE/RL 96-0006 and the SRD for 

"design basis events."

NUREG-1747 174
3. To meet the dose standards in the table, the facility analyses must consider less 

frequent but more severe earthquakes than the defined "design basis earthquake," 

which has a recurrence interval of 2000 years. 

To summarize the RU response, the 2000-year recurrence interval of the design basis 

earthquake (i.e., a frequency of occurrence of 5 x 10"4 yr -1) is not sufficiently low to meet 

the guidelines of the "extremely unlikely events" category of Table 2-1, which range from 

106 to 10"4 yr.". 

NRC Concerns 

NRC staff Comments: The NRC staff reviewed the BNFL Inc. proposal and the DOE/RU 

response, and agree with the position taken by the RU; i.e., that the BNFL Inc. interpretation 

should be rejected. The basis for the NRC staff's agreement on this issue is derived from the 

following considerations: 

1. A Commission Policy and Guidance White Paper on Risk-Informed and 

Performance-Based Regulation, dated March 1999, which refers to the PRA Policy 

Statement (60 FR 42622, August 16, 1995). This Policy Statement formalized the 

Commission's commitment to risk-informed regulation through the expanded use of 

PRA. The PRA Policy Statement states, in part, "The use of PRA technology should be 

increased in all regulatory matters to the extent supported by the state of the art in PRA 

methods and data, and in a manner that complements the NRC's deterministic approach 

and supports the NRC's traditional defense-in-depth philosophy." 

While no specific numerical criteria are mentioned in the White Paper or the PRA Policy 

Statement, an earlier (August 1986) policy statement ["Safety Goals for the Operation of 

Nuclear Power Plants; Policy Statement; Correction and Republication," 51 FR 30028, 

August 21, 1986.] concerning risks to the public from nuclear power plant operation 

stated that guidance to be derived from additional studies to be performed by the staff 

would be based on the following general performance guideline: 

Consistent with the traditional defense-in-depth approach and the accident 

mitigation philosophy requiring reliable performance of containment 

systems, the overall mean frequency of a large release of radioactive 

materials to the environment from a reactor accident should be less than 1 

in 1,000,000 per year of reactor operation. 

This performance guideline is specifically concerned with nuclear power plant operation, 

in which there are multiple levels of defense (i.e., containment systems). The guideline 

indicates that the containment systems are part of the accident mitigation philosophy, 

which means that an accident (initiating event) may occur more frequently than 1 in 

1,000,000 years of operation, as long as the mitigative features function in such a 

manner as to reduce the probability of a release below this value. Although there are 

many differences between a nuclear reactor and the proposed TWRS-P facility, it 

should be noted that the latter may have levels of some radionuclides comparable to a 

reactor in shutdown condition, but in more dispersible forms.

175 NUREG-1747
The TWRS-P facility is also being designed with mitigative systems, which are not 

identical to those in a nuclear power plant, but serve a similar purpose. However, in 

Section 6.0 of RUREG-98-18, the DOE-RU defines the event probability ranges in terms 

of the frequency of the event, not the frequency of consequences ["Regulatory Unit 

Position on Radiological Safety for Hanford Co-located Workers," RL/REG-98-18, Rev. 

0, September 16, 1998], e.g., a large release. Thus the "Estimated Frequency of 

Occurrence" in Table 2-1 (the second column of the table in Appendix A) is the 

frequency of the event (accident), not the frequency of the consequences of the event. 

Therefore BNFL Inc. must show that the dose standards in Table 2-1 will be met for 

events having the stated frequency ranges using mitigative features as necessary in 

order keep exposures to workers, co-located workers, and the public within the 

standards. 

Although the NRC guideline and the DOE-RU requirements are not intended to cover 

the same types of facilities, there is consistency in the regulatory approach; i.e., that 

given an extremely unlikely event (accident) having a credible frequency of occurrence 

between 1 x 106 yr' and 1 x 10-4 yr"1 (per DOE/RL-96-0006), the facility must be 

designed in such a marnner so as to reduce the consequences of the event to an 

acceptable level, usually expressed as a dose to workers and the public. 

2. If the NRC were to asssume regulatory authority over the TWRS-P project, and assuming 

no substantial changes are made in the guidance as outlined in Draft NUREG-1520 

(Standard Review Plan for the Review of a License Application for a Fuel Cycle Facility), 

then, according to Draft NUREG-1520: 

The applicant describes the maximum earthquake magnitude and peak 

ground acceleration at the site and its expected likelihood, in terms of return 

period at which the plant processes can be shut down safely with acceptable 

risk of radiological exposure to workers, public, and the environment. 

Applicant compares the design basis earthquake with the maximum 

earthquake accelerations expected on the site with a return period of 10,000 

years. The purpose of the comparison is to evaluate the likelihood of the 

design basis earthquake to ensure that such an event is properly considered 

in the applicant's lISA. [Standard Review Plan for the Review of a License 

Application for a Fuel Cycle Facility, NUREG-1 520, July 16, 1999, Office of 

Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory 

Commission, Washington, DC 20555-0001] 

This requirement obligates the applicant to consider earthquakes with a recurrence 

interval of no less than 10,000 years (i.e., at least 1 x 104 yr1 ). Therefore, if the 

TWRS-P facility were subject to NRC regulatory authority, the 2000-year recurrence 

interval proposed by BNFL Inc. for the design basis earthquake (i.e., a frequency of 

occurrence of 5 x 10-' yr -1) is not sufficiently low to meet this requirement. 

3. In 10 CFR Part 20, the annual dose limits of 5 rem to the worker and 0.1 rem to 

members of the public are stated, with the application of the principles of as low as 

reasonably achievable (ALARA) in both the design and operation of facilities. These 

dose limits correspond! to relative risks of latent cancer fatalities of 2 x 1 0. yr 1 and 

4 x 10s yr"', respectively. (This is based on Biological Effects of Ionizing Radiation

NUREG-1747 176
(BEIR) V risk estimates of 4 x 10i4 fatal cancers per rem for acute doses less than 10 

rem, and 1 x 10-3 fatal cancers per rem for acute doses greater than 10 rem.) For the 

workers, ALARA generally results in doses of 100-300 mrem (risks of 4 x 10- yr1

- to 1.2 

x 10-4 yr-1). Part 20 limits on decommissioning (10 CFR 20.1402, Radiological criteria 

for unrestricted use) and Part 61 limits on waste disposal (10 CFR 61.41, Protection of 

the general population from releases of radioactivity) display lower annual public limits of 

0.025 rem (a risk of 1 x 10-5 yr'), again subject to the application of ALARA. Typical 

public doses from commercial nuclear operations are well below 0.001 rem yr 1 (a risk 

below 4 x 10.7 yr1 ). 

The risk values stated in the previous paragraph are measures of consequence, not 

event frequency; however, an event frequency may be derived and compared to the 

frequency of the design basis earthquake (or the design basis event) if sufficient 

information is available concerning the source term and the mode(s) of exposure. While 

such a calculation is somewhat outside the scope of this paper, it may be noted that per 

Table 2-1, an extremely unlikely event (an event having a frequency of occurrence 

between 1 x 10'4 yr-1 and 1 x 10" yr1) must result in a consequence of no more than 25 

rem dose to a worker. Therefore, event frequencies corresponding to the doses for 

workers in the previous paragraph would need to be well below the frequency of 

occurrence of 5 x 104 yr -' proposed by BNFL Inc. 

4. The revised 10 CFR Part 70 discusses both high and intermediate consequence events 

in the context of frequency ranges. High consequence events are those that might 

exceed acute doses of 100 rem to workers and/or acute doses of 25 rem to individuals 

outside the controlled area. Section 70.61 (b) states that high-consequence events must 

meet a likelihood standard of highly unlikely. The risk of each credible high

consequence event must be limited, unless the event is highly unlikely, through the 

application of engineered controls, administrative controls, or both, that reduce the 

likelihood of occurrence of the event or its consequence. 

Intermediate consequence events are those that might exceed acute doses of 25 rem to 

workers and/or acute doses of 5 rem to individuals outside the controlled area. 

Proposed Section 70.61 (c) requires that intermediate-consequence events must meet a 

likelihood standard of unlikely. The risk of each credible intermediate-consequence 

event must be limited, unless the event is unlikely, through the application of engineered 

controls, administrative controls, or both, that reduce the likelihood of occurrence of the 

event or its consequence. 

The safety program may be graded such that management measures applied are 

commensurate with the reduction of risk attributable to that item. The frequency terms "highly unlikely" and "unlikely" are not defined in the proposed rule, but guidance is given 

in NUREG-1702, Section 3.7 of the draft TWRS-P Standard Review Plan (SRP), 

[Standard Review Plan for the Review of a License Application for the Tank Waste 

Remediation System Privatization (TWRS-P) Project, Draft Report, March 1999] which 

lists "highly unlikely" as corresponding to 1 x 101 yr"1 and below, and "unlikely" as 

corresponding to the frequency range of 1 x 10-2 yr'• to 1 x 10s yr'. (The Contractor

tailored standards, at 1 x 10"4 yr-' to 1 x 10s6 yr' for "highly unlikely" events, and 1 x 10.2 

yr1 to 1 x 10-4 yr1 for "unlikely" events, are more conservative than those stated in the 

draft TWRS-P SRP.) While these are goals and provide insight, this additional

177 NUREG-1747
conservatism of the flrequency/consequence standard provides some assurance that the 

BNFL Inc. design would meet the NRC guidance as described in the TWRS SRP, as 

long as the event frequencies considered in the analyses are not arbitrarily limited to a 

higher value such as the frequency of the design basis earthquake (i.e., 5 x 10 yr -1) as 

proposed by BNFL Inc. 

NRC Evaluation and Assessment 

To summarize, the NRC staff considers PRA and risk-based analyses to be important parts of 

the safety analyses, but not the only part. The NRC staff recognizes that risk analyses need to 

be supplemented with traditional deterministic analyses, appropriate conservatism, and 

defense-in-depth for effective regulatory decision making, as indicated in the Commission's 

policy statement. BNFL Inc.'s proposed "Interpretation of the Requirements for Meeting the 

Radiation Exposure Standards in Table 1 of DOE/RL-96-0006" would not be in conformance 

with Draft NUREG-1520, which states that the applicant must compare the design basis 

earthquake with the maximum earthquake accelerations expected on the site with a return 

period of 10,000 years (a frequency of 1 x 104 yr-). The contractor's proposal for the design 

basis earthquake would be less conservative than this standard, as well as the implied 

standards based on the 10 CFR Parts 20, 61, and 70 dose limits. As the RU indicated in its 

rejection of the proposal, the Contractor appears to confuse the design basis earthquake with 

the design basis event. 

A16.1 SEISMIC PROBABILISTIC RISK ANALYSIS 

Statement of the Issue 

As stated in Item A.15 the radiological consequences of events in the extremely unlikely range 

as defined by DOE standards are not supposed to result in unacceptable accident 

consequences (dose consequences). In order to assess the radiological consequences of 

seismic events with a return period of greater than 2000 years that was established as the 

design basis earthquake (DBE) for the facility, BNFL Inc. proposed a seismic probabilistic risk 

analysis be performed. The general methodology for that analysis has been provided by BNFL 

Inc. and it has been consistent with current PRA assessment methods. There are, however, 

specific areas within the application of the methodology where there are issues that are yet to 

be resolved. These issues are generally associated with the availability of failure data that are 

relevant for seismic fragility Curves that are used in the methodology. 

Specific Example 

As a result of the hazards analyses performed to date for the facility, it has been determined 

that the major contributor to the dose consequences for the events in the extremely unlikely 

range of occurrence is the loss of the containment/confinement boundaries for the materials 

being processed. This has been translated into the identification of process vessels and 

tankage as well as the process cells as these boundaries. Consequently, the basis for a 

fragility curve for a tank designed, for example, for a peak horizontal ground acceleration of 

0.24 g that shows an 80 percent probability of rupture at 0.80 g has not been established. 

Similarly, the structural failure of cell walls designed to the same acceleration that is stated to 

have a failure probability of 14 percent has also not been substantiated.

NUREG-1747 178
DOE Position

DOE has been in the 'process of attempting to follow the application of the PRA methodology 

through a series of PRA Status Meetings that were still ongoing at the time of the contract 

termination. This issue apparently remains open. 

NRC Concerns 

Since the consequences are evaluated in terms of the radiological doses, structural failure, for 

example, in the cell walls that are considered in this project as mitigation elements, does not 

necessarily define the onset of damage significant enough to permit loss of confinement of a 

portion or all of the radiological inventory. Therefore, to use the PRA methodology will require 

the capability to predict the progression of cracking/distress of the concrete sections under 

increasing seismic loads to define fragility curves that relate structural behavior to leakage 

pathways and subsequent loss of confinement. It is not clear if all of the input data can be 

adequately represented with distribution functions for the purposes of safety regulation. 

NRC Evaluation and Assessment 

Without additional information and qualification of these issues, it is not possible to determine if 

the proposed approach is suitable for quantifying the risk and assuring adequate safety of the 

proposed facility. 

A16.2 RADIOLOGICAL RELEASES FOR SEISMIC EVENTS 

Statement of the Issue 

Certain documents transmitted from BNFL Inc. to the DOE-RU in October and December 1999 

discuss the assessment of radiological releases from the proposed TWRS-P project. These 

documents discuss: 

1. The general methodology of assessing radiological releases.' 

2. Methods for assessing such releases following a seismic event"7 (BNFL Inc. stated that 

this report is based in part on the document in (1) above, and that this report is a part of 

the overall Seismic PRA Methodology contained in (3) below). 

96 BNFL Inc., letter to U.S. Department of Energy, with attachment, "Methods for Assessing Consequences of 

Potential Accidental Radiological Releases from the WTP," (RPT-W375-NS00001, Rev. 0, October 12, 1999), 

October 25, 1999 

97 BNFL Inc., letter to U.S. Department of Energy, with attachment, "Methods for Assessing Consequences of 

Potential Accidental Radiological Releases from the RPP-WTP Facility Following a Seismic Event" (RPT-W375

NS00006, Rev. B, December 7, 1999), December 8, 1999.

179 NUREG-1747
3. Aspects of the techniques of probabilistic risk assessment (PRA) methodology as 

related to seismic events, including risk quantification.98 

The attachment to an NRC memorandum99 to the DOE-RU addresses the NRC comments on 

these documents, as well as the RU's response to the BNFL Inc. methodology of assessing 

radiological releases, and the RU's response to BNFL Inc.'s seismic PRA document. 

Specific Example 

The seismic report (item (2) above) indicates that its methodology is based in part on the 

general dose assessment methodology report (item (1) above); however, the general report 

uses bounding, conservative values, whereas the seismic instead uses best estimate values 

and distribution functions. 

DOE Position 

DOE has accepted the basic, dose methodology approach taken by the Contractor. DOE has 

requested clarification of several points but appears receptive to the general, seismic approach 

being followed by the Contractor. 

NRC Concerns 

The NRC staff indicated in its comments that it would be helpful to include a definition of "best 

estimate values" and the sources of these data, as well as the justification for using these data, 

and the criteria for their selection. The staff also indicated that it would be useful to explain how 

the purpose of the Seismic Probabilistic Risk Analysis Methodology Report (and the related 

Seismic Probabilistic Risk Analysis Dose Assessment Methodology Report) are met using best 

estimate values and distribution functions. These two documents appear to imply that the 

primary reason for performing the analyses of dose consequences arising from seismic events 

is to determine the risk of incurring a given level of dose (i.e., the consequence) to the facility worker, co-located worker, and the public. It did not seem clear from either document whether 

PRA methodology will be used to identify and categorize those SSCs that may require additional strengthening or enhancement for situations in which conservative values of the 

various inputs are hypothesized to occur simultaneously. Accordingly, without knowledge of the 

specific input data and its associated level of conservatism, it was not possible to validate the 

results of the analyses. In addition, the NRC would expect validation of the model, the input 

data, and the results. It should be recognized that additional communications between the RU 

and BNFL Inc. may have resolved some of these concerns, but may have not come to the 

attention of NRC staff. 

NRC Evaluation and Assessment 

98 BNFL Inc., letter to U.S. Department of Energy, with attachment, "Seismic Probabilistic Risk Analysis Methodology" (RPT-W375-NS00005, Rev. B, December 17, 1999), December 20, 1999. 

99 Struckmeyer, R.K., U.S. Nuclear Regulatory Commission, memorandum to M. Leach, U.S. Nuclear Regulatory Commission, "Background Information and Comments on TWRS-P Project: "Dose Assessment Methodology Description, Seismic Probabilistic Risk Analysis Dose Assessment Methodology Report, and Seismic Probabilistic 

Risk Analysis Methodology Report." March 8, 2000

NUREG-1 747 180
Without qualification of these issues, it is not possible to determine if the proposed approach is 

suitable for determining acceptable risk and adequate assurances of safety from the proposed 

facility. 

A16.3 AVAILABILITY OF DOCUMENTS 

Statement of the Issue 

In its submittals, BNFL Inc. refers to a document that does not appear to be publicly available. 

Specific Example 

For example, BNFL Inc.'s SRD, Rev. 2d (November 2, 1999), Vol. II, under Safety Criterion 4.1

2 on page 4-4 references under "Implementing Codes and Standards," ASCE 4-98 (Draft), "Seismic Analysis of Safety-Related Nuclear Structures and Commentary." This document is 

apparently intended to update and replace ASCE 4-86 with the same title. The referenced 

document describes the seismic analysis methods that will be used to perform the seismic 

analysis of safety-related structures of the facility. The document provides a methodology for 

calculating seismic responses in structures and provides input motions for use in the seismic 

qualification of electrical and mechanical systems and components. 

DOE Position 

DOE has not pursued this matter nor the applicability of the draft standard. 

NRC Concerns 

The document in its current state of development does not constitute a standard that can be 

used as a reference since the ASCE Rules for Standards Committees have apparently not 

been fully executed for the standardization of the draft document. If BNFL Inc. wants to follow 

the draft requirements of such a document, those document provisions would have to be 

incorporated in toto into the text of the SRD document, and then acceptability would need to be 

determined by the DOE RU staff. 

NRC Evaluation and Assessment 

This draft standard needs to be reviewed for its acceptability for regulatory purposes.

181 NUREG-1747
A17. FIRE PROTECTION

Statement of the Issue 

There were several fire protection comments generated during review of the SAP and ISAR 

submittals. Most dealt with clarifications or omission of details. Since that time, there have 

been two Level 1 meetings on fire protection: September 1999 and March 2000. From these 

meetings, four major issues, and a number of minor issues remain. 

Specific Examples 

1. Process building construction classification: BNFL Inc. proposed a construction type 

which would not meet the requirements for Type 1 construction per National Fire 

Protection Association (NFPA) 220, nor the fire resistive construction types required by 

a strict interpretation of the Uniform Building Code (UBC). 

2. Steel fireproofing: The steel fireproofing issue is related to the construction classification 

issue. If fire resistance of the outside walls is not required, fireproofing of structural 

steel for these walls would not be required. In addition, BNFL Inc. wished to avoid 

fireproofing of structural steel in areas of low combustible loading. 

3. Protection of final high efficiency particulate air (HEPA) filters from fires: BNFL Inc. has 

stated in the fire protection Level 1 Meeting that no water sprays would be employed in 

filter plenums based on the superior qualifications of the circular filters to be used. 

4. BNFL Inc. has proposed modified requirements for the use of sprinklers in radioactive 

areas of the facility: 

a. C5 Areas (most contaminated with low fire loading): No sprinklers or steel 

fireproofing in C5 areas. 

b. C3 Areas (moderate contamination with varying degrees of combustible loading 

and safety significance): Sprinklers used selectively in these areas with justification 

provided where sprinklers are not used. 

c. C2 Areas (low potential for contamination, with in-situ and transient combustibles, 

such as maintenance supplies): These areas would likely be sprinklered. 

DOE Position 

1. DOE criteria require compliance with the UBC. The RU has commented to BNFL Inc. 

(Gibbs to Bullock, letter dated 5/3/00) that a justification for the selection of nonfire 

resistive construction type would have to include identification and analysis of building 

areas with unusual or concentrated hazards, bounding fuel packages, and critical 

structural members whose failure due to fire could structurally undermine a significant 

portion of the facility. Trhe analysis would also be expected to address any fire 

protection administrative controls, such as transient combustibles and vehicle access to 

the proximity of process buildings (e.g., control of exterior fire exposure hazards).

NUREG-1747 182
2. The RU has commented to BNFL Inc. (Gibbs to Bullock, letter dated 5/3/00) that the RU 

continues to expect that TWRS-P structural steel fireproofing will meet the requirements 

of the UBC. The RU also expects that any type of equivalency analysis will demonstrate 

that there are no fire hazards that present a potential threat to structural steel integrity. 

3. Both NRC and DOE fire protection criteria call for water spray protection in the final filter 

bank of the confinement system. At this time, the RU (Gibbs to Bullock, letter dated 

5/3/00) is requesting clarification of the BNFL Inc. position. 

4. DOE/RU requested that the contractor establish a definitive set of criteria for the use of 

sprinklers in a given area. The contractor wished to rely on the Fire Hazard Analysis 

(FHA) and judgement because of the uniqueness of the different fire areas in terms of 

type and distribution of combustibles, heat release rates, ventilation, etc. 

NRC Concerns 

1. NRC has stated (Tokar to Gibbs, letter dated 6/5/00) that the staff was in agreement 

with the RU criteria for resolution of this concern: a justification of non-fire resistive 

construction type based on analysis of hazards, fuel loading, and the expected fire 

protection administrative controls. 

2. NRC stated ( letter dated 6/5/00 M. Tokar, NRC, to D.C. Gibbs, DOE,) that the NRC 

staff was in agreement with the RU comments requiring an equivalency analysis for 

compliance with the UBC, and added two additional points: 

a. Lack of steel protection based on intended use would severely limit any future 

modifications for use of that area. 

b. Lack of fire suppression capability along with lack of steel protection in the same 

area makes administrative control of combustibles the only defense measure, and 

makes the fire protection in the area a vulnerability that would be identified in a FHA 

or Integrated Safety Analysis (ISA). 

3. Both NRC and DOE fire protection criteria call for water spray protection in the final filter 

bank of the confinement system. 

4. NRC reviewers feel that there is too much uncertainty regarding the concern about 

sprinkler coverage. 

NRC Evaluation and Assessment 

1. Regarding building construction classification, NRC criteria would require Type 1 

construction per NFPA 220. 

2. A letter providing comments on the Explosive Hazards 1 and 2 Topical Meetings 

(Pierson to Gibbs, letter dated 11/4/99) contained a short description of the NRC criteria 

for building construction and expressed concern about the stated BNFL Inc. position 

regarding fireproofing of structural steel.

183 NUREG-1747
3. Since protection of the final HEPA filters involves the basic design of the ventilation 

system, NRC staff preferred to make this a plant systems or radiation protection issue 

for the time being. 

4. NRC has requested a set of definitive criteria for determining the use of sprinklers in a 

given area. 

Finally, the Hanford site recently experienced a large brush fire that burnt approximately 50 

percent of the site's total land area. Circa 200,000 acres were burnt, along with some 

structures in west Richland. Additional fire breaks and berms had to be constructed around 

some of the facilities for protection. Therefore, the NRC would anticipate that the Contractor 

would also factor this experience in the facility's design. 

A18. EXPLOSIVE HA2ARDS 

Statement of the Issue 

The last topical meetings in which explosive hazards for the TWRS-P project were discussed in 

detail were in August and September 1999. The August Topical Meeting (Explosive Hazards I) addressed hazards from melter steam explosions, explosions from nitrate organic reactions in 

the melter'offgas system, over pressurization of an ion-column, ammonium nitrate explosion in 

the off-gas system, and a sugiar dust explosion in the feed preparation vessel. None of these 

events were considered credible by BNFL Inc., although specific probabilities were not 

presented to support such conclusions and there appeared to be an over-emphasis upon routine and near-normal conditions. Consequently, all of these events were left as potential 

issues by NRC and the RU pending further information and analysis. A resubmittal of the 

August 1999 Topical Report in March 2000 and information from design review meetings alleviated some of the NRC concerns about steam explosions, nitrate-organic reactions, and 

sugar dust explosions. Current NRC staff concerns about steam explosions are primarily the 

potential for a refractory failure allowing molten glass to contact water in the cooling jacket. Potential explosions caused by the contact of water and the cold cap were partially addressed 

by the BNFL Inc. responses to comments; nonideal behavior and misfeeds appear to require 

further analysis. Also, the BNFL Inc. process does not appear to be as vulnerable to 

radiological releases from organic-nitrate and sugar explosions as originally determined from 

the first submittal, although some questions still remain, primarily dealing with misfeeds and 

frequencies. Potential explosions from ammonium nitrate formation and over pressurization of 

the ion exchange column are still considered as open issues requiring considerable progress. 

The September Topical Meeting (Explosive Hazards II) was concerned with the potential for 

explosion of hydrogen gas in approximately 40 process tanks. At this time an active ventilation 

system for hydrogen build-up was proposed to take the place of the passive release system that was proposed earlier, probably for a single tank. This active system consisted of an air 

extract system with two 100 percent fans and one 100 percent backup fan. Process air is used 

for dilution during normal operation. An air vent system is provided for loss of off site power (with loss of process air) conditions. Although the staff considered this design a significant 

improvement over the passive system, there were concerns from the presentation and report. 

These concerns included: 

1. Hazard evaluations based on normal operating conditions rather than off normal.

NUREG-1 747 184
2. Rupture of vessel vent jumper was not considered in hazard evaluation. 

3. Loss of ability to maintain pressure control in vessel vent system was not considered. 

4. Large surges in offgas were not considered. 

5. The worse case detonation scenario was not properly evaluated. 

Specific Example(s) 

(Already cited.) In addition, at the June 2000 Topical Meeting on closing out issues, there was 

an ISM presentation on hydrogen explosion concerns. However, the presented ISM was very subjective with few quantitative analyses that supported its intended conclusion of no hydrogen monitoring requirements. Furthermore, previously unmentioned components of the system 

were introduced, such as the dampers and air flow monitors, without reliability analyses. 

DOE Position: 

DOE finds the additional information interesting but insufficient to adequately explain (and 

address) the issues. After the June 2000 Topical Meeting, DOE is requesting another 

demonstration of the ISM process on the hydrogen explosion concerns with more quantitative 

information and reasoning. 

NRC Concerns 

A review of the resubmittal in February 2000 also led to a re-evaluation of the open issues. 

The NRC staff concluded that rupture of the vessel vent jumper and loss of ability to maintain 

pressure control can probably be nearly eliminated by an expanded set of performance 

requirements on the system. The other concems remain however. 

The June 2000 Topical Meeting introduced dampers, flow monitors, and other equipment on the 

ventilation system for the first time. No reliability information was provided. However, a cursory review by NRC staff indicates a frequency of failure of circa 1 E-4/yr (assuming perfect operation of the fan system), which would not render hydrogen explosion events incredible. 

Details of the NRC staff's concerns with the original submittals and presentations for the 

Explosive Hazard Topical Meetings may be found in R. Pierson's, NRC, letter of November 4, 

1999, to D.C. Gibbs, DOE, entitled "NRC and CNWRA Comments on the Explosive Hazards 

Topical Meetings I and II, August 24, 1999 and September 28, 1999." Details of the review of 

the resubmittals may be found in M. Tokar's, NRC, letter of April 27, 2000 to D.C. Gibbs, DOE, 

entitled "NRC Response to March 23, 2000, Letter Concerning the Resubmittal of the August and September 1999 Topical Meeting Reports on Explosive Hazards." 

NRC Assessment and Evaluation 

It would be anticipated that the Contractor will complete the design in sufficient detail, with 

quantitative analyses, and with identification of IROFS, before the explosion issues can be 

adequately addressed and closed.

185 NUREG-1 747
A19. COMPLIANCE WITH INIQA-1

Statement of the Issue 

BNFL Inc., in its QA Program Implementation Plan (QAPIP) has committed to implementing a 

QA program in accordance with applicable requirements of American Society of Mechanical 

Engineers (ASME) NQA-1 -19139, as well as the DOE QA requirements of 10 CFR 830.120. 

BNFL Inc. has not adequately identified its commitments or interpretations related to the 

applicability of specific NQA-1 requirements, either as full and explicit commitments to NQA-1, 

nor have exceptions to and applicability of the provisions of the document been identified. 

Implementing procedures for NQA-1 requirements appear not to have been fully and completely developed, and there is no explicit commitment to the records storage and software 

requirements of NQA-1. 

Specific Example 

Implementing procedures for all NQA-1 requirements appear not to have been fully and 

completely developed, and there is no explicit commitment to the records storage and software 

requirements of NQA-1. BNFL. Inc. has not identified exceptions taken to NQA-1 -1989. 

DOE Position 

DOE considers the BNFL Inc. commitment in its QAPIP to NQA-1-1989, to be a full and 

adequate commitment at this point in the project. Program implementation and NQA-1 

requirement applicability will be addressed as necessary during inspections. 

NRC Concerns 

All applicable NQA-1 requirements needed for design, including records and software control 

programs, should be fully in-place during the current design phase, while others should be fully in-place before additional safety analyses, advance procurement or initial construction are 

begun. Correction and/or bac'kit of design analyses and documentation performed under an 

inadequate design control program could become necessary. 

NRC Evaluation and Assessment 

In the TWRS SRP, NUREG-1520, NRC endorsed NQA-1 -1989 as an acceptable QA program standard for the TWRS project. Exceptions or "tailoring" of the NQA-1 requirements should be 

addressed, supported as technically adequate or rejected as early in the project as practical to 

avoid costly and laborious retrofits. 

A.20 QUALITY ASSURANCE PROGRAM IMPLEMENTATION 

Statement of the Issue 

BNFL Inc. has committed to apply and implement the QA program across the entire project organization including the subcontractors for design, procurement, and construction. However, 

each of the various major subcontractors operate under their own individual QA program. There is the potential for conflicting and missing requirements for a project when a multitude of

NUREG-1747 186
QA programs are in-place. At this stage of the project there is concern regarding the adequacy of design control across the entire project for the detailed design and the beginning of 

procurement in areas such as the control of design inputs from the various subcontractors. The 

level of BNFL Inc. engineering verification and QA oversight of the subcontractors/team 

members activities to assure that adequate application and implementation of all QA 

requirements for ongoing design has not been evident to date. 

Specific Example 

BNFL Inc. conducts its activities in accordance with its QAPIP. GTS Duratek, who is designing the high and low level vitrifiers and the basic vitrification process, and offsite Bechtel 

engineering organizations operate to their own QA plans or manuals. 

DOE Position 

DOE considers that BNFL Inc. has committed to adequate QA program implementation and 

inadequacies will be addressed during inspections. 

NRC Concerns 

The adequacy and appropriateness of the QA program activities requires strong and effective 

QA oversight even when all participating organizations are operating to the same QA plan. The 

use of multiple QA plans by a variety of organizations and corporate entities in several locations 

requires application of QA oversight to assure adequacy and appropriate implementation by all 

participants. This level of oversight has not been evident in the project activities to date. 

NRC Evaluation and Assessment 

The appropriate and adequate implementation by all major team members and subcontractors 

of QA requirements should be verified prior to or early in the project design activities in order to 

preclude extensive avoidable retrofits for design, procurement or other project activities. 

A.21 SAFETY CLASSIFICATION OF STRUCTURES, SYSTEMS, AND COMPONENTS, 

AND GRADED QA 

Statement of the Issue 

A complete safety categorization of the structures, systems and components (SSCs) has not 

been fully developed by BNFL Inc. or provided to DOE. The Design Safety Features (DSF) 

deliverable that BNFL Inc. submitted in February of 1999 was intended to be an initial 

identification of the DSF which would include identification of those elements of the facility important- to-safety (ITS) including SSCs that give assurance that the SSCs will perform the 

required safety functions. BNFL Inc., in the DSF submittal, provided two categories of ITS 

SSCs: those elements that were to be ITS based on previous experience in the design of 

similar plants and the other category being those SSCs that were identified in the application of 

the ISM process. At the time of the DSF submittal, BNFL Inc. had identified 568 SSCs that 

were expected to be ITS and 1003 DSFs had been identified. It was also clearly stated that the 

total number of ITS SSCs identified to date did not represent the number of elements that 

would be ITS in the final facility. Associated with the safety categorization of the SSCs is the

187 NUREG-1747
related classification within the QA program that in this case is to use a graded approach to QA. 

Without a list of all SSCs for each of the safety categories, a logical and systematic graded QA 

program cannot be developed. Such a list has not been presented up to the time of the 

submittal of the Limited Consl:ruction Authorization Request (LCAR), dated June 26, 2000. 

Specific Example 

BNFL Inc. intends to use confinement as a DSF. For example, an element within the facility 

that can function as a secondary confinement barrier is one of the individual process cells within 

the pretreatment building. Credit will be taken for such confinement barriers as passive barriers 

achieved with the stainless steel liner of the cells, the seals on the penetrations and by the fact 

that there will be a limit on the thru-wall cracking of the reinforced concrete cell walls. These 

are considered by BNFL Inc., as DSF to protect against loss of material and to provide adequate 

shielding. The cell wall was designated as ITS. Other DSFs associated with the cells that were 

identified by BNFL Inc. include the use of design codes, use of a suitably qualified and 

experienced person ("SQEP") in design and construction, seismic design, material selection, 

nondestructive testing/inspection/QA during design and construction, shielding assessment, 

design to withstand maximurn cell temperature and analytical assessment of structural stability. 

In spite of the words in the D'SF submittal, and the fact that no complete list of ITS SSCs has 

been provided, the LCAR, dated June 26, 2000, contains the following statement. "Stainless 

steel liners are not designated as Important to Safety." This is apparently believed to be 

justified by the statement that, "Process cells are provided with liners (typically stainless steel) 

to prevent leakage of spilled process liquor into the soil column over time. The liners also 

facilitate decontamination of the process cells to minimize the contamination spread during 

operation and to facilitate deactivation and decommissioning. Containment of spilled liquor as a 

means to reduce operating risk to within exposure standards for workers or members of the 

public relies on the concrete sctructure itself and does not require liners." 

The DSF submittal contains the following statements. "Specification for materials used for 

stainless steel cladding have been developed based on extensive experience gained in 

cladding process cells on many projects. Specified tolerances for materials along with welding 

and weld test procedures ensure quality of secondary confinement provided by the cladding. 

All welded seams are examined to ensure they are free of unacceptable defects." The 

submittal also provided information that would indicate some safety level and quality levels are 

envisioned, although not well defined at this time. Listed below are some of the tests that were 

indicated would be performed. 

1. Visual inspection 

2. Dye penetrant examination. 

3. Ferrite meter check. 

4. Ultrasonic examination. 

5. Radiographic examination. 

6. Copper tests. 

7. Pneumatic pressure test. 

8. Ponding test. 

DOE Position

NUREG-1747 188
DOE has identified the need for the completion of the safety classifications and has expected 

this information to develop from the completion of the Integrated Safety Management system 

approach which had not been completed at the time of contract termination. While the LCAR 

has been rejected by DOE, its rejection was not based on this issue. This issue is expected to 

be an open issue when the project resumes. 

NRC Concerns 

Until there is an adequate list, meaning most major SSCs, of the SSCs in each of the 

designated safety categories and graded QA classes, and whatever other logic systems are 

associated with the facility provided, there is a concern that the initiation of construction under 

the LCAR may allow SSCs that should be ITS and in a certain quality class to be addressed in 

a substandard level when compared to the needs of the facility. 

NRC Evaluation and Assessment 

Because there is a strong emphasis on the design margins being only sufficient enough so that 

the target consequences are met based on the ISA, there is concern that the initiation of 

construction activities prior to a completed design being available for hazards analyses will 

result in judgements being made later that are not necessarily well founded when compared to 

what is built and in-place at that time. For any SSCs involved in any LCAR activity, the safety 

class and quality class should be soundly based or there should be conservative assumptions 

made and documented for the future verification. 

A.22 CHEMICALS AND THEIR SAFETY 

Statement of the Issue 

Several chemical safety issues were initially identified in the ISAR. The DSF submittal further 

analyzed one of these scenarios. However, the analysis may have been incomplete. From 

subsequent discussions, it is not clear if chemical safety issues (in either radioactive or 

nonradioactive areas of the facility) are being evaluated and addressed as necessary, and 

would be available for regulatory evaluation in the near term. In addition, the design of the cold 

chemical storage areas is very general and the detail may not be completed and available for 

specific safety analyses. Thus, the safety evaluation of those areas and their effects could not 

be performed. The changes in the storage and handling system concepts for the wet chemicals 

need to be resolved and integrated into subsequent safety submittals. 

Specific Example(s) 

1. Several chemical safety issues were initially identified in the ISAR, including storage of 

concentrated nitric acid, concentrated sodium hydroxide handling, anhydrous ammonia 

handling, and dry glass formers storage and handling. 

2. The Design Safety Features (DSF) subsequently analyzed the concentrated nitric acid 

spill. However, the analysis was inconsistent and incomplete. 

3.. The Contractor intends to subcontract much of the nonradioactive chemical handling 

area. This has contributed to the current lack of attention and detail in this area, and the

189 NUREG-1 747
lack of a complete hazards and operability (HAZOP) study of the chemical handling 

areas. 

4. The Contractor performed an initial safety screening of chemical spills (in the ISAR and 

DSF) using the Automated Resource for Chemical Hazard Incident Evaluation 

(ARCHIE) model, available from the U.S. Environmental Protection Agency (EPA) public 

domain website. ARCHIE was apparently selected because of its wide use for 

screening potential events, simple Gaussian plume algorithm, and little modeling experience is required for its use. For this application, however, there are a number of 

uncertainties regarding its use without considering uncertainties and the appropriate 

vapor pressures. As an example, ARCHIE indicated there would be no ITS SSCs from 

chemical incidents. However, the ISM Cycle 2 utilized a different model that indicated 

there might be ITS SSCs associated with the prevention and mitigation of concentrated 

nitric acid spills. 

5. The following additional examples accrue from the April and June 2000 Topical 

Meetings: 

a. Limited detailed information, particularly for quantitative and reliability analyses. 

b. Incomplete and uneven application of the HAZOP process-for example, offnormal 

conditions and the significance of pump and flow controls for static mixers are 

overlooked. 

c. Inconsistency with NRC guidance (the issued NUREG-1 702) and draft regulations 

(the revised 10 CF'R Part 70). 

d. Potential high level of risk, some two orders of magnitude or more greater than 

radiological risks. 

DOE Position 

From the April 2000 Topical Meeting minutes the RU questioned: 

1 . The proposal to have different connection types and sizes to prevent unloading, 

misloading, and inadvertent mixing of incompatible chemicals, because most trucks 

carry multiple connectors. 

2. The same piping and equipment in facility bulges for more than one chemical, thus 

risking an incident of mixing incompatible chemicals. 

3. The general risks from frequent tank truck deliveries to the facility (eight or more a day). 

At the June 2000 Topical Meeting, DOE/RU has stated that they are inclined to accept 

the Contractor's chemical safety approach that considers ITS only when a potential 

scenario involves multiple hospitalizations, a worker fatality, or exceeding Emergency 

Response Planning Guideline-3 (ERPG) levels for the CLW. 

NRC Concerns

NUREG-1747 190
The safety evaluations of the chemical areas cannot be completed without sufficient design and 

hazards information. In addition, the NRC is concerned that the proposed approach constitutes 

a significantly greater risk from chemicals than from radiological incidents and is inconsistent 

with commercial nuclear industry practice at fuel cycle facilities. The approach is also 

inconsistent with the revised 10 CFR Part 70 and related guidance that has been developed in 

conjunction with the industry. 

NRC Evaluation and Assessment 

The chemical safety area needs to be revisited and the design finalized and quantitatively 

analyzed. 

A.23 IODINE REMOVAL 

Statement of the Issue 

The status of iodine removal from the process system needs to be resolved and the supporting 

information would be expected in safety submittals. A condensed history of the iodine-129 

issue follows: 

1. From the ISAR review in February 1998, DOE/RU question 102 asked about the 

projected iodine-129 release rates as well as the effectiveness of absorption unit 

operations in the HLW primary offgas system. There was little response by the 

Contractor. 

2. Subsequently at the May 1999 Topical Meeting, BNFL Inc. stated that iodine-129 

removal was not required. The RU requested additional information that might support 

this assertion. On September 7, 1999, BNFL Inc. provided estimates of iodine releases 

and the associated doses in letter CCN 005716 and its Attachment 1. The RU 

questioned the assumptions in the letter because they were lower than the contract 

specifications. On October 5, 1999, BNFL Inc. clarified the situation by claiming that 

iodine controls "other than caustic scrubbers" (e.g., silver zeolite beds) were 

unnecessary. 

3. In March of 2000, BNFL Inc. provided additional information supporting the sole use of 

caustic scrubbers for iodine control, including the basis for decontamination factors 

(DF's), in CALC-W375HV-PR00092 (also referenced in the BNFL Inc. letter CCN 

012970, Dobson to Gibbs, dated 5/5/00). BNFL Inc. also submitted a draft of the Best 

Available Radiological Control Technology (BARCT).

191 NUREG-1747
Specific Example

The Contractor estimated a dose of circa 0.3 mrem/yr to the public from anticipated iodine-1 29 

releases, based upon a DF of 100 across the caustic scrubbing system. 

DOE Position 

The DOE/RU performed an independent dose calculation using conservative assumptions and 

obtained similar results as the! Contractor (see RU internal memorandum REG:JLP/00-RU

0318, dated April 11,2000). Consequently, the RU has concluded that the Contractor is 

proposing effective control strategies for iodine-1 29 and that the doses will be a small fraction 

of the annual limits. This conclusion still requires the use of an effective scrubbing technology, 

such as the caustic scrubber approach presented by BNFL Inc. at the March 2000 Topical 

Meeting. However, the RU has not limited the Contractor to caustic scrubbing as the 

decontamination technology.. 

NRC Concerns 

The NRC is concerned that effective control measures are applied to iodine due to its biological 

significance and known thyroid effects. These concerns are magnified by the arid environment 

around the Hanford site. At face value, the proposed caustic scrubber approach would seem to 

give acceptable dose results. However, there is a lack of detailed information on the design and approach - all caustic scrubbers do not perform equally well for iodine removal. 

Furthermore, an assessment of the caustic scrubbing systems reliability, performance 

(particularly with time), off normal effects, and safety significance would be anticipated. For 

example, failure of the system (a DF of 1) could result in doses of circa 30 mrem/yr to the public 

if it is not repaired or maintained in a timely fashion. This would exceed air permit limits 

(NESHAP - 10 mrem/yr) and would be at least an ALARA design issue from the NRC 

perspective. Additionally, the anticipated doses are strong functions of distance from the stack, 

and since the NRC considers the CLW as a member of the public, a functioning system could 

correspond to doses of circa 0.3 rem around the proposed facility's fenceline. The Hanford 

land use EIS anticipates site privatization initiatives and an industrial park about 1 mile from the 

200-East area. Thus, even if the NRC were to accept the CLW approach, anticipated public 

doses would still be higher than the 0.3 mrem/yr estimate. The NRC would also expect an 

assessment of offnormal effects upon the scrubbing system's performance, such as differing 

feed compositions (iodine amounts), scrubbing medium composition, packing deterioration, 

liquid/gas ratios, monitoring systems, and associated management measures. Like the RU, 

the NRC would not limit the Contractor to a specific scrubbing technology although the 

characteristics and experience of the technology would be factored into the NRC evaluation of 

its ability to meet the performance requirements. 

NRC Evaluation and Assessment 

The NRC would expect the previously mentioned concerns to be fully addressed in safety 

documentation.

NUREG-1747 192
INCREASED TANKAGE AND INVENTORIES

Statement of the Issue 

As the design has evolved during the Phase 1.B-1 period, the potential quantity of waste that 

could be stored within the tankage volume now being proposed for the facility has increased 

considerably and may contain the equivalent waste volume of up to three of the double-shell 

tanks (DST). It is not clear if additional increases in tank volumes will occur because a sizing 

rationale for the tanks has not been presented. For example, there are now six circa 380 Kgal 

tanks for LAW receipt and six circa 90 Kgal tanks for HLW receipt and storage. In addition, 

separated and concentrated radionuclides may also be stored (e.g., cesium, technetium, 

strontium/transuranic waste, and UF [ultrafiltration] concentrates). To date, the Contractor has 

not used a consistent and reasonably conservative approach for estimating tank and 

radionuclide inventories. For example, the mass balances in the FFP submittal use 

radionuclide values potentially at least an order of magnitude lower than the contract allows. 

Per the SRP, the NRC would expect that the design basis for the facility is clearly described 

and that the volumes and the associated activity have been adequately analyzed in the dose 

assessments for the various scenarios. 

Specific Example(s) 

Per the FFP submittal, there are at least 3 million gallons of tank storage space at the facility, 

as identified in the previous paragraph. This is equivalent to three DSTs at Hanford. 

DOE Position 

DOE does not have a clear position on tank storage. The large amount of tankage appears to 

be driven by DOE concerns for batching flexibility, HLW storage space, and higher plant 

capacities. 

NRC Concerns 

The tankage represents a significant radiochemical inventory in a mobilized or easily-mobilized 

form. 

NRC Evaluation and Assessment 

NRC evaluations indicate the tanks represent a significant source term for leakage, spill, 

misrouting, and external event accident scenarios. In addition, they will contribute to operations 

and maintenance activities at the facility (e.g., for pumps, instrumentation, agitators, sampling, 

breather filters, etc.). The NRC would anticipate that there would be safety controls associated 

with this large tankage. The presence and identity of these controls is not apparent in the FFP 

submittal and related documentation.

NUREG-1747

A.24

193
A-25 INSPECTION FEATURES

Statement of the Issue 

A complex radiochemical facility such as TWRS-P is likely to need numerous inspections over 

its design and operating periods. These would include inspection of SSCs, such as acceptance during construction, and corrosion of pipes and vessels, weld/connection integrity, instruments 

and calibration, cleaning, removal of deposits, and seal integrity during operations. In addition, 

additional inspections would focus on the operator and procedural issues, radioactive and 

material flow through the facility, and general conditions (of plant, people, and organization). The deliverables for this project did not contain or imply that inspections were included during the design or operating periods. 

Specific Examples 

Section 3.13, "Reliability, Availability, Maintainability, and Inspectability (RAMI)" of the 

Contractors ISMP, BNFL-51 93-ISP, rev. 4, dated December 2, 1998, requires that testability and inspectability of Safety Design Class systems and components be facilitated during design by such features as redundancy, that allow for a system or component to be removed from 

service for maintenance or testing without loss of safety protection and provisions. 

Contrary to the above, the Contractor had neither proceduralized nor implemented the 

requirement to consider inspectability and testability in the evolving facility design. No special features are identified in the designs submitted with the FFP material that would assist 

inspection of SSCs for constnjction acceptance or continued operation. Special ports for 

probes (e.g., eddy current, ultrasonic, radiation) and scopes are not apparent. There are many 

components that are in continuous use without built-in spares. Many of the pumps (e.g., the 

evaporator recirculation pumps) have neither a spare nor isolation valves to facilitate testing and inspection. 

DOE Position 

Members of the Design Process Inspection Team had identified this omission in the Design Process Inspection of January 2000. Because the ISMP represented part of the project Authorization Basis, this was; considered an Inspection Finding. This information was conveyed 

to the Contractor via letter 00-.RU-0210, Gibbs to Bullock, dated February 8, 2000, which 

transmitted the Design Process Inspection Report, IR-00-001. Additional follow-up has not 

occurred at this time. 

NRC Concerns 

NRC staff participated in the Design Process Inspection Team, and inspectability 

considerations were included in NRC Inspection Report 70-3091/0001. Without adequate 

means to inspect and test SSCs built into the design, it will be difficult for the Contractor to 

demonstrate adequate safety, particularly as the plant ages. Since it is required by regulation, then, without adequate design features to simplify inspection, the plant is likely to suffer 

capacity losses to accommodate the inspection needs.

NUREG-1747 194
NRC Evaluation and Assessment

NRC would anticipate that approaches and specific design features for inspection would be 

described as part of the management measures in a potential license submittal. 

A.26 CHEMICAL VERSUS RADIOLOGICAL RISK 

Statement of the Issue 

The NRC approaches the potential hazards and risks form radioactive materials, chemicals 

generated by radioactive materials, and chemicals that may impact the safe handling of 

radioactive materials in a uniform manner via an ISA. Per the revised 10 CFR Part 70 and the 

SRP, potential hazards of high consequence identified by the ISA should be rendered highly 

unlikely by design and controls, while potential hazards of medium consequences should be 

rendered unlikely by design and controls. The SRP further identifies 1 E-5/yr or lower as highly 

unlikely and the range of 1 E-2/yr to 1 E-5/yr as unlikely. The effects from chemicals are 

delineated in terms of emergency planning levels (essentially one hour or less exposures). The 

Contractor has identified process industry guidance as the basis ("standard") for potential 

chemical hazards at the TWRS-P facility and intends to use standard, off-the-shelf equipment. 

However, this appears to result in a significantly higher, accepted level of risk from chemical 

events as compared to radiological events. 

Specific Examples 

As an example with the FFP submittal design, an offgas release to the gallery around the LAW 

melters would likely have fatal consequences for the personnel present with a potential 

frequency of circa 1 E-3/yr. This is two orders of magnitude above the SRP value and 

potentially a thousand times greater than the SRD value for high dose, radiological accidents. 

The personnel would be exposed to radiological, chemical, and physical consequences (etting, 

temperature, and pressure); however, the radiological consequences would not likely constitute 

the greatest risk or cause the potential fatalities. No IROFS were identified for this event. 

DOE Position 

From the April 2000 Topical Meeting minutes, the RU questioned: 

1. The proposal to have different connection types and sizes to prevent unloading, 

misloading, and inadvertent mixing of incompatible chemicals, because most trucks 

carry multiple connectors. 

2. The same piping and equipment in facility bulges for more than one chemical, thus 

risking an incident of mixing incompatible chemicals. 

3. The general risks from frequent tank truck deliveries to the facility (eight or more a day). 

At the June 2000 Topical Meeting, DOE/RU stated that they are inclined to accept the 

Contractor's chemical safety approach that considers ITS only when a potential scenario 

involves multiple hospitalizations, a worker fatality, or exceeding ERPG-3 levels for the CLW, 

because the approach was previously approved in the SRD by the RU.

195 NUREG-1747
NRC Concerns

The proposed approach is inconsistent with NRC guidance, the revised 10 CFR Part 70 

regulations, and commercial fuel cycle approaches that balance chemical and radiological risks 

at fuel cycle facilities. It is worth mentioning that the American Institute of Chemical Engineers has printed material (see, for example, Guidelines for Engineering Design for Process Safety) that discuss in detail the desire for "inherently safer" plants with "multiple safety layers" (redundancy). In addition, best handling practices are available from manufacturers of specific 

chemicals (e.g., nitric acid) and chemical categories - these generally emphasize precautions and redundancy. The Contractor does not appear to be following this chemical process 

industry guidance. 

NRC Evaluation and Assessment 

The level of risk associated with chemical events is relatively high and incongruous with NRC 

regulations and guidance, and nuclear and process industry trends towards lower risk. It would 

seem that the potential risk level from chemical events should be lower and comparable to the 

radiological risk level, as noted by the revised 10 CFR Part 70 and the SRP. 

It should be noted that the use of Emergency Response Planning Guidelines (ERPGs) for 

analyzing chemical exposure may not be sufficiently conservative for some scenarios. ERPGs 

are intended to be used for evaluating emergency escape only, where the unit of the "effect" 

from exposure for 1 hour is not a "time weighted average" (TWA). Also, the ERPG values are 

deterministic and represent significantly different acute effects - ERPG-1 corresponds to 

irritation and annoyance, ERPG-2 represents an injury level that does not impede emergency 

egress, and ERPG-3 denotes a level corresponding to irreversible injury. Thus, while generally accepted as a screening tool, the use of ERPG values may be inappropriate for longer duration 

scenarios such as control room habitability. Instead, lower values (e.g., TWA) may be better 

choices for those situations. 

A.27 RELIANCE ON THE OPERATORS 

Statement of the Issue 

The proposed TWRS-P handles relatively large quantities of materials and activities for a 

radiochemical plant and contains numerous cells. While some of the cells are remotely 

operated, personnel access is planned for many of these cells for routine maintenance and 

recovery. In addition, the Contractor has indicated a preference for administrative controls and 

operator actions (including evacuation) over engineering and design controls. 

Specific Example(s) 

1. The Contractor has elected to use a locally-shielded LAW melter in lieu of the traditional, 

remotely operated mell:er. The replacement of melter consumables, such as bubblers 

and thermowells, is intended to be done manually using a Consumable Changeout Box 

(CCB). This might occur 10% of the time and, in addition to the CCB procedures, 

requires an operator procedure to place the melter into idle mode.

NUREG-1747 196
2. The same LAW melter is located in a gallery with two other melters. Operators will be 

present in the gallery 30 percent or more of the operating time. The Contractor intends 

to use evacuation procedures in the event of an accidental release of radioactive 

materials into the gallery. 

3. In the current design, the pump bulges and canister decontamination cells rely upon 

operator entry for maintenance. Per NRC regulations, these would likely be categorized 

as "very high radiation areas" and require procedures for decontamination, canister 

movement, and/or door interlocks to prevent overexposures. 

4. Many of the tables in the ISM Cycle 2 list "operator actions," procedures, and/or 

administrative controls. "Restricted access" is frequently listed as a primary control 

strategy to reduce operator exposure. 

DOE Position 

During the March, 2000 Topical Meeting on Melters and Offgas Treatment, DOE/RU viewed a 

computer-generated model of the operation of the Consumable Changeout Box (CCB). DOE 

appears to be willing to accept reliance upon the operators as part of the primary hazard control 

strategy. 

NRC Concerns 

This approach of relying on administrative controls and operator actions runs opposite to the 

preference stated in the SRP NUREG-1 702 and the revised 10 CFR Part 70 Statement of 

Considerations. 

NRC Evaluation and Assessment 

The NRC would look more favorably upon approaches that do not rely upon operator actions as 

either the primary or the secondary control strategy. 

A.28 TWRS-P SITE SPECIFIC GEOPHYSICAL AND GEOTECHNICAL 

INVESTIGATION REPORT 

Statement of the Issue 

The analysis and design of the structural foundation system of the existing site materials and 

those materials that may be transported to the site as well as the concrete and steel structures 

to be constructed on or in the site materials should reflect the site specific conditions. Pre

topical meetings on the subject of seismic analysis and design that were held in 1998 and 1999 

targeted the report of the site specific geophysical and geotechnical investigations and 

recommendations to be available in mid-1 999 to correspond to the start of the seismic analysis 

in March 1999 and the completion by the end of December 1999. This schedule would allow, 

based on the project schedules, to have the structural analysis completed by mid-April 2000. 

This schedule would provide for the completion of the development of structural and civil 

information for input to the LCAR. Repeated delays in the initiation of the work as well as in the 

completion of the work occurred in 1999 with meetings to discuss the data, findings and 

recommendations being postponed in August, September and October of 1999, and then no

197 NUREG-1747
meetings being identified. Until the LCAR, dated June 26, 2000, was submitted no information 

has been available relative to the site subsurface investigations. The LCAR identifies the 

document as, "Geotechnical Investigation Report," Shannon and Wilson, Inc., H-161-51, dated 

May 2000. 

Specific Example 

The basis for the analysis and design of the pretreatment building may be based on assumed 

information and the actual report of the geotechnical conditions has not been reviewed and 

evaluated by the regulatory authority. 

DOE Position 

DOE recognizes that this report and the information it contains is long overdue, however, the 

contract for the work was apparently under DOE control, not BNFL Inc. 

NRC Concerns 

While the report is late and untimely, the data and recommendations in the report must be 

thoroughly reviewed for adequacy and the final regulatory evaluation and recommendations will 

have to be reflected in the analysis and design completed to date. Assumptions made during 

the period of time when the design parameters from the actual report were not available must 

be verified as conservative. The adequacy of the complete analysis and design work should 

not be influenced by the fact that rework will have to be performed. 

NRC Evaluation and Assessment 

NRC did not receive the report in the LCAR material provided by the RU just as the project was 

being terminated, so no evaluation or assessment of this issue was made.

NUREG-1747 198
APPENDIX B 

January 29, 1997 

MEMORANDUM OF UNDERSTANDING BETWEEN THE NUCLEAR 

REGULATORY COMMISSION AND THE DEPARTMENT OF ENERGY 

COOPERATION AND SUPPORT FOR DEMONSTRATION PHASE (PHASE 1) 

OF DOE HANFORD TANK WASTE REMEDIATION SYSTEM PRIVATIZATION 

ACTIVITIES 

B1. PURPOSE 

The purpose of this Memorandum of Understanding (MOU) between the Nuclear Regulatory 

Commission (NRC) and the Department of Energy (DOE) is to establish the basis for 

cooperation and mutual support during the demonstration phase (defined as Phase I) of 

DOE's Tank Waste Remediation System-Privatization (TWRS-P) activities. An objective of 

this DOE/NRC interaction is the development and execution of a comprehensive regulatory 

program by DOE that is consistent with NRC's regulatory approach for protecting workers, the 

general public, and the environment. DOE's regulatory program is to be structured to facilitate 

the possible transition of regulatory responsibilities from DOE to NRC at the start of the 

full-scale operations phase (defined as Phase II). During Phase I, DOE is responsible for 

implementing the TWRS Privatization regulatory program. This MOU provides for cooperation 

and mutual support in an integrated effort that provides for: 

1. DOE to acquire capability to implement a program of nuclear safety and safeguards 

regulation consistent with NRC's regulatory approach. 

2. NRC to acquire sufficient knowledge and understanding of the physical and operational 

situation at the Hanford waste tanks and the processes, technology and hazards 

involved in Phase I activities, to enable NRC (a) to assist DOE in performing reviews in 

a manner consistent with NRC's regulatory approach and (b) to be prepared to develop 

an effective and efficient regulatory program for the licensing of DOE contractor-owned 

and contractor-operated facilities that will process waste at Hanford during Phase II. 

B2. INTRODUCTION 

1. Background 

During 1991, the DOE established the TWRS Program at the Hanford site to manage, retrieve, 

treat, immobilize, and dispose of certain radioactive waste in a safe, environmentally-sound, 

and cost-effective manner. The requirements and commitments for the TWRS cleanup 

activities are documented in the Hanford Federal Facilities Agreement and Consent Order, also 

known as the Tri-Party Agreement (TPA). Under the TPA, DOE, the U.S. Environmental 

Protection Agency (EPA), and the Washington State Department of Ecology have agreed to a 

timetable for cleanup of the Hanford site.

199 NUREG-1747
DOE, through the TWRS Program, is making a fundamental change in its contracting approach 

at Hanford, utilizing privately-owned facilities on the Hanford site for processing waste which 

contains special nuclear material. This change in contracting approach also necessitates a 

fundamental change in DOE's approach to regulation and oversight. 

To accomplish the TWRS requirements, DOE plans to privatize treatment operations for the 

Hanford tank wastes. The TWRS-P is divided into two phases, a demonstration phase (defined 

as Phase I) and a full-scale operations phase (defined as Phase II). During both phases, DOE 

will purchase waste treatment services from a DOE contractor-owned, contractor-operated 

facility under a fixed-price type of contract; DOE will provide the feedstock to be processed. The 

DOE TWRS-P Contractor must finance the project; design the equipment and facility; apply for 

and receive required permits and licenses; construct the facility and bring it on line; operate the 

facility to treat waste; and deactivate the facility. 

DOE will undertake nuclear safety and safeguards regulatory responsibility associated with the 

TWRS-P activities during Phase I. The EPA and the State of Washington have responsibility to 

regulate environmental issues and the Occupational Safety and Health Administration has 

responsibility to regulate occupational safety. NRC's participation during Phase I will primarily be of a cooperative nature foir the purposes of information transfer and assisting DOE in the 

establishment of a regulatory program that is consistent with NRC's regulatory approach for 

protecting workers, the general public, and the environment. 

This MOU describes the relationship between NRC and DOE for activities conducted during Phase I only. The relationship between NRC, DOE, and the DOE TWRS-P Contractors during Phase II remains to be clarified by legislation and/or regulatory requirements. 

2. Phase Descriptions 

Phase I 

Phase I is a proof-of-concept/commercial demonstration-scale effort. The objectives of 

Phase I are to: (a) demonstrate the technical and business viability of using privatized 

facilities to treat Hanford tank waste, (b) define and maintain required levels of safety and safeguards, (c) maintain environmental protection and compliance, and 

(d) substantially reduce life-cycle costs and time required to treat Hanford tank waste. 

Phase II 

Phase II will be the full-scale production phase, in which the facilities are to be 

configured so that all the remaining tank waste can be processed. The objectives of 

Phase II are to (a) implement the lessons learned from Phase I, and (b) process all tank 

waste into forms suitable for final disposal. The current DOE proposal is to have NRC 

assume full regulatory, responsibility (consistent with the manner in which NRC regulates its licensees) for Phase II, although certain operational, statutory, and regulatory issues 

must be clarified before the proposed Phase II regulation by NRC can be implemented. Current estimates are that DOE procurement documents and NRC regulatory 

requirements for Phase II would be needed by the year 2004. 

This MOU does not apply to Phase II activities.

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200
B3. AUTHORITY

1. Department of Energy 

Sections 31, 91, and 161 of the Atomic Energy Act of 1954, as amended; Section 104 of the 

Energy Reorganization Act of 1974; and, Section 301 of the DOE Organization Act authorize 

DOE to provide for the safe storage, processing, transportation and disposal of hazardous 

waste, including radioactive waste, resulting from nuclear materials production and weapons 

production. In addition, with regard to activities under DOE's jurisdiction, Section 161 .i.(3) of the 

Atomic Energy Act of 1954, as amended, permits DOE to prescribe such regulations or orders 

as it may deem necessary to govern DOE activities authorized by the Atomic Energy Act of 

1954, as amended, including standards and restrictions governing the design, location, and 

operation of facilities used in the conduct of such activity, in order to protect health and to 

minimize danger to life or property. 

2. Nuclear Regulatory Commission 

Sections 53, 57, 62, 63, 81, 103, 104, and 161 b, of the Atomic Energy Act of 1954, as 

amended, and Section 201 (f) of the Energy Reorganization Act of 1974 authorize NRC to 

license and establish by rule, regulation, or order, standards and instructions to govern the 

possession and use of special nuclear material, source material, or byproduct material to 

protect health or to minimize danger to life or property, or to promote the common defense and 

security. This agreement is entered into pursuant to these and other applicable authorities, 

including the Economy Act of 1932, as amended. 

B4. FOUNDATION UNDERSTANDINGS 

1. This MOU applies to Phase I only. 

2. DOE will regulate the DOE TWRS-P Contractors during Phase I under the terms and 

conditions agreed upon by DOE and the DOE TWRS-P Contractors, and will be 

responsible for the regulatory oversight of all design, construction, operational, and 

event-response activities. NRC will have no regulatory authority over the DOE TWRS-P 

Contractors during Phase I. 

3. No regulatory action, process, or practice established by DOE during Phase I will be 

binding on NRC during any possible NRC regulatory oversight of DOE TWRS 

Privatization Contractors during Phase I1. 

4. NRC's regulatory approach is based (a) on reviewing the applicant's systematic and 

integrated identification of potential accidents and interactions resulting from radiological 

and related process chemical and fire hazards, and (b) on ensuring adequate protection 

against those hazards which could impact on the safety of the worker, the general public 

and the protection of-the environment.

201 NUREG-1747
B5. AGREEMENTS BETWEEN PARTIES 

1. Responsibilities 

Department of Energy 

The Manager, Richlannd Operations Office, will be responsible for implementing the 

terms of this agreement. The TWRS Regulatory Official, who reports to the Manager, 

Richland Operations Office, will be the DOE point of contact for all communications 

relating to carrying out the provisions of this agreement. 

Nuclear Regulatory Commission 

The Director of the Office of Nuclear Materials Safety and Safeguards (NMSS) will be 

responsible for implementing the terms of this agreement. The Chief of the responsible 

Branch within NMSS will be the NRC point of contact for all communications related to 

carrying out the provisions of this agreement. 

2. General Provisions 

a. At the foundation of the DOE privatization approach is a predictability and reliability 

feature embedded in DOE's contracts with the TWRS-P Contractors-namely 

contractual commitments for DOE regulatory actions within specific time periods. 

Essential to timely- and orderly DOE regulatory actions is the awareness by NRC of 

these contractual commitments and the need for timely interaction between DOE 

and NRC at all levels. 

b. If an issue arises in the implementation of this MOU which cannot be resolved at 

the agency point-of-contact level, the NRC and DOE agree to refer the matter 

within 30 days to the Director, NMSS, and the Manager, Richland Operations 

Office, for appropriate action. 

c. It is the intent of both parties to conduct the TWRS Regulatory Program in an open, 

public, and professional manner. NRC and DOE recognize the importance of 

providing timely and accurate information to the public regarding regulatory matters 

that may affect the protection of workers, the general public, and the environment. 

Meetings between NRC and DOE staff in connection with this MOU will be 

governed by NRC policy on open meetings (59 FR 48340; September 20, 1994). 

NRC will participate with DOE in public meetings and other public interactions, as 

appropriate. All transmittals between DOE and NRC regarding TWRS Privatization 

activities will be made publicly available, consistent with NRC and DOE policies and 

requirements, at an established local public document room. 

d. Each agency recognizes that it is responsible for the protection, control, and 

accounting of classified, proprietary, and procurement-sensitive information; 

Safeguards Information (SGI); and Unclassified Controlled Nuclear Information 

(UCNI).

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202
e. Each agency will be responsible for processing, under its established program(s), allegations-declarations or statements or assertions of impropriety or inadequacy 

whose validity has not been established-associated with the regulated TWRS-P 

activities covered by this MOU. Each agency will keep the other agency informed, 

as appropriate, of such allegations, the allegations' status, and the allegations' 

resolution. Each agency will assure that allegations are promptly referred to the 

agency or entity that has jurisdiction over the allegation. 

f. support of the DOE TWRS Privatization activities, DOE will provide private office 

space and equipment, if needed, for NRC in the vicinity of the TWRS Regulatory 

Unit in the Richland, Washington area. DOE will provide the NRC with ready access 

to current TWRS regulatory information; access to key individuals in the Regulatory 

Unit for consistency discussions; access to TWRS general information, tank farm 

status and operational issues, and safety perspectives; and access to Hanford site 

safety perspectives. 

3. Regulatory Interaction Activities 

a. Site Familiarization 

NRC will need to acquire knowledge of the physical and operational situation for the 

Hanford waste tanks and of the processes, technologies, and hazards involved in 

processing the tank wastes. The following activities will be performed to provide this 

familiarization: 

1. NRC will visit the Hanford site, as necessary, to examine the conditions of the tank 

farms as they may relate to TWRS-P. As part of NRC's orientation, DOE will 

provide to NRC information on: 

"* The physical conditions and operational requirements necessary for safe 

storage, retrieval, transfer, and processing of the tank waste. 

"* Evaluations of the criticality potential for TWRS Privatization activities. 

"* Radiation levels of the waste and chemical forms of the waste. 

"* Contamination levels in the areas of the planned TWRS Privatization facilities 

and tanks. 

"* Hydrogen generation/flammable gas situation of tanks. 

"* Organic complexant/nitrate oxidizer situation of tanks. 

"* Other possible hazards associated with the waste. 

* Available or planned waste movement systems. 

* The Resource Conservation and Recovery Act (RCRA), Comprehensive 

Environmental Response, Compensation, and Liability Act (CERCLA), and 

Atomic Energy Act of 1954 for the TWRS.

203 NUREG-1747
II. DOE will provide! NRC access to the tank farms, tank farm records and 

documentation, and other information conceming operational conditions and events 

that NRC may desire in order to understand the TWRS Privatization project and 

associated hazards, processes, and conditions. 

Ill. Upon request by NRC, DOE will brief or hold discussions with NRC on issues 

related to the TWRS Privatization effort. The locations, timing, and content of these 

meetings will be agreed upon by the points of contact for each agency. 

IV. NRC may occasionally conduct reviews and special audits or inspections at DOE's 

request to provide! objective perspective on selected regulatory issues. 

b. Regulatory Familiarization 

To assist the DOE in establishing the capability to regulate consistent with NRC 

concepts and principles, the NRC will provide detailed briefings, guidance documents, and support in developing important administrative and technical program elements of a 

regulatory program. NRC will provide DOE access to regulatory training provided by NRC to its staff on a space available basis and, with specific agreement, will provide DOE opportunity to observe NRC's regulatory activities. 

c. Development of DOE TWRS Regulatory Program 

DOE guidance specific: to the regulation of DOE TWRS-P Contractors will be prepared and issued by DOE. The guidance is for use by the DOE's TWRS Regulatory Unit in its 

execution of the regulatory reviews and resulting regulatory actions and is provided as 

information to the DOE TWRS-P Contractors for their preparation of regulatory submittals. The guidance will cover those submittals required of the Contractors by DOE 

such as the Quality Assurance (QA) program, essential set of safety standards and 

requirements (including the site-specific design basis), integrated safety management plan, safety assessment, construction authorization request, operating authorization 

request, operational reports and assessments, and deactivation authorization. DOE will 

be responsible for issuiing this guidance in its final form. 

The following activities will be performed by NRC and DOE to develop the guidance: 

I. NRC will provide DOE with established and evolving NRC guidance and position documents as input for DOE to consider in the development and updating of its 

guidance for the DOE regulatory review. NRC will assist DOE in developing a DOE 

inspection program that will be applied during design, fabrication construction (e.g. acceptable codes and standards for concrete, electrical, welding, etc.), installation, 

and qualification testing. 

I1. DOE will develop guidance for the review of Contractor submittals and DOE reviews 

of TWRS-P activities. NRC will review and provide a basis 

for its comments on DOE's draft guidance to identify areas that may not be 

consistent with NRC's regulatory approach.

NUREG-1747 204
Ill. NRC will participate, as appropriate, with DOE in the joint development of guidance, 

based on industry standards, e.g., American Nuclear Society/American National 

Standards Institute (ANS/ANSI), for issuance by DOE as guidance for the DOE 

TWRS-P Program. 

d. Regulatory Program Implementation 

Specific DOE regulatory activities planned include design basis review, QA program 

evaluation, standards approval, initial safety evaluation, construction authorization and 

inspection, operating authorization oversight, and deactivation authorization. These 

actions will begin in Fiscal Year 1997 and continue throughout Phase 1. The following 

activities will be performed by DOE and NRC in fulfillment of their respective 

responsibilities under this MOU: 

1. DOE will be responsible for safety (e.g. design basis) and safeguards reviews and 

determining acceptability of DOE TWRS-P Contractors' submittals against the DOE 

TWRS guidance. DOE will have final decision authority for regulatory 

implementation during Phase I and for all interactions with the DOE TWRS-P 

Contractors. 

II. NRC will review and provide a basis for its comments on DOE TWRS-P 

Contractors' submittals to identify any areas that are not consistent with NRC's 

regulatory approach. These submittals will include all documents which address the 

technical and quality basis for the TWRS facilities and which could affect nuclear 

and process safety and safeguards in design, construction and operation. 

NRC will assist DOE in evaluating submittals and in verifying effective 

implementation of: 

"* Design: design basis, design verification, level of design detail and 

documentation, design specifications, calculations and drawings, and 

procurement specifications. 

"* Quality assurance for design, procurement, construction, pre-operational 

testing and operation. 

"* Operator training and qualification 

"* Human factors. 

* Emergency response.

205 NUREG-1747
B5. OTHER PROVISIONS

1. Nothing in this MOU will limit the authority of either agency to independently exercise its 

authority with regard to matters that are the subject of this MOU. 

2. Nothing in this MOU will be deemed to establish any right nor provide a basis for any 

action, either legal or equitable, by any person or class of persons challenging a 

government action or a failure to act. 

3. This MOU will be effective upon signature and upon satisfaction of conditions in Section 

VI.4 and will remain in effect until the end of Phase I. This agreement may also be 

terminated by mutual agreement or by written notice of either party submitted six 

months in advance of 'termination. Amendments or modifications to this agreement may 

be made upon written agreement of the parties. 

4. This MOU will become effective and remain in effect during such time periods when 

Congress authorizes and provides appropriate funding (or when there is another 

acceptable form of reimbursement) for NRC's participation in this project. 

5. Activities within the scope of this MOU and within the scope of appropriated resources 

are mutually agreed to be without reimbursement of cost for either organization. Special 

activities such as described in Sections V.c.1 .d and V.C.2 may be negotiated for cost 

reimbursement as needed.

NUREG-1747 206
APPENDIX C 

POINT PAPER SUMMARIES 

Cl. MATERIALS CONSIDERATIONS OF PROPOSED TANK WASTE REMEDIATION 

SYSTEMS 

Technical approaches are being formulated for the treatment of DOE tank wastes into vitrified 

waste forms suitable for long term storage and disposal. Tank wastes from the Hanford site 

are likely to involve significant processing efforts for separating radionuclides from 

nonradioactive species and concentrating them in the high level waste (HLW) glass. The 

appropriate selection of materials of construction and their inservice corrosion/erosion 

performance monitoring impact the safety of the facility primarily from the perspective of 

adequate confinement of radioactive materials, and this paper provides an overview of 

materials selection concerns based upon the knowledge of the design approach at this point in 

time, corresponding to perhaps a 5 percent design level (i.e., very preliminary design). Both 

general and specific corrosion/erosion are likely to occur over the Tank Waste Remediation 

System (TWRS) operating lifetime of some 30-plus years. Potential areas of the TWRS

Privatization (TWRS-P) facility where corrosion/erosion induced loss of confinement could 

result in the potential for radiation exposures exceeding regulatory limits have been identified as 

follows: 

1. Tanks. 

2. Pressurized transfer piping. 

3. Pump seal areas. 

4. Melters. 

5. Melter offgas treatment areas. 

Analyses estimate both unmitigated and mitigated consequences from potential accident 

scenarios in these areas, using a conservative approach suitable and normalized volumes for 

safety categorizations and preliminary designs. Several corrosion/erosion scenarios have 

potential accident consequences to the workers and the public of sufficient severity and risk (in 

the 2E-2/yr to 6E-2/yr range) such that prevention (reduce probability) and mitigation (reduce 

consequences) become necessary, requiring the identification of items relied upon for safety. 

These include the following: 

1. Leakage from HLW and low activity waste (LAW) storage tanks, with potential risks of 

1.4E-3/yr to 1.4E-4/yr. 

2. Pressurized transfer lines with potential risks around 4E-5/yr. 

3. Pump seal areas with potential risks up to 4E-2/yr. 

4. Corrosion failures in the HLW melter areas with potential risks up to 1.2E-2/yr. 

Potential risks from corrosion/erosion related loss of confinement of the separated cesium 

storage tank and LAW melter areas are lower in the 1 E-6/yr to 1 E-7/yr range.

207 NUREG-1747
There are no clear definitions of "acceptable risk." However, the NRC public dose limit from 

routine operations and the average U.S. worker risk (all causes) both correspond to around 5E

5/yr risk, and it would be anticipated that "acceptable risk" would be some percentage of this. A 

small percentage (0.1-0.5 percent) of the average cancer risk (currently around 2E-3/yr) has 

also been mentioned. It seems likely that the total, potential risk level for a TWRS-P like facility would have to be below 1 E-5/yr and probably close to 1 E-6/yr in order to be acceptable to the 

public. Therefore, some equipment is likely to be designated as relied on for safety in order to 

address corrosion/erosion loss of confinement concerns and reduce the potential risks. The 

two most likely approaches are prevention, which depends heavily upon operator actions to 

detect corrosion/erosion and avoids loss of confinement events, and mitigation, which relies 

upon the cell and ventilation system for additional confinement after failure. Ideally, the 

selection of the appropriate materials of construction, in conjunction with robust designs, redundant features, safety controls, adequately conservative corrosion/erosion margins, inspections, monitoring, and operator intervention are likely to provide adequate assurances of 

safety and offer the potential for reducing materials-related accident risk to more acceptable 

levels (circa 2E-6/yr). 

This effort has identified a relative lack of published corrosion and erosion information under 

expected operating conditions of the TWRS-P, particularly in the melter and offgas areas. Also, failure and release data for accident analysis are not well established for vitrification facilities in 

the expected TWRS-P environment. It is anticipated that additional information will become 

available from DOE and contractor efforts as the program continues to develop. 

This paper has considered selected materials-related safety considerations with respect to the 

proposed TWRS-P facility. Since radioactive materials will be confined in vessels, pipes, 

pumps vitrification melters, waste canisters, and other associated confinements, the appropriate performance- and life-limiting factors due to corrosion, erosion, and the combined effects of 

corrosion and erosion have been considered in general terms. However, the design of the 

TWRS-P facility is very preliminary and ongoing as of December, 1999. Reportedly, the current 

status corresponds to about 5 percent of preliminary design, with major design changes occurring. Detailed information on materials selection for various major process vessels, piping, and other confinement is still not available. Careful monitoring of this critical aspect should be considered when the Contractor submits the construction authorization request 

(CAR) in November 2000. 

C2. PROCESS SAFETY OF PROPOSED TANK WASTE REMEDIATION SYSTEMS 

Technical approaches are being formulated for the treatment of U.S. Department of Energy (DOE) tank wastes into vitrified waste forms suitable for long term storage and disposal. Tank 

wastes from the Hanford site are likely to involve significant chemical processing for separating radionuclides from nonradioactive species with the intent of concentrating the radioactive 

species in the vitrified HLW and concentration of the nonradioactive species (including most of 

the sodium) into vitrified LAW. The current plan envisions treatment in a private contractor 

owned facility on the Hanford site with subsequent storage of the vitrified HLW at a DOE facility onsite. Disposal of vitrified LAW occurs in a near-surface DOE disposal unit at Hanford. 

This paper presents analyses for chemical and process safety at potential TWRS-P facilities 

using generic and conceptual process approaches proposed by DOE contractors. These

NUREG-1747 208
analyses identify the following areas of concern from the perspective of chemical and process 

safety: 

1. Radiochemical inventories. 

2. Process efficacy. 

3. Organic ion exchange resin/nitrate interactions. 

4. Crystalline silicotitanate (CST) drying. 

5. Organic materials. 

6. Radiolysis. 

7. High temperature operations. 

8. Nonradioactive chemical effects upon radiochemical processing. 

Several of these areas of concern have events that can be analyzed at this early stage of 

design using a conservative, bounding approach suitable for structures, systems, and 

components (SSC) categorization. The remaining areas of concem are discussed qualitatively, but require a more detailed process and facility design for quantification. Analyses estimate 

both unmitigated and mitigated consequences and potential risks from these events. Several 

event scenarios involving resin interactions, CST, large radiochemical inventories, melter cold 

cap, and cold chemical releases have potential accident consequences to the workers and the 

public of sufficient severity such that the corresponding risk may not be acceptable, based upon 

comparisons to current limits. Consequently, prevention (reduce frequency and probability) and 

mitigation (reduce consequence severity) become necessary, requiring the identification of 

items relied on for safety. Analyses estimate the combined, unmitigated risk to the receptor at 

100 meters as approximately 2.4E-2/yr, about an order of magnitude above the equivalent risk 

of the 10 CFR Part 20 radiation worker annual dose limit of 5 rem. Melter and organic resin 

scenarios dominate the potential unmitigated risk at 100 meters, accounting for about 94 

percent of the risk total. However, the remaining risk from the identified potential events still is 

an order of magnitude greater than the average occupational risk for U.S. workers of 4.8E-5/yr. 

Fortunately, relatively simple and effective prevention and mitigation methods are available to 

reduce the potential risk from the TWRS-P facility. Passive prevention methods rely upon high quality, inspected, and tested components with conservative design and corrosion margins. This is a standard approach for nuclear facilities. Active prevention uses controls to avoid 

operating sequences outside the design envelope that are precursors to events with 

consequences; the controls can be based upon instrumentation or administrative procedures. Mitigation reduces consequences primarily by confinement (e.g., cells with high efficiency particulate air [HEPA] filters), with adequate levels of defense-in-depth. Again, this is a 

standard approach in the nuclear industry. Prevention and mitigation offer the potential to 

reduce the risk from TWRS-P operations to around 2E-6/yr. This result is less than 5 percent 

of the average occupational risk, around 1 percent of the risk from the average background 

dose, and around 0.1 percent of the average cancer fatality rate and is likely to be acceptable. 

Obviously, DOE and its contractors will include experimental testing as part of the program leading to the design, construction, and operation of the TWRS-P facility. Few appropriate safety related parameters, such as failure rates, modes, and release fractions, are available for 

HLW processing and vitrification facilities. It would be beneficial if the measurement of such 

safety parameters could be included in the DOE program.

209 NUREG-1 747
A potential rupture of cold chemical storage tanks containing ammonia and nitric acid would 

have onsite and offsite effects exceeding Emergency Response Planning Guideline-3 levels 

and require evacuation of the facility. Consequently, the facility design should include 

provisions to address such an event, by either dedicated breathing air to control and operator 

areas of the facility during Such an event, a remote control facility, automated 

operation/shutdown, effective mitigation of the chemical releases, or other means. In addition, 

some portions of the facility should be designed as shelter areas from potential events, 

including chemical releases.

NUREG-1747 210
APPENDIX D 

DISCUSSION OF POTENTIAL ISSUES FOR REGULATORY TRANSITION 

This attachment provides a discussion of issues related to the potential transition of regulatory authority for the Tank Waste Remediation System/Waste Treatment Plant (TWRS/WTP) facilities from U.S. Department of Energy (DOE) to the U.S. Nuclear Regulatory Commission 

(NRC) based upon their status as of June 2000. Many of these issues are summarized in Main 

Reference 17 and have been discussed between the DOE and NRC over the 4-year length of 

the program. From the viewpoint of the NRC staff, most of the issues would be addressed by the legislation that enables NRC regulatory authority over the TWRS/WTP facilities or NRC 

external regulation of DOE facilities, and by continued refinement and detailing of the proposed facility designs. The remaining issues relate to DOE programmatic activities and not regulation. If the NRC had remained involved in the program, continuing activities and design refinements 

might have changed some of the conclusions. 

D1 EMERGENCY PLANNING 

Issue 

Emergency planning is presently site-wide and is under the control of the DOE. Since the NRC 

normally regulates emergency planning at licensee facilities, this aspect of the TWRS/WTP 

contractor operation would become subject to NRC regulation. 

Discussion 

The emergency planning requirements of DOE and NRC are very similar. NRC requires some 

applicants for licenses under 10 CFR Part 70 to provide emergency plans.1°° Further guidance specific to the format and content for the emergency plans is given in Regulatory Guide 3.67.101 

DOE's requirements for emergency planning are contained in DOE Order 151.1,102 which 

supersedes the 5500 series of DOE Orders. However, the Hanford Emergency Response Plan'"3 still invokes the 5500 series of DOE Orders until the new Order can be implemented. Further detailed guidance is provided in the Emergency Management Guide."° The guidance 

100 Code of Federal Regulations, Title 10, Energy, Section 70.22(1)(ii). 

101 Nuclear Regulatory Commission (U.S.)(NRC). Regulatory Guide 3.67, "Standard Format and Content for 

Emergency Management System." NRC: Washington, D.C. January 1992. 

102 Department of Energy (U.S.)(DOE). DOE Order 151.1, "Comprehensive Emergency Management System." 

DOE: Richland, Washington. September 25, 1995. 

103 Department of Energy (U.S.)(DOE). DOE/RL 94-02, "Hanford Emergency Response Plan." DOE: Richland, 

Washington. 1994 

104 Department of Energy (U.S.)(DOE). DOE G 151.1, "Emergency Management Guide." DOE: Richland, 

Washington. August 1997.

211 NUREG-1747
in this document is very similar to that found in the NRC Regulatory Guide. It is, however, more 

specific. 

Both DOE and NRC have a classification scheme for emergency events. DOE suggests three 

levels (alert, site area emergency, and general emergency),"05 while the NRC guidance 

provides for two different emergency event classifications (alert and site area emergency).1°6 Both agencies support the use of offsite emergency resources in the event of emergencies and 

suggest written agreements between these local, state, and federal entities.'°',108 

Many DOE facilities have other facilities in close proximity and areas of public access (e.g., rivers, state or federal roads, etc.) within a short distance of the facility. In this situation, DOE 

usually develops a site-wide or reservation-wide emergency plan that incorporates the facility

specific plans. The approach is consistent with that of the NRC; however, more coordination of 

emergency planning activities; with local, state, and Federal agencies is required. The DOE 

Field Office usually provides this coordination, particularly when there are multiple contractors 

on a site.' 9 

In both its Integrated Safety Management Plan11,111 and its Initial Safety Analysis Report 

(ISAR), BNFL Inc. states that its emergency management plan will be implemented consistent 

with the Hanford Emergency Response Plan such that an integrated, timely, emergency 

response is assumed. 

DOE's emergency planning activities cover the entire Hanford Reservation, which 

encompasses many more facilities than just the proposed TWRS/WTP facilities, and provides the primary interface with the appropriate local and state agencies. Facility-specific emergency plans are also prepared by the contractor; however, these plans primarily contain the facility

specific information required 1-0 comply with DOE Orders and define the conditions under which 

the DOE Operations Office is apprised of an emergency and the coordination between the 

105 Department of Energy (U.S.)(DOE). DOE G 151.1, Emergency Management Guide, Vol. 3, Table 3.1, 

"Summary of Emergency Classes." DOE: Richland, Washington. August 1997. 

106 Nuclear Regulatory Commission (U.S.)(NRC). Regulatory Guide 3.67, Section 3, "Standard Format and 

Content for Emergency Plans for Fuel Cycle and Materials Facilities." NRC: Washington, D.C. January 1992. 

107 Department of Energy (U.S.)(DOE). DOE G 151.1, "Emergency Management Guide." DOE: Richland, 

Washington. August 1997. 

108 Nuclear Regulatory Commission (U.S.)(NRC). Regulatory Guide 3.67, Section 3, "Standard Format and 

Content for Emergency Plans for Fuel Cycle and Materials Facilities." NRC: Washington, D.C. January 1992. 

109 Department of Energy (U.S.)(DOE). DOE G151.1, Vol. 1, "Emergency Management Guide." DOE: Richland, 

Washington. August 1997. 

110 BNFL Inc. BNFL-5193-ISP-01, Rev. 2, "TWRS-P Project Integrated Safety Management Plan." BNFL Inc.: 

Richland, Washington. April 17, 1998. 

ill BNFL Inc. BNFL-5193-ISP-01, Rev. 0, "TWRS-P Project Integrated Safety Management Plan." BNFL Inc.: 

Richland, Washington. September 1997.

NUREG-1747 212
facility and DOE. As noted above, DOE then interfaces with the appropriate local and state 

authorities. 

BNFL Inc. has stated in its submittals that it will develop its emergency plans consistent with the 

Hanford Emergency Response Plan and that it will work with DOE and its contractors to provide 

for integrated emergency management of the Hanford site. 112 BNFL Inc. has also committed to 

design its emergency management program to work within the existing emergency 

management programs and to entering into agreements with DOE and the DOE contractors 

operating other facilities on the Hanford site to share resources and equipment.1 13 The NRC 

staff has reviewed Section 9 of BNFL's ISAR, which pertains to emergency management, and 

found that the format and content were consistent with the NRC's guidance. Furthermore, the 

NRC staff found no significant open items in the area of emergency management aside and 

that the DOE guidance is both generally consistent with the NRC's approach and more 

prescriptive. The NRC staff did find that the emergency management plans of TWRS/WTP 

should be integrated with those of the site, and responsibilities, authorities, etc., established in 

written documentation. 

If regulatory oversight were to transition to the NRC, then the TWRS/WTP contractor's 

emergency program would likely fall under NRC jurisdiction. NRC may require the contractor 

to modify its Emergency Response Plan to be consistent with the NRC guidance. For example, 

the discussion of emergency planning in the ISAR defines three categories of emergencies, 

which is consistent with the DOE guidance. If NRC were to assume regulatory oversight, then 

the TWRS-P emergencies may have to be reclassified and other changes may be required to 

provide compliance with the NRC Regulatory Guide.1 3 

The NRC staff has noted that the emergency classifications currently proposed by BNFL Inc. 

are consistent with the Hanford Emergency Response Plan but are inconsistent with the 

emergency event classifications in 10 CFR Part 70. The NRC staff noted that the TWRS-P 

project presents different hazards from those of the more traditional fuel cycle facilities so that a 

need for extending the emergency classifications to include the general emergency category 

might be warranted depending on the conclusions of the integrated safety analysis. However, 

the TWRS-P emergency event classifications would still not be consistent with the NRC 

guidance. The NRC staff has considered a number of alternatives ranging from modifying the 

regulations to allowing DOE to maintain the role of lead agency for emergency response. The 

alternatives ranging from modifying the regulations to allowing DOE to maintain the role of lead 

agency for emergency response. The alternatives also included using the licensing process to 

define the emergency event requirements. The NRC staff's concern is that, if the TWRS-P 

project had emergency event classifications different from other 10 CFR Part 70 licensees, this 

could be a source of NRC staff confusion in the NRC's Emergency Response Center during an 

actual event and that there could be confusion on training and procedures. 

The TWRS-P situation is somewhat analogous to the Washington Public Power Supply System 

(WPPSS) nuclear plants, WNP-1 and WNP-2, that are situated on leased land on the DOE 

112 BNFL Inc. BNFL-5193-ISP-01, Rev. 2, Section 3.10, "TWRS-P Project Integrated Safety Management Plan." 

BNFL Inc.: Richland, Washington. April 17, 1998. 

113 BNFL Inc. BNFL-5195-ISAR-01, Rev. 0, Section 9.1, "TWRS-P Project Initial Safety Analysis Report." BNFL 

Inc.: Richland, Washington. January 12, 1998.

213 NUREG-1747
reservation. However, the WPPSS facility is not situated in close proximity to other facilities, 

and there are no DOE workers in the immediate vicinity of the WPPSS facilities. WPPSS has 

established two Memoranda cf Understanding (MOU) with DOE; one for cooperation and 

mutual use of emergency response personnel, and a second for use of DOE's decontamination 

facilities in the event of personnel contamination. WPPSS also has established agreement with 

the local and state emergency response organizations. Similarly, the Siemens fuel fabrication 

facility, which as an NRC licenses under 10 CFR Part 70, is located on property that it owns but 

which is adjacent to the DOE: reservation. Like WPPSS, Siemens has established MOUs with 

DOE and the interested local and state agencies. 

Preliminary NRC Staff Assessment 

If the TWRS/WTP facilities were to transition to NRC regulation, the NRC staff would expect 

integration issues with other Hanford site operations to be fully addressed, communications and 

responsibilities clearly defined, and the appropriate authorities for the TWRS/WTP Emergency 

Plan (EP) managers to direct the Hanford site EP as necessary based upon the Integrated 

Safety Analysis (ISA). This would likely be formalized in MOUs similar to the existing NRC 

facilities on and adjacent to the Hanford site. The NRC staff does not view this as a significant 

issue. 

D2 RADIOACTIVE WASTE REGULATORY AUTHORITY 

Issue 

DOE is concerned about the regulator, authority, and requirements for the radioactive wastes 

generated by TWRS-Privatization (TWRS-P). 

Discussion 

The DOE plans for long term storage and/or disposal of the vitrified low activity waste (LAW) 

from the TWRS-P facility locally (i.e., on the Hanford site) in near-surface facilities. The vitrified 

high-activity waste will be stored locally and subsequently sent to a Federal repository for 

permanent disposal. 

The current framework for regulation of DOE radioactive waste provides for the NRC regulation 

of high level waste (HLW) and DOE regulation of other types of radioactive waste. A key 

consideration in identifying DOE-regulated waste is whether the waste was generated incidental 

to HLW reprocessing. Such wastes might include spent resins, failed melters, and high 

efficiency particulate air (HEPA) filters in addition to vitrified LAW. The critical importance of 

this determination is that waste determined to be incidental to HLW reprocessing is not 

classified as HLW and can be addressed by low level waste (LLW) regulations and 

storage/disposal facilities. Such facilities could be regulated by either DOE, NRC, or the State 

of Washington. If DOE decided to use existing DOE storage/disposal facilities at Hanford, 

these would not be subject to regulation by NRC. A significant portion of this issue is devoted 

to the examination of the nuances of identification of waste incidental to HLW reprocessing. 

Under the current regulatory structure, the NRC staff regards the identification of waste 

incidental to HLW reprocessing as the pivotal issue in assessing the adequacy of DOE plans 

for onsite disposal of LLW generated by TWRS-P.

NUREG-1747 214
The incidental waste issue stems from NRC's broad definition of HLW and the absence of a 

formal definition of incidental waste. HLW is defined as irradiated reactor fuel; liquid wastes 

from the first cycle of a solvent extraction system, or equivalent, and the concentrated wastes 

from subsequent solvent extraction cycles, or equivalent, in an special nuclear fuel (SNF) 

reprocessing facility; and solid materials into which those liquid wastes have been converted. 

Disposal of HLW requires a geologic repository. Land disposal requirements of radioactive 

wastes are contained in 10 CFR Part 61-such wastes are usually referred to as LLW. 

Incidental wastes are not mentioned in the regulations. 

Appendix F of 10 CFR Part 50 sets the NRC policy relating to siting of fuel reprocessing plants and related waste management facilities. The policy sets the time limits for on-site storage of 

high level radioactive wastes and also defines high level radioactive waste. In 1970, with the 

promulgation of 10 CFR Part 70, the Atomic Energy Commission specifically noted that the 

term high level waste did not include waste resulting from HLW reprocessing plant operations such as ion exchange beds, sludges, and contaminated laboratory items, such as clothing, tools, and equipment. These excluded wastes were referred to as incidental waste." 4 

There are operational an economical incentives to demonstrate that waste is not high-level waste."' Once the highly radioactive component of waste is separated out, the resultant less 

radioactive (and potentially more voluminous) LAW is like incidental waste and is in a regulatory 

dilemma; it is not clearly covered under present NRC regulations and may be disposed of in 

accordance with DOE orders for LLW disposal1 16 at DOE facilities provided the Washington 

Administrative Code (WAC) is satisfied. Under the present system, unless the NRC determines 

that this LAW/incidental waste is not HLW, the waste must be disposed of as HLW in a federal 

repository. However, if sufficient levels of radioactivity are removed from the LAW/incidental 

wastes, their potential hazards would be comparable to LLW. Hence, their disposal performance requirements could be achieved by near-surface facilities as well. 

In January and July 1990, the states of Oregon and Washington unsuccessfully petitioned the 

NRC to formally promulgate rules for determining that wastes generated incidental to HLW 

reprocessing are not HLW. In denying the petition for formal rulemaking, NRC restated the 

criteria for determining the waste to be incidental and not HLW." 7 The NRC criteria were also 

stated in a letter to DOE and are summarized below:'18 

Criterion One: Wastes have been processed (or will be further processed) to remove key radionuclides to the maximum extent that is technically and economically 

practical. 

114 Atomic Energy Commission (U.S.), Washington, D.C. Federal Register. Vol 35, p. 17533. November 14, 1970. 

115 U.S. Department of Energy (D. Wodrich) presentation to Oregon Department of Energy. "Classifying Hanford 

Tank Low Activity Waste Fraction." DOE: Richland, Washington. August 27, 1996. 

116 Department of Energy Order 5820.2A defines LLW as all radioactive waste that is not high level, transuranic, 

spent fuel, or byproduct material. 

117 Department of Energy (U.S.). Washington, D.C. Federal Register. Vol 58, p. 12342. March 4, 1993. 

118 Bemero, R., U.S. Nuclear Regulatory Commission, letter to J. Lytle, U.S. Department of Energy, March 2, 1993.

215 NUREG-1 747
Criterion Two: Wastes will be incorporated in a solid physical form at a concentration that 

does not exceed the applicable concentration limits for Class C LLW as set out 

in 10 CFR Part 61. 

Criterion Three: Wastes are t0 be managed, pursuant to the Atomic Energy Act, so that safety 

requirements comparable to the performance objectives set out in 10 CFR Part 

61, Subpart C, are satisfied. 

In short, the NRC position is tlhat the vitrified LAW and incidental wastes can be treated like 

LLW if they have similar characteristics and hazards as LLW. In November 1996, the DOE 

Richland Operations Office (DOE-RL) requested NRC agreement that the Hanford tank waste 

planned for onsite disposal is incidental waste (i.e., not HLW) and, therefore, would not be 

under NRC licensing authority." 9 The basis for this request was an RL technical basis report 

that included an interim performance assessment of the waste disposal site. The NRC 

concluded that the Hanford tank waste planned for onsite disposal met the criteria for 

classifying it as incidental waste; however, because of the preliminary nature of the tank waste 

performance assessment submitted by RL, several conditions were stipulated.120 

From the DOE/Regulatory Unit (RU) perspective, any agreement RL has in place with the NRC 

related to classification of incidental waste will not be affected by a transition to NRC regulation, 

so long as NRC grants the materials license to the TWRS-P Contractor on the basis of 

possession and not ownership of the waste. Under the current arrangement, DOE will retain 

title to all materials in the waste envelopes provided to the Contractor and in all intermediate 

and final waste products.121 The NRC regulation under which the TWRS-P Contractor would 

likely be regulated (10 CFR Part 70) does not specifically require the contractor to take title or 

ownership of the waste material being processed.'22 

The State of Washington is an Agreement State123 as provided by Section 274 of the Atomic 

Energy Act' 24 (AEA). That provision of the AEA permits individual states to assume, with the 

concurrence of the NRC, regulation of some activities that would otherwise be regulated by the 

NRC. Low-level radioactive waste management and disposal are among the nuclear activities 

for which regulation can be assumed by Agreement States and the State of Washington 

currently regulates such activities within their borders except at DOE facilities. Agreement 

119 Kinzer, J., U.S. Department of Energy, letter to C. Paperiello, U.S. Nuclear Regulatory Commission, 

November 7, 1996. 

120 Paperiello, C., U.S. Nuclear Regulatory Commission, letter to J. Kinzer, U.S. Department of Energy, June 9, 

1997. 

121 U.S. Department of Energy. TVVIRS Privatization Contract No. DE-AC06-RL1 3308, Section C.4.c, page C-9. 

DOE: Richland, Washington. September 26, 1996. 

122 Code of Federal Regulations, 77Ve 10, Energy, Section 70.1 (a), "Domestic Licensing of Special Nuclear Material." 

123 Agreement State means a state that has developed a program for regulation of those aspects of radioactive 

materials and radiation-producing devices not reserved exclusively for the NRC and has signed an agreement with 

the Commission to assume that responsibility. 

124 Atomic Energy Act of 1954, as amended.

NUREG-1747 216
State regulations must be compatible with NRC regulations, but need not be identical. Because 

Agreement State programs are essentially an extension of the NRC regulatory regime, 

transition to external regulation of DOE facilities in an Agreement State suggests that the 

Agreement State would likely be responsible for regulation of LLW generated by TWRS-P. Any uncertainty with respect to the appropriate regulatory entity for management of such wastes 

could be explicitly addressed in federal legislation mandating external regulation of TWRS/WTP 

or DOE nuclear activities. 

In a similar vein, Washington is a member of the Northwest Interstate Compact on Low-Level 

Radioactive Waste Management and has enacted into law the governing requirements125 of that 

organization. The compact approach to providing for management of LLW was created by Federal statute and was designed to encourage development of regional LLW disposal facilities 

in part to promote economic viability. The Federal statue12

" allows the compacts to exclude 

waste from outside the compact. However, the Northwest compact language does not require 

that all LLW generated within the borders of the member states be disposed of at the compacts regional site. The compact even allows individual LLW generators to establish and maintain 

LLW management facilities for their sole use.'27 Furthermore, Federal waste facilities are 

excluded from the definition of "facilities" for the purpose of the compact statute. 128 

Some radioactive materials from remediation of DOE facilities are being disposed of in onsite 

facilities permitted under the Comprehensive Environmental Response, Compensation, and 

Liability Act (CERCLA).'2 In an effort to eliminate potential redundancies in the requirements of CERCLA and the DOE Order on radioactive waste management, DOE adopted a policy' 30 

allowing for the substitution of CERCLA requirements to satisfy similar requirements of the 

DOE Order if the CERCLA requirements could be demonstrated to yield results equivalent to 

the DOE requirements. In addition, DOE has prepared guidance to clarify how compliance with 

the policy can be achieved."3' In fact, one of the DOE facilities that motivated DOE to develop 

and adopt the CERCLA policy was the Environmental Restoration Disposal Facility (ERDF) at 

the Hanford site. 

125 Revised Code of Washington, Title 43, Chapter 145, Section 010, "Northwest Interstate Compact on Low-Level 

Radioactive Waste Management." 

126 Low-Level Radioactive Waste Policy Amendments Act of 1985, Pub. L. 99-240, as amended (January 15, 1986). 

127 Revised Code of Washington, Title 43, Chapter 145, Section 010, Article IV, "Regional Facilities." 

128 Revised Code of Washington, Title 43, Chapter 145, Section 010, Article II, "Definitions." 

129 Comprehensive Environmental Response, Compensation, and Liability Act. 1980/1986. 

130 Department of Energy (U.S.)(DOE). CERCLA Policy: "Policy for Demonstrating Compliance with DOE Order 

5820.2A for Onsite Management and Disposal of Environmental Restoration Low level waste under the 

Comprehensive Environmental Response, Compensation, and Liability Act.," DOE: Washington, D.C. May 31, 

1996. 

131 U.S. Department of Energy, submittal to Defense Nuclear Facilities Safety Board, "Guidance for complying with 

DOE Order 5820.2A, 'Radioactive Waste Management' for Onsite Management and Disposal of Low level waste 

(LLW) Resulting from Environmental Restoration Activities," January 9, 1997.

217 NUREG-1747
Both the current DOE Order on radioactive waste management and the proposed draft revision 

of the Order 1 2 include the assumption that HLW is mixed waste unless demonstrated 

otherwise. Mixed waste is waste that includes both a radioactive component and a hazardous 

component (as defined in the U.S. Environmental Protection Agency (EPA) Resource 

Conservation and Recovery Act (RCRA) regulations'1). Such waste is subject to the 

requirements of EPA and to applicable radioactive waste management requirements. In 

addition, the Federal Facilities Compliance Act1 d imposes specific deadlines and other 

requirements on management of mixed waste at DOE facilities. An indirect consequence of the 

mixed waste assumption is that LLW and transuranic waste derived from HLW are likewise 

assumed to be mixed waste unless they are demonstrated to be otherwise or the HLW from 

which they are derived has been demonstrated to be otherwise. Fortunately, EPA regulations identify vitrification as the best achievable technology (BAT) for the hazardous constituents in 

HLW and LAW. 

In the authorizing legislation for NRC regulation of LLW, 1

' the responsibility for management of 

waste with radionuclide concentrations in excess of those allowed for Class C waste is 

delegated to the DOE. Consequently, the NRC has not promulgated regulations applicable to 

disposal of such waste. The IDOE, on the other hand, does not classify LLW in the same 

manner as the NRC. DOE does not have explicit requirements for greater than Class C 

(GTCC) waste, but rather manages it to meet the usual LLW performance requirements of 

DOE. Transition to external regulation of DOE may produce a situation in which there are no 

explicit requirements for GTCC waste. The NRC staff believes the waste from vitrification and 

decommissioning will exhibit radionuclide concentrations less than Class C. In its role as final 

arbiter of classifications of waste incidental to HLW reprocessing, NRC can disallow such 

classification of specific waste streams that exceed Class C concentrations. That practice would leave the HLW repository as the most likely disposal facility for such waste, and 

certification of such waste forms for emplacement in the repository would be required. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this a significant issue for regulatory transition because the 

TWRSIWTP facility is being designed so that the vitrified and incidental wastes will have 

radionuclide levels meeting Class C criteria or lower. Appropriate qualification testing and 

documentation will be needed. 

D3. SAFEGUARDS AND SECURITY 

Issue 

132 Department of Energy (U.S.)(DOE). DOE Order 435.1 (Draft), "Radioactive Waste Management," (distributed 

for internal departmental concurrence). DOE: Richland, Washington. July 1, 1998. 

133 Hazardous Waste Implementing Regulations under the Resource Conservation and Recovery Act 40 CFR 260 

through 270, U.S. EPA. 

134 Federal Facilities Compliance Act of 1992, Pub. L. 102-386, as amended. 1992 

135 Low-Level Radioactive Waste Policy Amendments Act of 1985, Pub. L. 99-240, as amended. (January 15, 1986).

NUREG-1747 218
The application of safeguards and security requirements for both the DOE and NRC licensed 

facilities is based on the classification of the materials processed and/or stored at the facility. It 

appears there are many similarities in the DOE and NRC classifications of the materials for 

TWRS/WTP, but there are minor differences. 

Discussion 

The physical protection requirements for various classifications of special nuclear materials are 

very similar between DOE and the NRC. The requirements for the protection of special nuclear 

materials are provided in the 5600 series of the DOE Orders, and the NRC requirements are 

contained in 10 CFR Part 73. Both DOE and NRC also recognize the differences in the 

quantities and attractiveness characteristics (isotope, materials type, concentration, enrichment, 

and physical form) of the materials in the application of those requirements.136 

The Hanford waste tanks include 177 tanks containing approximately 214 million curies in 

approximately 55 million gallons of waste. Most of the tanks contain several kilograms of U-235 

at low concentrations and enrichment levels. Pu-239, also at low concentrations, is contained 

in 18 of the tanks, but small quantities (10 to 100 grams occur in all the tanks. 

The Initial Safety Analysis Report (ISAR) for TWRS-P'3 characterizes the waste as containing "minor amounts of special nuclear material in the tank waste to be processed by the TWRS-P 

facility." 

The NRC notes that plutonium and U-235 are contained in the TWRS-P tanks in a Formula 

Quantity of Special Nuclear Material (i.e., greater than 5,000 grams - also called a Category I 

quantity). The protection requirements for Category I material are given in 10 CFR 73.45. The 

NRC would classify the materials as Category I since it is based on the characterization of the 

feed materials and the definition of Category I material as provided in 10 CFR 73.2. Section 

73.2 states, "Formula quantity means strategic special nuclear material in any combination in a 

quantity of 5,000 grams... This class of material is sometimes referred to as a Category I 

quantity of material." 

From the perspective of material control and accounting (MC&A), the NRC has indicated that 

because DOE sites contain different isotopes and higher concentrations or levels of enrichment 

of plutonium than NRC-licensed facilities, the categories in the NRC regulations do not overlap. The NRC has also indicated that sampling programs may be required for process control of the 

vitrification recipe as well as the pretreatment subsystems. 

To further complicate the differences in classifying the material, the ISAR states, "For the 

TWRS-P facility to successfully process this waste, it must separate the radionuclides from the 

diluted waste while taking into account the additional chemicals (e.g., chelating agents)." This 

raises the question as to what isotopes are to be separated, to what concentrations, where will 

they be located within the facility, and what levels of protection will be provided at those 

locations. Guidance provided to the TWRS-P Contractor states: "The material to be provided to 

136 Code of Federal Regulations, Title 10, Energy, Sections 70.4, 73.2, and 74.4 and U.S. Department of Energy 

Order 5633.3B. 

137 BNFL Inc. BNFL-5193-ISAR-01, Rev. 0, "Tank Waste Remediation System Privatization, Initial Safety Analysis 

Report." BNFL Inc.: Richland, Washington. January 12, 1998.

219 NUREG-1747
approaches have the potential to result in different consequence estimates that may produce differing classifications for systems, structures, and components (SSC) required for risk 

reduction to workers and the public. As a result, SSCs designated for worker protection may require upgrading to meet NRC requirements when regulatory oversight of the TWRS/WTP 

facilities transitions to the NRC. 

Discussion 

The DOE/RU position is that CLWs are a subset of workers and should not be classified as 

members of the public."45 

The classification of an individual as a CLW is found in DOE safety analysis reports and a 

proposed standard."',47 In the DOE proposed standard for accident analysis, the CLW is 

defined as a worker in a fixed population outside the daily process safety management controls 

of a given facility area. In practice, this fixed population normally refers to the workers at an 

independent facility area located some distance from the reference facility area. The 

classification is particularly relevant to accident planning and emergency management. 

Although the CLW classifications are not found in DOE rules, the definition for a general employee and a member of the public in the DOE rule 10 CFR Part 835, "Occupational Radiation Protection," provides the basis for the RU position that CLWs are considered a 

subset of workers rather than as members of the public." A general employee is an individual 

who is either a DOE employee or a DOE contractor employee, an employee of a subcontractor 

to a DOE contractor or a visitor or a DOE contractor employee, an employee of a subcontractor 

to a DOE contractor or a visitor who performs work for or in conjunction with DOE or utilizes 

DOE facilities.'49 As a worker at an independent facility area, a CLW is a general employee under that activity and may receive an occupational exposure. A member of the public means 

an individual who is not occupationally exposed to radiation or radioactive material. An 

individual is not a "member of the public" during any period in which the individual receives 

occupational exposure."° Thus, by this DOE logic, a CLW should not be classified as a 

member of the public because CLWs may receive occupational exposure under 10 CFR 

Part 835. It should be noted that, under this approach, the CLW does not have to have 

145 Department of Energy (U.S.)(DOE). RL/REG-98-18, Revision 0, "Regulatory Unit Position on Radiological 

Safety for Hanford Co-Located Workers." DOE: Richland, Washington. August 17, 1998. 

146 Department of Energy (U.S.)(DOE). EH-12-94-01, Vol. 2, Appendixes, "Method of Assessment of Worker Safety Under Radiological Accident Conditions at Department of Energy Nuclear Facilities." DOE: Richland, Washington. 

June 1994. 

147 Department of Energy (U.S.)(DOE). DOE-SP-STD-3005-93 Proposed, "Definitions and Criteria for Accident 

Analysis." DOE: Washington, D.C. March 5, 1993. 

148 Code of Federal Regulations, TitJe 10, Energy, Part 835, "Occupational Radiation Protection, Final Rule." 

December 1993. 

149 Code of Federal Regulations, Titbe 10, Energy, Section 835.2, "Definitions, General Employee." December 1993. 

150 Code of Federal Regulations, Title 10, Energy, Section 835.2, "Definitions, Member of the Public." December 

1993.

NUREG-1747 222
dosimetry, bioassay monitoring, evacuation drills, personnel protective equipment (PPE), or 

radiation training, but can receive occupational exposure up to 5 rem per year. Some examples 

of CLW by this approach include postal workers, delivery drivers, secretaries, normal trades 

contractors, food service vendors, and Native American Indians. 

DOE has not established a universal accident dose standard value for CLWs to be used for the 

design of nuclear facilities. Rather, DOE has proposed a method to assess worker safety 

under accident conditions."' 1 The RU referenced the proposed method in the top-level standards document, DOE/RL-96-0006.'2 This method of assessing the adequacy of 

radiological design provisions and safety under accident conditions for DOE facilities classifies 

some individuals as CLW.'5 3 DOE has reported on accident dose guidelines proposed by 

Contractors for radiological consequence as a function of accident frequency for workers, 

CLWs, and the public. 1

5 For radiological consequence assessment, the Contractor-proposed 

accident dose standards for CLW's are identical to the values for workers at the reference 

facility for unlikely and highly unlikely events. Some DOE accident assessments do not 

distinguish CLWs from workers at the reference facility."s The impact of the accident dose 

value selected from the CLW to the facility design and selection of safety provisions under 

accident conditions depends on numerous factors. These include the proximity of the CLWs to 

the reference facility, the nature of the site and meteorological conditions, and the specifics of 

the accident, source term, and release. 

Once the risk posed by an accident is quantified in terms of consequences to the worker, the 

CLW and the public, the need for safety controls/SSCs can be determined. Safety class SSCs 

are those systems, structures, and components required to mitigate the consequence or 

prevent an accident that poses an unacceptable risk to the public. Similarly, safety significant 

SSCs are required to prevent or mitigate the consequences or prevent an accident that poses 

an unacceptable risk to the worker or the CLW. Safety significant SSCs are subjected to more 

stringent design, fabrication, and construction requirements than SSCs that have no specific safety role. Safety class SSCs are subjected to even more stringent requirements than safety significant SSCs. Thus, the separate category of CLW might reduce the requirements of some 

SSCs. 

The NRC approach usually does not recognize such an entity as the CLW. Other workers and 

people outside the specific facility and its area yet working on or visiting the DOE Hanford site

for any reason-would be considered members of the public. This is because such CLW may 

be operating under different management organizations and radiation protection programs, and 

151 Department of Energy (U.S.)(DOE). EH-1 2-94-01, Volume 1, Main Report, "Method for Assessment of Worker 

Safety under Radiological Accident Conditions at Department of Energy Nuclear Facilities." DOE: Richland, 

Washington. June 1994. 

152 Department of Energy (U.S.)(DOE). DOE-RL-96-0006, "Top-Level Radiological, Nuclear, and Process Safety 

Standards and Principles for TWRS Privatization Contractors." DOE: Richland, Washington. February 1996. 

153 Department of Energy (U.S.)(DOE). EH-12-94-01, Volume 1. DOE: Richland, Washington. June 1994. 

154 Department of Energy (U.S.)(DOE). EH-12-94-01, Volume 1. DOE: Richland, Washington. June 1994. 

155 Department of Energy (U.S.)(DOE). E H-12-94-01, Volume 1. DOE: Richland, Washington. June 1994.

223 NUREG-1747
hence, the integration and control may be inadequate for managing the organizations and 

potential events. For TWRS-P, DOE plans to use a large portion of the Hanford site to define 

CLW, which typically results in minimum distances to members of the public (i.e., with lower 

dose and risk limits) of approximately 10 miles. In contrast, the NRC focuses on the concept of "controlled area" by the facility operator; at NRC facilities this is usually associated with a fence 

although the controlled area can extend beyond the fence and even beyond the site boundary. Thus, the NRC approach usually corresponds to a shorter distance (usually 100-200 meters) to 

the public for accident analysis purposes, which can translate into more items relied on for 

safety [IROFS]. However, the key concept is the TWRS/WTP operator's authority to exercise 

control over the Hanford site emergency plans; if such authority is granted by DOE to the 

TWRS/WTP operator, then the "controlled area" concept may be satisfied, and the issue is 

moot. This is the approach under discussion with a proposed mixed oxide (MOX) facility at 

Savannah River that will undergo NRC licensing15. 

The NRC approach uses accident analyses to estimate consequences (for workers and the 

public) and compare these to standards and limits in the regulations and guidance. Certain 

NRC regulations, including 10 CFR Part 72, provide numerical standards for evaluating the 

impacts of accidents on individual members of the public at the controlled area boundary. The 

revision to 10 CFR 70 includes accident standards for the facility worker, members of the 

public, and the environment but not for CLWs. In addition, the dose limits require comparison. Under the TWRS-P regulatory program the dose limits were 25 rem for the worker, CLW, and 

the public for highly unlikely events (the public had a target dose goal of 5 rem). In DOE, highly unlikely corresponds to the frequency range of 1 E-4/yr to 1 E-6/yr. From the perspective of the 

NRC staff, the revised 10 CFFI Part 70 and standard review plan (SRP) have a worker dose 

limit of 100 rem and a public dose limit of 25 rem for high consequence events, and 

corresponding limits of 25 remn and 5 rem for intermediate consequence events. High 

consequence events are to be rendered highly unlikely (1 E-5/yr or less in frequency) by safety controls, and intermediate consequence events are to be rendered unlikely (in the 1 E-2/yr to 

1 E-5/yr frequency range) by safety controls. Thus, at face value, the DOE limits are more 

restrictive although the fuel cycle SRP157 does allow grading of the frequency limit in inverse 

proportion to the magnitude of the consequences and this could result in the limits overlapping between frequencies of 1 E-51yr and 1 E-6/yr. The application of as low as reasonably achievable (ALARA) (required under either DOE or NRC regulation) would probably render 

such differences insignificant. 

As an aside, the NRC notes the DOE regulatory approach should consider future site changes planned that are likely to reduce the distances to the public for accident evaluation purposes 

(see Issue A.13 in Appendix A). 

Preliminary NRC Staff Assessment 

156 Persinko, A., U.S. Nuclear Regulatory Commission, memorandum to M. Leach, U.S. Nuclear Regulatory Commission, "Summary of Meeting with Duke Cogema Stone & Webster to Discuss Technical Topics Associated 

with the Mixed Oxide Fuel Fabrication Facility," February 24, 2000. 

157 Nuclear Regulatory Commission (U.S.)(NRC). NUREG-1520, "Standard Review Plan for the Review of a 

License Application for a Fuel Cycle Facility." NRC: Washington, D.C. 2000.

NUREG-1747

224
The NRC Staff considers this to be a resolvable issue either by the application of appropriate 

controls over the site or the application of ALARA. 

D6. WASTE OWNERSHIP 

Issue 

The NRC would most likely regulate a TWRS/WTP facility under 10 CFR Part 70, a special nuclear materials license. The contract provides that DOE retains ownership of the waste 

material, even when it is inside the Contractor's facility. DOE is concerned this part of the 

contract may require revision, transferring material ownership to the Contractor when it enters 

his facility and back again to DOE when it leaves. 

Discussion: 

Section C.4.c of the contract between DOE and BNFL Inc. 1

5 states that DOE will retain title to 

the waste envelopes provided to the Contractor and all intermediate and final waste products. If regulatory oversight were to transition to NRC, DOE has a concern that the NRC may require transfer of the ownership of the waste to the Contractor as long as the intermediate and final 

waste products remained in TWRS/WTP Contractor facilities. Transfer of ownership of the 

waste was not provided for in the DOE contract with the TWRS-P Contractor; thus, the contract 

would have needed renegotiation if the NRC required waste ownership transfer. Information on 

waste ownership is not yet available for the new Management and Operations (M&O) style 

contracts proposed by DOE that replace the privatization approach. 

From an NRC perspective, the TWRS-P facilities would likely be subject to a license under 10 

CFR Part 70 which pertains to the issuance of licenses to receive title to own, acquire, deliver, receive, possess, use, and transfer special nuclear material.'" 9 The TWRS-P facility would be 

required to obtain a specific license under 10 CFR 70.22 to possess the special nuclear 

material in the TWRS waste; however, transfer of ownership of the waste is not required by 10 

CFR Part 70. 

NRC has issued special licenses to a number of licensees for possession of DOE-owned 

material. For example, Pacific Northwest National Laboratory was issued a special license for 

the conduct of research and development activities in the 300 Area of the Hanford site using DOE-owned fissile material. A private company, Applied Radiant Energy Corporation in 

Lynchburg, Virginia, has also received a special license for its possession of DOE-owned 

cesium capsules. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this an issue. 

D7. PAYMENT FOR TRANSITION TO AND REGULATION BY THE NRC 

158 Contract Number DE-RP06-96RL1 3308, Amendment A0005, August 24, 1998. 

159 Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear Material."

225 NUREG-1747
Issue

Regulation of TWRS/WTP facilities requires funding. 

Discussion 

Regulatory costs are difficult ito estimate accurately, especially for a new type of plant for which 

there is no licensing experience. Presently, the regulatory costs for the TWRS-P effort are born 

by DOE and NRC; both from budgeted line item funding. It is expected that, during transition, 

regulatory costs will increase for the NRC and decrease for DOE. 

The funds for the current NRC activities related to the TWRS-P facility were a line item in the 

budget of about $2 million annually"6° (for approximately 13 full time equivalents [FTEs]). For 

comparison, the DOE RU costs for developing and executing a full scope nuclear, radiological, and process safety regulatory program have remained between $7 and $8 million per year, roughly split equally between contractor and federal employee. The average annual cost of one 

NRC staff person is around $260,000.00.'6' Therefore, $8 million corresponds to about 32 

FTEs at the NRC. Given that the NRC has already developed regulations and guidance for 

regulating TWRS/WTP facilities, it is reasonable to expect that NRC costs would be 

significantly lower than $8 million for regulatory transition and licensing/certification activities, 

prior to full and stable operation. Following license issuance and full transition to NRC 

regulatory oversight, these costs would be expected to decrease slightly. NRC is, by statute, a 

full recovery agency for which all costs of operation are recovered by charges levied upon licensees.'62 Surcharges could also be added to cover other NRC overhead costs. Over a 10

year period, it would be expected that the costs for NRC regulation would be lower than the 

current costs of the regulatory activities. 

There is no legislative bar to NRC assessing annual fees (under 10 CFR Parts 170 and 171) to 

Federal agencies who hold NRC licenses.1' 3 However, for either privatized or M&O facilities, 

DOE is not expected to be a licensee and, therefore, would not be subject to direct NRC fee 

assessment. 

Preliminary NRC Staff Assessment 

NRC involvement in TWRS-P has been funded via a line item in the budget. Future regulation 

of a TWRS/WTP facility by tihe NRC would require legislation that would also identify the 

160 NRC line item budget for the Special Projects Branch regulatory activities for TWRS in 1998 is $1.9 million plus 

about $500 K for consultants' fees. 

161 Code of Federal Regulations, Title 10, Energy, Section CFR 170.20, "Average Cost per Professional Staff-Hour." 

162 Omnibus Budget and Reconciliation Act of 1990. 

163 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-050, "Development of Legislative Issues for Licensing a Mixed Oxide Fuel Fabrication Facilibt," pp. 33-35. NRC: Washington, D.C. March 16, 1998. [This reference 

provides an analysis of who pays NRC relative to the MOX fuel fabrication facility in which DOE is currently paying for NRC assistance on design and operation issues where DOE becomes the licensee. However, the analysis does 

not apply to the vitrification facility if :!he TWRS/WTP contractor were to become the licensee.]

NUREG-1747 226
funding mechanism, be it by line items, fee collection, DOE payments, or a combination thereof. 

It is anticipated that the NRC regulatory costs would be a small fraction of the actual DOE 

expenditures to the contractors on the tank waste programs. The NRC Staff does not see 

funding of regulatory activities as a significant issue. 

D8. TRI-PARTY AGREEMENT 

Issue 

DOE is a party to the Tri-Party Agreement (TPA) and is responsible for TPA commitments. 

DOE has expressed a concern that NRC regulatory activities during and after regulatory 

transition could impact the TPA commitments, and, thus, DOE might request the NRC to 

become involved in the TPA. 

Background 

The Hanford Federal Facility Agreement and Consent Order (often referred to as the Tri-Party 

Agreement, or TPA1

") is a legal agreement between DOE, EPA, and the State of Washington 

Department of Ecology (Ecology). Its legal authority stems from the RCRA and the CERCLA. 

The TPA contains provisions for the overall environmental management of the Hanford site. 

This includes provisions for management of hazardous waste treatment, storage, and disposal 

(TSD), and hazardous waste permitting. 

The TPA includes details concerning the implementation of remedial and corrective actions 

including closure and post-closure activities. The TPA defines the respective roles, 

responsibilities, and interrelationships between DOE, EPA, and Ecology concerning these 

environmental management activities. It also defines interrelationships and responsibilities between RCRA and CERCLA. In addition, the TPA delineates authorities, identifies 

enforcement provisions, and provides for dispute resolution among the parties.' 

The TPA contains an Action Plan that establishes plans for compliance with RCRA and 

CERCLA and the Washington State Hazardous Waste Management Act (HWMA). In addition, 

the Action Plan sets milestones for CERCLA and RCRA activities that must be met by DOE. 

Failure to meet these milestones can result in fines and lawsuits. These milestones represent 

the actions necessary to ensure acceptable progress toward Hanford site compliance with 

RCRA, CERCLA, and HWMA.`3 

The TWRS-P Contractor had agreed via Clause H.22 of the contract1 65 to plan and perform work consistent with the requirements of the TPA even though it is not a signatory. It would be 

anticipated that future contractors under the new contracts would be asked to do the same. 

As discussed in Section 1, DOE is using contract-based regulation for the regulatory oversight 

of the TWRS/WTP waste immobilization contractor(s). In the context of contract-based 

164 Department of Energy (U.S.)(DOE). DOE/RL-96-25, Revision 0, "Policy for Radiological, Nuclear, and Process 

Safety of TWRS Privatization Contractor." DOE: Richland, Washington. July 3, 1996. 

165 Department of Energy (U.S.)(DOE). "TWRS Privatization Contract No. DE-RP-96RL13308, Amendment A0005." 

DOE: Richland, Washington. August 23, 1998.

227 NUREG-1747
regulation, the regulatory agency has an incentive to conduct the necessary regulatory activities 

in such a manner that the project schedule is not unduly delayed and DOE's external 

commitments are satisfied, while simultaneously adhering to the principle of independence (from the DOE-RL TWRS program office) and its primary role of ensuring adequate safety1

6 

(Main Reference 17). The NRC would have no vested interest in regulating the TWRS 

privatized waste immobilization contractor in such a manner that would assure that DOE's TPA 

milestones could be satisfied '(Chapter 7.0, Main Reference 17). 

The NRC has assumed regulatory oversight of several DOE facilities or former DOE facilities. 

These facilities include the gaseous diffusion plants (GDPs) at Paducah and Portsmouth (in this 

case, DOE has leased the facilities to the U.S. Enrichment Corporation, which is the licensee to 

the NRC), and a storage facility at the Idaho National Engineering and Environmental 

Laboratory (INEEL) for dry storage of Three Mile Island waste. In all cases, the NRC has not 

become a party to the existing Federal Facility Compliance Agreements. In fact, the NRC 

enabling legislation probably would preclude it from becoming a party to the TPA. Furthermore, 

the NRC Principle of Independence would be compromised since its regulatory decisions would 

have to be made in the context of their impact upon schedules, milestones, and costs. Thus, 

the NRC would not attempt to become involved in the TPA and DOE would maintain sole 

responsibility for satisfying the provisions of the TPA. 

The NRC does not generally commit to support external schedules such as licensee review 

schedules or the TPA. Any NIRC regulatory actions that might affect DOE commitments to the 

TPA would probably be based upon significant findings and issues that would affect TPA 

commitments regardless of NRC involvement. 

Preliminary NRC Staff Assessment 

DOE is a party to the TPA anrd is responsible for TPA commitments. The NRC staff would 

expect the NRC to remain a non-party to the TPA, consistent with other regulators (e.g., Defense Nuclear Facilities Safety Board (DNFSB), WDOH) and NRC precedence. The NRC 

staff does not see this as a significant issue. 

D9. DOE STOP WORK AUTHORITY 

Issue 

"Contract-based" regulation has been used on TWRS-P. A transition contractor is in place and 

new contracts will be issued. "he DOE is unsure if it should retain its stop work authority under 

the contract and, if it does, should DOE conduct readiness reviews. Such activities could 

conflict with NRC regulatory oversight. 

Discussion 

From a regulatory perspective, DOE would not have any safety concerns for TWRS-P 

operations in Phase II that could not be adequately protected by NRC, the Occupational Safety 

166 Department of Energy (U.S.)(DOE). RL/REG-97-10, Rev. 1 'Radiological, Nuclear, and Process Safety 

Regulation of TWRS Privatization Contractors Regulatory Plan." DOE: Richland, Washington. January 7, 1998.

NUREG-1747 228
and Health Administration (OSHA), the EPA, and the State of Washington oversight. However, 

DOE may choose to retain a residual stop work authority in order to safeguard against the 

unknown and uncertainties. 

The vitrification of DOE waste will be in the hands of a single contractor or contractor team. 

DOE will likely maintain oversight to ensure that the contractor meets its programmatic 

commitments to DOE to immobilize the waste in the tanks on an agreed-upon schedule. DOE 

will wish to protect its interests and commitment, in the event that the contractor delivers a 

product to DOE that fails to meet technical specifications satisfactory to DOE under the terms 

of the contract, by insisting on a non-safety-related "stop work" provision. This provision may 

prevent the production of more "off-spec" vitrified material. An appropriate method for handling 

(and possibly, reprocessing) off-spec products would need to be arranged by DOE. Fixing the 

technical problems or operational problems might involve a review of the contractor's technical 

approaches to vitrification of DOE waste {a "technical readiness review") by DOE prior to further 

waste processing. DOE might also incorporate a stop work authority for safety purposes into 

the contract. 

If regulatory oversight responsibilities were to transition to the NRC, the contract(s) would 

require modification to define NRC as the authority for the radiological and nuclear process 

safety aspects of the TWRS-P effort. As a regulator, the NRC would anticipate DOE 

maintaining stop work authority in some form under its contracts, including retaining the 

authority to conduct a readiness review prior to restart. The NRC staff believes such activities 

would be transparent to its regulation provided that all other regulatory requirements and 

licensing conditions were being met by the TWRS/WTP facility. 

Preliminary NRC Staff Assessment 

The DOE currently has stop work authority for safety and programmatic concerns under the 

contracts. The NRC Staff anticipates both the NRC and DOE would have stop work authority 

for safety concerns after regulatory transition, and the Staff does not consider this to be an 

issue. 

D1O. CONTINUED SAFETY OVERSIGHT BY DOE 

Issue 

The DOE currently has total programmatic responsibility for the TWRS-P project. That 

programmatic responsibility includes a safety oversight content, which is exercised by the RU. 

It should be clear how DOE's safety responsibilities will be discharged or modified by external 

regulation by the NRC. 

Discussion 

The Contractor, as the owner of the plant, has the prime responsibility for safety. TWRS/WTP 

facilities will be built on the DOE Hanford site and DOE expects to retain ownership of the 

waste materials at all times. While the new contracts may not be finalized for several months, it 

is anticipated that some form of contract reimbursement or incentive will be included for the 

production of containers containing the vitrified wastes.

229 NUREG-1747
In the wider world of commercial chemicals, toll-processing of chemicals may include 

operations for compounding, formulating, milling, custom blending, repackaging or separating 

and purifying materials which remain the property of the supplier. Companies toll-processing chemicals for others are answerable to OSHA for maintaining a safe workplace for employees and to EPA for protection of the environment. Operational safety is the responsibility of the toll

processor; the responsibility of the chemical supplier is to ensure that the toll-processor is made 

aware of the hazards associated with the process materials and does not usually extend to 

ensuring that the toll-processing is conducted safely. Typically, a chemical company wishing to 

subcontract out certain steps involved in the manufacture of a product will invite bids for the 

work and, for strategic reasons, provide feedstock to two or more toll-processors. The chemical 

supplier will want to ensure that the toll-processors are technically competent to perform whatever it is their tasked to do, meet the supplier's product specifications, and deliver the 

product on a mutually agreed upon schedule. At no time will the chemical supplier wish to take 

responsibility for safe operation of the toll-processor's facility, even if that facility is located 

adjacent to that of the chemical supplier's. However, if the supplier and toll-processor operate 

contiguous facilities, the two companies may agree to support each other in emergency 

planning and firefighting operations if an accident should occur. 

For Hanford and TWRS/WTP, it is anticipated that DOE will take more interest in the safety 

aspects than a company contracting for toll-processing of materials due to the singular nature 

of the wastes, the plant location on the Hanford site, the presence of DOE facilities in close 

proximity to the planned location of the vitrification facility, the high profile of the project, and the 

large costs involved. 

Preliminary NRC Staff Assessment 

The DOE currently is the reguilator for the proposed TWRS/WTP activities. After transition, the 

NRC would become the regulator. The NRC anticipates the DOE would want to perform some 

safety oversight activities, perhaps in a manner analogous to a corporate headquarters unit 

providing oversight to an operating facility. The NRC staff does not consider this to be an issue 

for regulatory transition. 

Dl. THE APPLICATION OF 10 CFR PART 2 HEARING REQUESTS 

Issue 

Under NRC regulation, the licensing of TWRS-P might be subjected to hearing requests under 

the "Rules of Practice," 10 CFR Part 2. This may require some time and has the potential to 

impact DOE schedules. 

Discussion 

The Rules of Practice, 10 CFR Part 2, provide for hearings. An Administrative Law Judge may be appointed whose principal objective would be to ensure that the process for airing all 

propositions is fair and equitable. 

The NRC staff understands that 10 CFR Part 2 applies to 10 CFR Part 70 actions unless 

Congressional action is taken through amendment of the Administrative Procedures Act or as 

part of the legislation authorizing NRC regulation of TWRS/WTP. However, the NRC Staff

NUREG-1 747 230
believe that 10 CFR Part 2 hearings under 10 CFR Part 70167 (without the Atomic Safety 

Licensing Board [ASLB] or discovery) are more focused on specific issues than general 

licensing and rulemaking hearings 1

' (with the ASLB) and are intended to conclude 

expeditiously. Hearings requested for granting a license to possess special nuclear material 

under this section of the code may be conducted according to Subpart 1 of 10 CFR Part 2. 

The certification of the GDPs, licensed under 10 CFR Part 76, required special legislation and 

was not subjected to a 10 CFR Part 2 public hearing process. This was because these were 

existing facilities and the facilities were certified rather than licensed. For Hanford, either 

licensing or certification could be used for NRC regulation. 

Preliminary NRC Staff Assessment

Whether the route is licensing, certification, or Congressional mandate the NRC staff does not 

foresee a significant schedule impact from the hearing process and, therefore, does not 

consider this to be a significant issue for regulatory transition. 

167 Code of Federal Regulations, Title 10, Energy, Part 2, subpart (1), Section 2.1205(h) "Request for a hearing: 

petition for leave to intervene," and Section 2.1209, "Power of Presiding Officer," which give wide latitude to the 

presiding officer in determining that "the specified areas of concern are germane to the subject matter of the 

proceeding and that the petition is time." Section 2.1211 (a), "Participation by a person not a party" allows a person 

who is not a party to make a limited appearance to state his of her views on the issues. A limited appearance is not 

considered part of the decisional record. 

168 Code of Federal Regulations, Title 10, Energy, Part 2, subpart (h), Section 2.805 (a), "Participation by interested 

person," deals principally with rulemaking and affords opportunity for interested persons to participate "through the 

submission of statements, information, opinions, and arguments." There may also be informal hearings for 

interested persons. The opportunity on general matters appears much greater than allowed under Section 2.1211.

231 NUREG-1747
D12. CONTRACTUAL OBLIGATIONS FOR FEED DELIVERY 

Issue 

DOE will continue to deliver radioactive feed material to the Contractor. Presumably, DOE will 

agree to abide by the direction of the NRC-regulated Contractor as to the rate and 

characteristics of this feed material. 

Discussion 

The issue concerns the translfer of DOE-owned tank waste to the TWRS/WTP facility after 

transition to NRC regulation. The waste specifications in the current contract are principally based on a hypothetical Phase I plant which is operated for only 5-9 years. The facility is being designed to be operable for a period that can reasonably be expected to enable it to treat waste 

that is not enveloped by the current specifications in the contract. It is estimated that 

approximately 5 percent of the waste in the tanks is outside the envelope determined by the 

contract specifications. Although there exists conceptual solutions for dealing with this situation 

(e.g., blending with other tank contents or inert substances), the details of the waste 

characteristics may not become known until after their contents have been homogenized, which 

will occur only shortly before processing. 

Preliminary NRC Staff Assessment 

The NRC staff expects there wNill be specifications and safety requirements for the wastes and 

the plant operations. If the wastes cannot be blended, then an amendment process could be 

pursued by the licensee (this 'has already occurred at the GDP's). This is not an issue for 

regulatory transition. 

D13. COMMUNICATION PLAN WITH STAKEHOLDERS 

Issue 

There are regulatory requirements for soliciting comments from the public and stakeholders. 

Discussion 

From the DOE perspective, input from interested stakeholders needs to be solicited. This 

includes the Contractors themselves, consensus groups such as the Hanford Advisory Board 

(HAB) and the Tribal Nations, Hanford workers, and other members of the public who have 

expressed interest. 

DOE has a responsibility to obtain the input of stakeholders and the Tribes as the effective 

clients of all the agency's work. Furthermore, the RU has an Openness Policy1 69 that makes it 

169 Department of Energy (U.S.)(DC'E). RL/REG-97-04, Rev. 2, "Policy for Openness and Openness Plan." DOE: 

Richland, Washington. June 1998. The radiological, nuclear, and process safety regulation of TWRS Privatization 

contractors provide that the Hanford site shall be transacted publicly and candidly.... "The Tribal National and Public 

should be involved in decisions concerning privatization contractor regulations. Their involvement improves our processes and products and helps (to) ensure safety. We welcome and encourage this involvement."

NUREG-1747 232
clear that stakeholders' views are important and will be addressed. The RU is satisfying this 

policy in a number of ways, both in providing information and in listening to the views of others 

in open meetings and within such groups as the HAB. However, regulation does not occur in a 

vacuum, and the NRC recognizes DOE's responsibilities for communication1"'. The NRC also 

requires it of its licensees. 

Previous DOE attempts at openness, particularly in interactions with the Agreement Tribes, 

have been to provide all DOE documents without focus or any attempt to indicate to the 

recipient as to the relevance of those documents. From the recipients' point of view, this flood 

of paper obscured communication."' 

Preliminary NRC Staff Assessment 

The NRC anticipates that the existing DOE communication plans (e.g., with the Hanford 

Advisory Board) would continue during and after transition to NRC regulation, as they are 

separate from the regulatory processes. In addition, the NRC also maintains public 

communication and openness as part of its regulatory practices. Therefore, the NRC staff does 

not consider this a significant issue. 

D14. COST BENEFIT OF NRC REGULATION 

Issue 

DOE is concerned that NRC regulation might increase the costs of tank waste treatment. 

Discussion 

As background, it should be noted that the cost of NRC regulation is greater than that actually 

expended on an individual facility,"7 because collected fees include the costs of NRC's 

overhead activities. 

From the perspective of Hanford and DOE, the issue is whether the transfer of regulatory 

authority from DOE to NRC provides a net benefit in safety for the additional costs involved, 

assuming that actual NRC costs will also include a surcharge to fund the above items.' 73 It is 

not clear whether surcharges might be added to NRC fees for TWRS-P regulation. 

170 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-050, "Development of Legislative Issues for Licensing a 

Mixed Oxide Fuel Fabrication Facility," p. 3, item 7. NRC: Washington, D.C. March 16, 1998. 

171 Donna Pouwakee, Manager, Waste Management, Nez Perce Indian Tribe, Consultancy with Council of Energy 

Resource Tribes, Lewiston, Idaho, 1992. 

172 "In 1995...nuclear power plant operators paid over 80 % of all user fees. Almost half of that amount went to pay 

for NRC overhead and management costs." Nuclear Energy Oversight, Nuclear Energy Institute, p. 3. May 1998. 

173Code of Federal Regulations, Title 10, Energy, Section 171.16, "Schedule of Materials Annual Fee Surcharges."

233 NUREG-1747
The original 1996 MOU between the Secretary of Energy and the Chairman of the NRC was 

signed in the expectation of a net benefit in safety.174 Just prior to that time, DOE had been the 

subject of audits by the General Accounting Office (GAO) and by the DNFSB which pointed to 

critical deficiencies in DOE safety programs and oversight. 

Also qualitatively, there may appear to be a clear and visible benefit from separation of the 

operator from the regulatory--just as the NRC was separated from the Atomic Energy 

Commission. To effect this separation while working with DOE, the NRC has established a 

clear separation of its activities from those of the DOE/RL TWRS program. 

Both DOE and NRC have reaffirmed their belief that "there would be clear benefits from 

external regulation of worker and nuclear safety at DOE facilities."17

" 

Preliminary NRC Staff Assessment 

The NRC staff notes that the likely, direct regulatory costs would be lower than the current 

costs (see Section D7 of this appendix) and the regulatory program and authority would be 

clearer, which would likely translate into lower, indirect costs. However, the total costs 

associated with regulation will probably be a small portion of the total program cost. 

Consequently, the NRC staff considers the NRC regulation of TWRS/WTP facilities to be a 

national policy issue that should not be decided by cost/benefit analyses. Therefore, the NRC 

staff does not consider this to be a significant issue. 

D.15 DEFENSE NUCLEAR FACILITIES SAFETY BOARD ROLE 

Issue 

The Defense Nuclear Facilities Safety Board (DNFSB) currently has an oversight role for the 

TWRS/WTP program. It is not clear that this would continue after transition to NRC regulation. 

Discussion 

The DNFSB was created in 1988 to provide independent, external oversight of DOE defense 

nuclear facilities.176 The specific functions of the DNFSB are to review and evaluate standards, 

conduct investigations, analyze design and operational data, review facility design and 

construction, and make recornmendations to the Secretary of Energy."7 The DNFSB has the 

174 U.S. Department of Energy and U.S. Nuclear Regulatory Commission Memorandum of Understanding, 

November 1996. 

175 Statements by "officials of both agencies" before a House Commerce Subcommittee, May 20, 1998. NUCLEAR 

WASTE NEWS, page 203, May 21, 1998. Elizabeth Moler's and Shirley Jackson's testimony (18 and 17 pp.) are obtainable through BPI DocuDial as Nos. 48-2946 respectively. 

176 Pub. L. 100-456. September 29, 1988 and United States Code, Section 2286. 

177 Atomic Energy Act, as amended, Section 312. 1954

NUREG-1747

234
authority to conduct hearings, establish reporting requirements for the Secretary of Energy, and 

assign resident inspectors at DOE defense nuclear facilities.178 

With regard to the DNFSB's scope of authority, a DOE defense nuclear facility means any of 

the following:" 

1. A production facility or utilization facility18, 1

81 that is under the control or jurisdiction of DOE 

and that is operated for national security purposes, but the term dose not include: 

a. Any facility or activity covered by Executive Order No. 12344, dated February 1, 1982, 

pertaining to the Naval nuclear propulsion program; 

b. Any facility or activity involved with the transportation of nuclear explosives or nuclear 

material; 

c. Any facility that does not conduct atomic energy defense activities; or 

d. Any facility owned by the United States Enrichment Corporation. 

2. A nuclear waste storage facility under the control or jurisdiction of the Secretary of Energy, 

but the term does not include a facility developed pursuant to the Nuclear Waste Policy act 

of 1982 and licensed by the NRC. 

Based on this definition, it is not clear whether the TWRS/WTP facility is a DOE defense 

nuclear facility. This definition does indicate that exceptions have been made for certain other 

facilities that appear to meet the basic criteria for being classified as a DOE defense nuclear 

facility but whose activities are overseen by agencies other than the DNFSB. 

If the TWRS-P facility is determined to be a DOE defense nuclear facility and subject to DNFSB 

oversight after NRC regulation begins, the TWRS/WTP contractor will be subjected to dual 

oversight. This could result in the NRC and the DNFSB applying overlapping resources on 

TWRS-P activities and expending resources to resolve any conflicting recommendations made 

178 Atomic Energy Act, as amended, Section 313. 1954 

179 Atomic Energy Act, as amended, Section 318. 1954 

180 Atomic Energy Act, as amended, Section 11, defines a production facility as (1) any equipment or device 

determined by rule of the Commission to be capable of the production of special nuclear material in such quantity as 

to be significant to the common defense and security, or in such manner as to affect the health and safety of the 

public; or (2) any important component part especially designed for such equipment or device as determined by the 

Commission. 

181 Atomic Energy Act, as amended, Section 11, defines a utilization facility as (1) any equipment or device, except 

an atomic weapon, determined by rule of the Commission to be capable to making use of special nuclear material in 

such quantity as to be of significance to the common defense and security, or in such manner as to affect the health 

and safety of the public, or peculiarly adapted for making use of atomic energy in such quantity as to be of 

significance to the common defense and security, or in such manner as to affect the health and safety of the public; 

or (2) any important component part especially designed for such equipment or device as determined by the 

Commission.

235 NUREG-1747
by the two organizations. Additionally, the TWRS/WTP contractor may have to employ 

additional resources to satisfy the needs of the two oversight bodies.1 82 

Preliminary NRC Staff Determination 

The NRC has determined that legislative authority is required for it to regulate TWRS/WTP 

facilities. As part of such legislation, the NRC staff anticipates that regulatory roles will be 

clearly defined, and based upon the precedent established at the enrichment and SNF storage facilities that the NRC regulates at DOE sites, the Staff does not expect a role for the DNFSB. 

The NRC staff does not consider this to be an issue. 

D16. PRICE-ANDERSON 11IDEMNIFICATION 

Issue 

Price-Anderson indemnification may need changes in its application if the TWRS/WTP facility 

transitions to NRC regulation. 

Discussion 

Under Subsection 170.d of the Atomic Energy act of 1954, as amended,"8 the Secretary of 

Energy is required to enter into indemnity agreements with anyone conducting activities under 

contract with DOE that involve risk of public liability and that are not subject to Subsection 

170.b., which addresses financial protection requirements for NRC licensees, or Subsection 

170.c., which specifies indemnification agreements for NRC licensees. 

In 1996, DOE issued a contract to BNFL Inc.'8 to conceptually design the TWRS-P tank waste 

treatment facility. This contract incorporates by reference the Department of Energy Acquisition Regulations (DEAR) clause on nuclear hazards indemnity, 185 which provides indemnification to 

BNFL Inc. pursuant to Subsection 170.d. of the AEA. This clause states, in part, "...the 

contractor will not be required to provide or maintain, and will not provide or maintain at 

Government expense, any form of financial protection to cover public liability...." The most 

recent contract negotiated with BNFL Inc.186 incorporates this same clause. The new, M&O 

type contracts are likely to contain the same clauses. 

182 DNFSB actions to date with respect to TWRS/WTP are consistent with the assumption that they interpret their 

scope of responsibility to include juri:sdiction over the program. 

183 42 USC 2210, The Atomic Energy Act of 1954, as amended, January 6, 1997. 

184 Department of Energy, (US)(DOIE). DE-AC06-96RL-1 3308, Amendment A0005. DOE: Richland, Washington. 

August 24, 1998. 

185 Department of Energy, (US)(DOIE). DEAR 952.250-70, "Nuclear Hazards Indemnity Agreement." DOE: 

Richland, Washington. January 1992. 

186 Department of Energy, (US)(DOCE). DE-AC06-96RL-13308, "Amendment A0005." DOE: Richland, Washington. August 24, 1998.

NUREG-1747

236
If the TWRS/WTP facility transitions to NRC regulation, the operator would most likely become 

the single licensee. In this case, the Secretary of Energy would no longer be required by the 

AEA to offer indemnification to the contractor, because the Contractor would be "...subject to 

financial protection requirements under Subsection b. or agreements of indemnification under 

Subsection c...."187 (It should be noted that, although Subsections 170.a., b., and c. of the AEA 

pertain to NRC licensees, it is not clear whether Subsection 170.d. would prohibit DOE from 

indemnifying an NRC licensee.)1" Additionally, if DOE were to no longer provide 

indemnification, the contract may need other clauses and changes (e.g., removal of the DEAR 

clause). 

With the TWRS-P contractor as an NRC licensee under 10 CFR Part 70, the NRC has optional 

regulatory authority under Subsection 170.a. of the AEA to require financial assurance, and if 

such financial assurance is required, indemnification may also be provided. Subsection 170.b. 

(1) requires that the amount of financial assurance be determined "...on the basis of criteria set 

forth in writing...." Subpart B to 10 CFR Part 140189 provides this criteria and contains NRC's 

requirements for financial protection and indemnity agreements for applicants and licensees 

other than Federal agencies and nonprofit educational institutions. 

Currently, 10 CFR Part 140, in addressing applicants for a license or licenses issued under 10 

CFR Part 70190 considers only plutonium processing facilities, fuel fabrication plants, and 

uranium enrichment facilities. Specifically, Subsections 140.13a and 140.13b address the 

financial protection required of those types of facilities. 

In applying discretionary authority for Price-Anderson indemnity coverage, the NRC has in the 

past not extended the coverage if the potential public liability from a possible nuclear accident 

would not exceed the amount of the commercially available insurance. Such financial 

assurance and indemnification have been required as license conditions for a small number of 

10 CFR Part 70 facilities, including plutonium processing and fuel fabrication facilities. 

Earlier this year, DOE and the NRC evaluated a number-of regulatory issues associated with 

NRC licensing a mixed oxide (MOX) fuel fabrication facility. The MOX facility would be located 

on a DOE site and designed, constructed, and operated by a private contractor. DOE would 

own the MOX facility; therefore, the NRC has stated its intention to make DOE and the 

operating contractor co-licensees. Regarding Price-Anderson indemnification for the MOX 

project, both DOE and NRC agreed that the operating contractor would be eligible for DOE 

indemnification if NRC did not extend its discretionary coverage, and hence, legislative 

clarification would not be required.1 91 However, the DOE-NRC evaluation did not discuss the 

187 Atomic Energy Act of 1954, as amended, Section 170.d. 1954 

188 Atomic Energy Act of 1954, as amended, Section 170. 1954 

189 Code of Federal Regulations, Title 10, Energy, Part 140, "Financial Protection Requirements and Indemnity 

Agreements." 

190 Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear Material." 

191 Section-98-050, "Development of Legislative Issues for Licensing of a MOX Fuel Fabrication Facility," March 16, 

1998.

237 NUREG-1747
fact that if DOE indemnifies the operating contractor (which is also a co-licensee), DOE may 

assess civil penalties against the contractor for violations of DOE Nuclear Safety 

Requirements,192 thereby establishing the potential for dual DOE-NRC regulation of the 

contractor. 

Preliminary NRC Staff Assessment 

The NRC staff believes the legislation enabling NRC regulation will define any Price-Anderson 

indemnification arrangements. The NRC staff does not consider this an issue. 

D17. TIMING TO AVOID DELAYS FROM REGULATORY TRANSITION 

Issue 

If regulatory transition were to occur, DOE would like the timing to offer the best chance of 

avoiding unnecessary delays, iin removing and immobilizing the tank wastes. 

Discussion 

From the perspective of DOE, the timing for transition needs to be selected so as to reduce the 

potential for disruption to a minimum. Some timing considerations tend to conflict. In 

programmatic terms, it would be beneficial for DOE that transition occurs sometime after full 

and confirmed operation so that it could be assured that the plant meets its specifications. On 

the other hand, to minimize potential retrofit changes in meeting NRC's regulatory expectations, it might be preferable to make the transition as soon as possible. (This can also be 

accomplished by resolving aUl regulatory differences prior to construction and transitioning 

later.) 

Legislation may be required if the 10 CFR Part 2 public hearings process is not to be invoked, 

perhaps by NRC's acceptance of DOE's prior authorization basis, or if there is intent to certify, rather than license, the facility. Legislation or Office of Management and Budget action may be 

needed on funding issues. DOE is of the opinion that potential delays involved in legislation 

and NRC review can best be accommodated in a process conducted in parallel with the 

DOE/RU authorization process during the TWRS-P design and construction activity. 

As part of the MOU, ongoing cooperation has existed between DOE and NRC in identifying the 

full spectrum of differences in regulatory oversight between the two agencies. This work is 

intended to resolve technical differences in such a manner that the design will meet NRC 

requirements or that NRC will accept DOE's prior authorization basis. 

DOE can authorize the design through construction on the basis of the PSAR. At that point, the 

RU intends a second licensing step to authorize operation after review and approval of a Final 

Safety Analysis Report (FSAR). This junction, normally involving revised documentation 

between the PSAR and FSAR, could beneficially be the time at which the documentation and 

the Safety Analysis Report (SAR) are upgraded and changed to meet NRC requirements for 

192 Code of Federal Regulations, Title 10, Energy, Section 820.20(b), "Basis for Civil Penalties."

NUREG-1747 238
format and content. With the forgoing cooperative work in hand, the disturbance could be 

minimal. 

In order to have confidence that the design of the vitrification plan will operate as specified and 

that planning towards remediating the tank farm(s) will be successful, DOE must have 

confidence that the plant is constructed as planned. This includes actual design and 

construction. It could also be desirable, from DOE's perspective, to confirm operation and the 

end products of each section of the plant, including the pretreated wastes and the quality of the 

glasses. 

The NRC staff believes that there are three alternatives for the timing of regulatory transition: 

1) immediately, 2) overlapping, or parallel, regulatory transition process during design, or 3) 

regulatory transition during operations. Alternatives one and two correspond to licensing of the 

vitrification plant, for which rules and guidance already exist. Option 3 represents a certification 

route. 

Preliminary NRC Staff Assessment 

The NRC staff believes the legislation establishing NRC regulation of TWRS/WTP will define 

the timing for transition. Transition as soon as possible in the near-term offers the clearest 

benefits. 

D18. DOE FINANCIAL INTERESTS 

Issue 

DOE has expressed concerns about protecting its financial interests in the vitrification facility. 

Discussion 

DOE has already invested several hundred million dollars in the Phase I activities. The 

completion of the design and construction activities in Phase I may amount to another 

$4 billion 1

". Thus, DOE has considerable financial interests in the program. DOE protects those interests, particularly through frequent meetings with the TWRS contractors to assess 

progress and by an inspection program during preliminary design and construction. Ordinarily, 

NRC would not become involved with the regulatory process until after completion of 

preliminary design. The NRC has no financial interests to protect. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this a regulatory transition issue because the NRC's mission 

is protection of the workers, the public, and the environment. 

193 Nuclear News Flash, September 1, 2000.

239 NUREG-1747
D19. QUALITY ASSURANCE

Issue 

The DOE TWRS Quality Assurance (QA) program requirements and the TWRS-P contractor's 

DOE-approved QA programs may need to be modified in order to facilitate the transition to 

NRC regulation. 

Discussion 

The QA of TWRS/WTP activities is governed by the DOE QA rule (10 CFR 830.120).194 

Although organized and worded differently, the DOE QA rule requirements are intended to be 

generally consistent with basic commercial nuclear industry QA requirements (such as 

NQA-1 195) and NRC QA requirements.196.197 The DOE QA rule requirements are expressed at a 

higher level than commercial nuclear industry and NRC QA requirements, and provide for 

flexibility in the approaches that may be taken to meet the requirements. As a result, a QA 

program could be developed to meet the requirements of the DOE QA rule without meeting the 

NRC QA requirements. DOE: has prepared an implementation guide for the QA rule, 98 that 

provides additional insights into what DOE considers an acceptable QA program. 

The DOE QA rule does contain requirements that go beyond what is normally expected of an 

NQA-1 or NRC QA program. However, the preamble to the DOE QA rule states, "Just as the 

NRC has endorsed NQA-1 as an acceptable way (there are others) for their licensees to 

implement the requirements of (10 CFR Part 50) Appendix B, DOE contractors may use NQA-1 

as a way to implement the rule."199 

In order to determine the impact of transitioning from a QA program that meets the DOE QA rule to a program that meets NRC QA requirements, the NRC QA requirements must be 

known. At this time, the NRC has defined the QA requirements it would apply to a high level 

194 Item 3 of Section 6.0 of the "Policy for Radiological, Nuclear, and Process Safety Regulation of TWRS 

Privatization Contractors," (DOE/RL-96-25) states that the contractor's set of standards and requirements "shall 

contain, as a subset, the nuclear safety requirements in 10 CFR Parts 830, 834, and 835 that are enforceable under 

10 CFR Part 820." 

195 ANSII"ASME NQA-1, Quality Assurance Requirements for Nuclear Facilities. 

196 The NRC QA requirements being referred to are thos in 10 CFR Part 50, Appendix B, Quality Assurance Criteria 

for Nuclear Power Plants and Fuel Rieprocessing Plants; 10 CFR Part 60, Subpart G, Quality Assurance; 10 CFR 

Part 71, Subpart H, Quality Assurance; and 10 CFR Part 72, Subpart G, Quality Assurance. 

197 See Section 3.5 of the DOE/EM's Comparison of NRC and DOE Requiremnts for Independent Spent Nuclear 

Fuel Storage Facilities dated June 1995 and section 4.1.6 of the DOE/EM's Office of Spent Fuel Management Comparison of Department of Energy TWRS Top-Level Standards and Principles and Nuclear Egulatory Commission Draft Standard Review Plan for Fuel Cycle Facilities. 

198 Implementation Guide for US with 10 CFR 830.120, Quality Assurance, DOE G 830.120. 

199 Department of Energy (U.S.), Washington, D.C. Federal Register Vol. 59, pp. 15843-15849 (Sec I1.C.30). 

April 5, 1994.

NUREG-1747 240
waste (HLW) processing facility. The existing 10 CFR Part 70m does not contain specific QA 

requirements for SNM licensees except for licensees who intend to possess and use SNM in a 

plutonium processing and fuel fabrication plant. For such a plant, 10 CFR Part 70 indicates that 

the QA program description should include a discussion of how the criteria in 10 CFR Part 50, 

Appendix B will be met.-°1 

Specific references to 10 CFR Part 50, Appendix B or NQA-1 are not anticipated to be included. 

In concert with development of the new draft 10 CFR Part 70, NRC has developed two SRPs; 

some for facilities possessing and using SNM and some specifically for TWRS/WTP facilities. 

The current SRP for TWRS-P states that the NRC reviewer should review the application to 

determine whether the applicant, for items relied on to prevent or mitigate accidents of high 

consequences, has either: 

1. Described its (and its principal contractor's) organization, organizational authority, and 

organizational responsibilities for QA; provided a commitment to implement and maintain 

the QA program to comply with the applicable elements of NQA-1; or 

2. Described its (and its principal contractor's) organization, organizational authority, and 

organizational responsibilities for QA, and addressed the extensive checklist provided in 

the appendix to this section of the SRP. 

In either case, the applicant should also (a) describe how the QA program will be graded 

for items of lesser or no effect on consequences of concern and (b) list the items relied on 

for safety as determined by the applicant's integrated safety analysis. 

In its QA Program and Implementation Plan (QAPI P), 202 which had been approved by the RU, 203 

BNFL Inc. had committed to comply with the applicable elements of NQA-1, DOE/RW-0333P, 204 

and NUREG-1293.2o0 Additionally, the TWRS-P contract required that DOEIRW-0333P be 

applied.2'

0 Although the QAPIP appeared to indicate that the basic requirements of NQA-1, 

Part I (1 994a) apply and that the NQA-1 supplementary requirements were to be addressed as 

200 Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Licensing of Special Nuclear Material." 

201 Code of Federal Regulations, Title 10, Energy, Draft Part 70, "Domestic Licensing of Special Nuclear Material." 

July 30, 1998. 

202 BNFL Inc., BNFL-5193-QAP-01, Revision 4, "Quality Assurance Program and Implementation Plan." BNFL Inc.: 

Richland, Washington. 1998. 

203 Gibbs, D.C., U.S. Department of Energy, letter to M.J. Bullock, BNFL, Inc., "Approval of BNFL-5193-QAP-01, 

Rev. 4, 'Quality Assurance Program and Implementation Plan,' (QAPIP)," June 2, 1998. 

204 Department of Energy (U.S.)(DOE). DOE/RW-0333P, "Quality Assurance Requirements and Description for the 

Civilian Radioactive Waste Management Program." DOE: Richland, Washington 

205 Nuclear Regulatory Commission, (U.S.)(NRC). NUREG-1293, "Quality Assurance Guidance for a Low-Level 

Radioactive Waste Disposal facility." NRC: Washington, D.C. 1991. 

206 Department of Energy (U.S.)(DOE). RLIREG-98-13, Revision 0, Table 3.1.1 and Section 4.1, "DOE Regulatory 

Unit Evaluation Report of the BNFL Inc. Quality Assurance Program and Implementation Plan." DOE: Richland, 

Washington. May 1998. indicates that application of DOE/RW-0333P is a contractual requirement.

241 NUREG-1747
required by project specific activities, the language in the QAPIP was not sufficiently clear for 

the NRC to make a definitive judgment on which elements of NQA-1 had been committed to. 

However, the QAPIP is clear that DOE/RW-0333P applies. DOW/RW-0333P requirements are 

derived to a large extent from NRC QA requirements (10 CFR Part 50, Appendix B; 10 CFR 

Part 60, Subpart G; 10 CFR Part 71, Subpart H; and 10 CFR Part 72, Subpart G) and NQA-1 

requirements. DOEIRW-0333P has been approved by the NRC under 10 CFR Part 60 and 10 

CFR Part 71 for radioactive material packages. 207 Moreover, DOE is applying DOE/RW-0333P 

to the TIM-2 Independent Spent Fuel Storage Installation, which is in the process of being 

licensed by the NRC.2 

Based on the TWRS-P commitment to apply DOE/RW-0333P, the NRC concluded that the 

TWRS-P QA program description satisfied the QA requirements of the 1998 draft of 10 CFR 

Part 70 and the guidance in the corresponding draft of the TWRS SRP regarding a commitment 

to meet NQA-1 (the Basics and Supplementary Requirements in Part I [from the former NQA-1] and the requirements in Part II [from the former NQA-2] but not the Nonmandatory Appendices 

Preliminary in Part Ill). 

Preliminary NRC Staff Assessment 

The DOE TWRS Quality Assurance (QA) program requirements and the TWRS-P contractor's 

DOE-approved QA programs may need to be modified in order to facilitate the transition to 

NRC regulation. The NRC staff believes the differences in QA requirements are small and the 

adequacy of the QA program really depends upon its implementation. The NRC Staff does not 

consider this a significant transition issue. 

D20. IMPACTS ON OTHER HANFORD SITE FACILITIES 

Issue 

DOE is concerned that there may be impacts upon non-TWRS/WTP facilities at Hanford after 

regulatory transition to the NRC. 

Discussion 

Some NRC requirements are known to be different from DOE standards and practices 

established for the design and safe operation of facilities at Hanford. Examples include, risk 

evaluation guidelines including the concept of the co-located worker, design basis seismic and 

207 Travers, W.D., U.S. Nuclear Regulatory Commission, letter to R.A. Milner, U.S. Department of Energy, February 

29, 1996. 

208 License Application for the Idaho National Engineering Laboratory Three Mile Island Unit Two Independent Spent Fuel Storage Installation, Chapter 6, Revision 0, October 1996.

NUREG-1 747

242
tornado magnitudes, and natural phenomena hazards evaluation criteria.2° 0

9,21 However, the 

NRC may find such DOE standards and practices acceptable, even if different. As a recent 

example, NRC has accepted the DOE's approach for determining the design basis earthquake 

and tornado design requirements for the new dry spent fuel storage facility at the INEEL, which 

is undergoing a licensing review under 10 CFR Part 72.211 

Where DOE's approach is not acceptable to the NRC, the NRC will require TWRS/WTP to 

comply with its different requirements, after transition to regulation of the facility by NRC. 

Although the NRC would have no authority to impose these requirements on non-TWRS/WTP 

facilities, the imposition of NRC requirements on TWRS/WTP could impact on other facilities 

broadly in two ways. First, designing or operating the facility to meet NRC requirements may require design and operational changes at neighboring facilities or from the Hanford site as a 

whole. Second, in cases where the facility is designed or operated to meet certain NRC 

requirements that are different than DOE standards and practices, obviously different safety requirements between TWRS/WTP and its neighboring facilities may be questioned by 

stakeholders, and pressure for upgrading non-TWRS/WTP facilities may result. 

An example of the first type of potential impact involves NRC's interpretation that a co-located 

worker at a DOE site will be considered a member of the public for accident dose evaluations 

under NRC regulation.212 If the TWRS/WTP facility is designed so that any individual outside of 

its controlled area of a few hundred yards meets NRC's public accident dose limits, then there 

is no effect on neighboring facilities. However, if the facility is not so designed, the NRC could 

expect the contractor to exercise some control over Hanford worker access to surrounding 

facilities and to include such workers in the TWRS/WTP facility radiation protection and 

monitoring program. 213 Also, the NRC could expect the DOE Richland Operations Office and 

local agencies to redefine the site's established emergency planning zones and evaluation 

practices, and to revise emergency planning agreements. 

Whether the second type of impact occurs is unclear, as evidenced by three recent examples. 

A new TMI-2 spent fuel storage facility at INEEL will be licensed by the NRC and located within 

the security fence of the DOE-regulated Idaho Chemical Processing Plant (ICPP). New N 

reactor spent fuel conditioning and storage facilities at Hanford are being designed to NRC

equivalent requirements while modifications to existing, nearby project facilities are 

accomplished in accordance with DOE requirements. For both projects, public support for the 

move to NRC requirements has been positive, and there has been no public concern expressed 

209 "Comparison of DOE and NRC Requirements for Spent Nuclear Fuel Storage Facilities." SCIENTECH. 1995. 

210 Westinghouse Hanford Company (WHC). WHC/DB-003, Additional NRC Requirements for the K Basins Spent 

Fuel Project. WHC: Richland, Washington 

?11 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-071, "Exemption to 10 CFR 72.102(f)(1) Seismic 

Design Requirements for Three Mile Island Unit 2 Independent Spent Fuel Storage Installation." NRC: Washington, 

D.C. April 8, 1998. 

212 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-038, "Hanford Tank Waste Remediation System 

Privatization Co-Located Worker Standards." NRC: Washington, D.C. March 4, 1998. 

213 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-038, page 3, "Hanford Tank Waste Remediation 

System PRIVATIZATION Co-Located Worker Standards." NRC: Washington, D.C. March 4, 1998.

243 NUREG-1747
over the safety of the neighboring DOE regulated facilities. The DOE project managers attribute these positive results to frequent and open communications with their Stakeholders. 

On the other hand, at Hanford, at least one Stakeholder representing a public interest group has already expressed an interest in formally participating in any NRC licensing of the 

vitrification plant. 

The technical impacts to non-TWRS/WTP facilities of regulatory transition need to be identified, 

evaluated, and addressed on an individual basis as such issues arise. Coordinating the 

transition to external regulation of the tank farm storage and vitrification facilities will likely reduce the impacts to the non-vitrification portion of the tank farms, which is the non

TWRS/WTP facility most susceptible to these impacts. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this an issue for regulatory transition. 

D21. RESPONSIBILITY FOR OCCUPATIONAL SAFETY 

Issue 

For its facilities, DOE has regulatory authority for occupational safety. For NRC licensed 

facilities, the OSHA has such authority. If the TWRS/WTP transitioned to NRC regulatory oversight, OSHA might assume regulatory authority for occupational safety. 

Discussion: 

The Occupational Safety and Health Act prohibits OSHA from regulating where another Federal 

agency chooses to exercise its own authority.214 DOE has authority for regulating worker safety 

at its facilities under the Atomic Energy Act of 1954.21 Accordingly, DOE contractors are 

responsible for worker occupational health and safety;2

"6 DOE enforces compliance with both 

OSHA and DOE requirements for worker occupational health and safety. 

Traditionally, DOE's authority has extended to all facilities at a site, even new facilities which 

are typically government-owned, contractor-operated (GOCO) facilities. The original TWRS-P 

facility would have been company-owned, company-operated (COCO). However, OSHA 

declined to accept regulatory authority for the facility, and consequently, the RU regulates the 

occupational safety program. In 1993, the Secretary of Energy committed to replace self

regulation of worker safety with external regulation, and OSHA was undertaking pilot programs 

214 OSHA Act of 1970, Section 4(b)(l). 1970. 

215 Atomic Energy Act of 1954, Section 161.i(3). 1954 

216 Department of Energy (U.S.)(DOE). Order 440.1A, "Worker Protection Management for DOE Federal and 

Cotnractor Employees." DOE: Washington, D.C. March 27, 1998.

NUREG-1747 244
at certain DOE facilities,21

" although it is not clear when, or if, OSHA would accept and exercise 

oversight authority over DOE facilities. 

If the TWRS/WTP project were to transition to regulation by NRC, the 1988 MOU between NRC 

and OSHA218 could apply. (it would also be possible to develop a specific MOU to address the 

TWRS/WTP project, like the MOU written for the GDPs.219) Under the 1988 MOU, NRC 

exercises regulatory responsibility for worker radiological health and safety at an NRC-licensed 

facility. Furthermore, as stated in the 1988 MOU, NRC would have the responsibility for 

regulating the "...chemical risk produced by radioactive materials" and "...plant conditions that 

affect the safety of radiological materials and thus present an increased radiation risk to 

workers." These latter responsibilities include regulating certain aspects of fire and chemical 

safety at NRC facilities. In addition, under the 1988 MOU, during inspections of radiological 

and nuclear safety, NRC personnel may identify occupational safety concerns, normally within 

the scope of OSHA's responsibility, and may receive complaints from an employee about 

OSHA-covered working conditions. In such instances, NRC would report the matter to the 

licensee's management. NRC inspectors may elevate OSHA safety concerns to the attention 

of NRC management who would inform the OSHA Regional Office, when appropriate. On a 

case-by-case basis, NRC and OSHA conduct joint inspections at NRC-licensed facilities. 

Conversely, OSHA provides NRC with information that it may obtain concerning worker 

radiological health and safety. 

Preliminary NRC Staff Assessment 

The NRC staff anticipates that the enabling legislation for NRC regulation would clearly define 

the responsibilities. This is not a significant issue for regulatory transition. 

D22. TAILORING OF REQUIREMENTS 

Issue 

The NRC and DOE use different approaches for the tailoring of requirements for safe mission 

performance. 

Discussion 

This issue concerns the different approaches used by DOE and NRC for establishing 

requirements for achieving adequate safety and reconciliation of any differences in the 

approaches to facilitate seamless transition from DOE to NRC regulation, should transition 

occur at a future date. As discussed in more detail below, even though the DOE approach to 

establishing requirements is somewhat different than the NRC approach, the requirements 

217 Voluntary Protection Programs are in place at WIPP, Allied Signal, both Star sites, and Weldon Springs (Merit 

site). OSHA trials are currently being conducted at ORNL and ANL. 

218 Nuclear Regulatory Commission (U.S.), Washington, D.C. "Memorandum of Understanding between the NRC 

and OSHA; Worker Protection at NRC-Licensed Facilities." Federal Register. Vol. 53, p. 43950. October 31, 1988. 

219 Nuclear Regulatory Commission (U.S.), Washington, D.C. "MOU between the NRC and OSHA with Respect to 

the Gaseous Diffusion Plants." Federal Register. Vol. 61, p. 40249. August 1, 1996.

245 NUREG-1 747
resulting from either approach will likely achieve adequate safety, and significant reconciliation 

may not be necessary. 

The DOE approach is a bottom-up process based on the principles of Integrated Safety Management. DOE contractors identify, propose, and justify safety standards and 

requirements by selecting frorn appropriate DOE and industry standards and requirements, with 

full consideration of the work processes and specific hazards associated with a facility. In the 

case of the TWRS project, the DOE approach for radiological, nuclear, and process safety° places on the TWRS Contraclor the responsibility to achieve (a) adequate safety of the workers 

and the public, (b) comply with applicable laws and legal requirements, and (c) conform to DOE 

stipulated top-level safety standards and principles. 2

" Within this framework, the contractor 

identifies the work needed, evaluates the associated hazards, selects appropriate standards, 

and justifies the adequacy of the selected standards.m 

The traditional NRC approach is a top-down process. Adequate protection for the general public, workers, and the environment with regard to matters under the authority of NRC is 

achieved through compliance with established requirements, in the form of rules, regulations, and orders embodied primarily in 10 CFR Energy. The NRC's requirements have evolved over 

the years through the formal rulemaking process and are continually subject to modification and 

change as new information becomes available. Such requirements are established for classes 

and types of facilities and processes. License applicants are required to demonstrate 

compliance with all applicable requirements before a license is granted by the NRC. When a 

license is granted following NRC's detailed reviews of the applicant's information and analyses, the NRC concludes that adequate safety will be achieved because applicable rules, regulations, and orders will be met.2 3 

The revised 10 CFR Part 70, which the NRC would likely apply to TWRS/WTP facilities, is more 

relevant.22 4 The revised 10 CFR Part 70 contains top-level performance-based requirements and baseline design criteria, rather than NRC's traditional prescriptive, deterministic 

220 Department of Energy (U.S.)(DOE). RL/REG-97-1 0, Rev. 1, "Radiological, Nuclear, and Process Safety 

Regulation of TWRS Privatization Contractors Regulatory Plan." DOE: Richland, Washington. January 1998. 

221 Department of Energy (U.S.)(DOE). DOE/RL-96-0006, Rev. 0, "Top-Level Radiological, Nuclear, and Process 

Safety Standards and Principles for the Privatization Contractor." DOE: Richland, Washington. February 1996. 

= Department of Energy (U.S.)(DOE). DOE/RL-96-0004, Rev. 0, "Process for Establishing a Set of Radiological, Nuclear, and Process Safety Standards and Requirements for TWRS Privatization." DOE: Richland, Washington. 

February 1996. 

223 For example, see Code of Federal Regulations, Title 10, Energy, Sections 70.23(a), 70.31(a) and 70.32(a)(8) on license approval and issuance. 

224 NRC statement during a May 28, 1998 public meeting on the revised Code of Federal Regulations, Title 10, 

Energy, Part 70.

NUREG-1747

246
requirements.3m As noted by NRC, under the revised Part 70, "the contractor must design 26 its 

facility to meet the baseline design criteria and other requirements that are identified in the 

NRC's regulations and perform an integrated safety analysis of the facility design to confirm 

that appropriate controls will be provided.... This approach incorporates both flexibility of a risk

informed, performance-based safety analysis as well as baseline design criteria from which a 

judgment is reached that there is reasonable assurance of adequate safety." The applicant's 

license application summarizes the results of the integrated safety analysis, the facility design, 

compliance with the prescribed standards and requirements, and the ability of the facility to 

withstand/mitigate the hazards identified in the integrated safety analysis. 

Therefore, there is nothing inherent in the TWRS/WTP requirements development process, 

compared to the draft Part 70 process, that would preclude the selection of a set of standards 

and requirements for the TWRS/WTP facility that will achieve adequate safety in either an NRC 

or a DOE regulatory environment. However, this will not guarantee that the NRC would accept the TWRS/WTP standards and requirements as sufficient should the facility transition to NRC 

regulation. In fact, NRC has already acknowledged that the DOE and NRC approaches are 

intended to reach the same endpoint but has expressed the following concerns with the 

approach for TWRS-P: 

1. Safety requirements were proposed and conditionally approved during the pre-conceptual 

design phase of the facility. This attempt to reach judgment on adequate safety was 

performed prior to the majority of design information being available. 

2. For both regulatory processes, the determination of adequate safety is somewhat subject 

to engineering judgment and, as such, should transition occur, additional requirements and 

potential backfits could be needed. 

3. The situation is exacerbated by (a) DOE's tailoring approach, which allows the TWRS-P 

Contractor to pick and choose in proposing a set of applicable safety requirements, and (b) 

the appearance that the RU is increasingly influenced by cost and schedule considerations 

during its regulatory decision making process (from Chapter 7.0, Main Reference 17). 

Preliminary NRC Staff Assessment 

The NRC staff believes this issue is not significant if regulatory transition occurs in the near

term because the end result should be comparable. 

225 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-185, "Proposed Rulemaking-Revised Requirements for 

the Domestic Licensing of Special Nuclear Material." NRC: Washington, D.C. July 30, 1998. 

226 Nuclear Regulatory Commission (U.S.)(NRC). SECY-98-185, "Proposed Rulemaking-Revised Requirements for 

the Domestic Licensing of Special Nuclear Material." NRC: Washington, D.C. July 30, 1998. Note that the wording 

on pages 23 and 44 of SECY-98-185 indicates that an applicant must consider/address the baseline design criteria, 

not necessarily meet them.

247 NUREG-1747
D23. ACHIEVEMENT OF ADEQUATE SAFETY

Issue 

DOE and NRC may reach different conclusions about the achievement of adequate safety and 

this may require reconciliation during the possible future transition of TWRS/WTP to NRC 

regulation. 

Discussion 

This concerns the implementation of measures to avoid or to identify, assess, and disposition differing conclusions between DOE and NRC on the achievement of adequate safety as a result 

of the different approaches. Such measures are best implemented early in the project's life cycle and continue on an ongoing basis to preclude a backlog of differences that would have to 

be resolved as part of the transition process to NRC regulation. 

The DOE regulatory approach for TWRS-P safety (radiological, nuclear, and process) places 

on the TWRS/WTP Contractor the responsibility to achieve (a) adequate safety for the workers 

and the public (b) comply with, applicable laws and legal requirements, and (c) conform to DOE 

stipulated top-level safety standards and principles. 2

"

7 Within this regulatory framework, the TWRS-P Contractor identifies the work needed, evaluates the associated hazards, selects 

appropriate standards, and justifies the adequacy of the selected standards.m As discussed in Section D.22, the DOE regulatory approach for TWRS-P is potentially similar to the NRC approach anticipated in the revised 10 CFR Part 70. Nevertheless, the NRC has expressed the 

following concerns about the! process for TWRS-P: 

1. Safety requirements were proposed and conditionally approved during the pre-conceptual design phase of the facility. This attempt to reach judgment on adequate safety was 

performed prior to the majority of design information being available. 

2. For both regulatory processes, the determination of adequate safety is somewhat subject to engineering judgment, and as such, should transition occur, additional requirements and 

potential backfits could be needed. 

Preliminary NRC Staff Assessment 

The NRC staff believe this is a technical issue that requires additional design effort by the 

contractors. 

227 Department of Energy (U.S.)(DOE). RL/REG-97-10, Rev. 1, "Radiological, Nuclear, and Process Safety Regulation of TWRS Privatization Contractors Regulatory Plan." DOE: Richland, Washington. January 1998. 

228 Department of Energy (U.S.)(DOE). DOE/RL-96-0004, Rev. 0, "Process for Establishing a Set of Radiological, Nuclear, and Process Safety Standards and Requirements for TWRS Privatization." DOE: Richland, Washington. 

February 1996.

NUREG-1747 248
D24. EFFECT OF SCHEDULE DELAYS ON THE OVERALL RISK

Issue 

DOE is concerned about balancing the risks of the TWRS/WTP facility with the risk of 

continued storage of the tank wastes. 

Discussion 

DOE has expressed interest in balancing the risks posed by continued operation of the tanks 

with the risks posed by operation of the vitrification plant, regardless of the regulatory 

framework. In particular, DOE is concerned that delays in TWRS/WTP operations due to 

regulatory matters (DOE and/or the NRC transition) would translate into increased risks 

because the tank farm system is old and potentially leaking. In theory, the regulatory 

requirements for a TWRS/WTP facility could be different (and reduced) against a baseline of 

the existing hazards of tank storage, which could be reduced by the operation of the facility. From the viewpoint of DOE, early startup, high availability, and high capacity operation of the 

vitrification plant could serve to reduce the risks posed by the tanks and the overall risk of HLW 

at Hanford, even if requirements on the vitrification plant were relaxed. 

NRC requirements have been established to minimize risks of operation of a facility relative to a 

baseline of non-operation, even though NRC can consider balancing risks within a facility. However, NRC operates to codified regulations (10 CFR Part 70) and, while provision for risk 

benefits and balancing can be made via the risk-informed, performance based approach, it can 

be difficult to manage and balance the different risks on a quantitative basis with limited 

information on the design and safety features of the TWRS/WTP facility. Regulatory 

requirements could be administered with recognition of the real risks from the degrading tanks 

as baseline risks. 

The risks to continued operation of the tank farm are analyzed in the Basis for Interim 

Operation (BIO),9 which temporarily substitutes for a comprehensive Safety Analysis Report 

(SAR). The tanks also represent immediate concerns. Steel tanks that were built upwards of 

55 years ago to a design life of 10 years ultimately will fail. The collapse of tanks due to 

overburdened loads, perhaps under seismic excitation, while they still contain waste represents both an immediate public concern and a long-term remediation challenge. 

Preliminary NRC Staff Assessment 

The NRC staff anticipates that the scope of NRC regulation-the TWRS/WTP facilities, tank 

farms, other Hanford facilities-will be defined in the enabling legislation. A single regulatory 

entity offers consistency and the potential to consider total risks from tank waste management. 

However, absent specific design and quantitative analysis, and NRC regulation of the entire 

Hanford Site operations, it is not possible to assess balancing total risk for the tank waste/HLW 

M Department of Energy (U.S.)(DOE). HNF-SD-WM-BIO-001, Rev. 0, Executive Summary page ES-1 3, 'The 

Basis for Interim Opration of the Tank Waste Remediation System Facilities." DOE, Richland, Washington. August 

1998.

249 NUREG-1747
related systems at Hanford. However, achieving adequate assurances of safety is likely to be 

based upon individual facility analyses. 

D25. RESOURCE ALLOCATION ACROSS TWRS FACILIES 

Issue 

DOE has expressed concerns about balancing the resources available for the safety features of 

the TWRS/WTP (the vitrification facilities in particular) across the entire TWRS complex. 

Discussion 

The resources that support safety operation of the vitrification plan and the tanks are both 

drawn from the same source (i.e, public funds) and are, therefore, ultimately in competition with 

one another. DOE's concern is that the risk of the deteriorating tanks is increasing as the tanks 

continue to age thereby increasing the likelihood of tank leaks and accidents. Approximately 25 

percent of the tank farm equipment fails upon demand. Therefore, this issue is directed toward 

the DOE observation that regulatory oversight needs to be applied to the vitrification plant and 

the tanks in such a manner as to minimize the overall risk from both. This could be done by providing a consistent regulatory framework for all parts of the TWRS/WTP program-the tanks 

and the vitrification plant. 

Given the increasing risk of continuing the status quo with the tanks, DOE wants to emphasize the importance that all efforts be made to ensure that the design, licensing, and operation of the 

vitrification plant occur on schedule. However, while such a conclusion may be logical, DOE 

does not want to be seen as recommending regulation of the vitrification plant to any lesser 

standards than are applied to their other nuclear facilities. In management of the Hanford site, DOE ensures the safety of all operations according to the risk of those operations, and 

allocates funding and resources accordingly. From DOE's perspective of total risk, this might correspond to expedited schedules or safety features different from a facility constructed 

individually. 

Preliminary NRC Staff Assessment 

The NRC is concerned with adequate assurances of safety for the workers, the public, and the 

environment. Once safety and regulatory requirements are met for TWRS/WTP, resource 

allocation is an internal DOE matter. The NRC staff expects the scope of NRC regulation to be 

defined in the enabling legislation. 

D26. BACKFITIRETROFIT 

Issue 

Changes may be required as the TWRS/WTP facility transitions to NRC regulation.

NUREG-1747 250
Discussion

The TWRS-P facility could be subjected to new or modified requirements imposed by NRC if 

regulatory transition occurs. These new or modified requirements may require that the 

TWRS-P facility be modified. Terms such as "backfit" and uretrofit" are commonly used to 

describe modifications that are made in response to a new or modified requirements. These 

terms, the regulatory actions and precedents associated with them and the differences between 

them are discussed further below. 

The NRC has backfitting provisions in its regulations for reactors, independent spent fuel 

storage installations, and GDPs. 230 In these regulations, backfitting is generally defined as the 

modification of, or addition to, systems, structures, or components of a facility; or to the 

procedures or organization required to operate a facility, any of which may result from a new or 

amended provision in the NRC rules or the imposition of a regulatory staff position interpreting 

the NRC rules that is either new or different from a pervious NRC staff position. Backfitting only 

applies to facilities that have received a license (including a construction permit) or a 

certification. For these facilities (except for a few special conditions noted in the regulations21 ), 

the NRC can require a backfit only when it determines that there is a substantial increase in the 

overall protection of the public health and safety or the common defense and security to be 

derived from the backfit and that the direct and indirect costs of implementation for that facility 

are justified in view of this increased protection. 

Backfit provisions are expected to be ultimately included in 10 CFR Part 70.2 In response to 

an Nuclear Energy Institute (NEI) Petition for Rulemaking that requested a backfitting provision 

for 10 CFR Part. 70 similar to the one in 10 CFR 50.109, the NRC staff proposed that "a 

qualitative backfit mechanism, similar in purpose to a 10 CFR 50.109 provision, be considered 

after the safety bases, including the results of the ISA, are established and incorporated in the 

license, and after licensees and staff have gained a few years of experience with 

implementation of the ISA requirement. This mechanism would not apply to modifications 

identified as a result of the initial ISAs that are needed to assure protection of public health and 

safety; these modifications would be required for compliance with the revised 10 CFR Part 

70.233 

23o See Code of Federal Regulations, Title 10, Energy, Section 50.109, "Backfitting," Section 72.62, "Backfitting," 

and Section 76.76, "Backfitting." 

231 A backfit analysis is not required if regulatory action is necessary to ensure that a facility provides adequate 

protection to the health and safety of the public; the modification is necessary to bring a facility into compliance with 

a license/certification or NRC rules or orders, or into conformance with written commitments by the licensee; or the 

regulatory action involves defining or redefining what level of protection to the public health and safety should be 

regarded as adequate. 

232 Code of Federal Regulations, Title 10, Energy, Part 70, "Domestic Ucensing of Special Nuclear Material." The 

draft 10 CFR Part 70 issued on July 30, 1998, does not include a backfit provision. However, because of the 

differences of opinion on whether to include a backfit provision in Part 70, the Commission is requesting public 

comment on its intent to defer consideration of a qualitative backfit provision in Part 70. 

233 Nuclear Regulatory Commission (U.S.)(NRC). SECY-97-137, Attachment 1, Section 8.0, "Proposed Resolution 

to Petition for Rulemaking Filed by the Nuclear Energy Institute." NRC: Washington, D.C. June 30, 1997.

251 NUREG-1747
Changes that are required to be made when an existing facility applies for an NRC license or 

certification are termed "retrofits" and do not meet the criteria for being classified as a backfit. 

Thus, NRC backfit provisions do not apply to retrofits. In fact, the NRC does not have specific retrofit provisions in its regulations. The Portsmouth and Paducah GDPs, which were existing, DOE-regulated operational facilities prior to being transitioned to NRC regulation, had to meet 

or commit to meet the applicable NRC regulations in order to receive certifications. For these 

plants a plan for achieving compliance with respect to any areas of noncompliance with the 

NRC's regulations had to be submitted with the application for an initial certificate of 

compliance.2 " Although a backfitting rule for the GDPs was issued, the NRC interpreted the 

rule not to apply until after the initial NRC certification process -was completed. 235 It should be 

noted that the GDPs were trarnsitioned to NRC regulation approximately 50 years after their 

design and 45 years after construction, while transition of the TWRS/WTP facilities would likely 

occur within 5 years of the design activities. 

Preliminary NRC Staff Assessment 

The NRC anticipates that some Backfits/Retrofits may be necessary upon transition of the 

TWRS/WTP to NRC regulatory authority. However, the NRC staff expects these to be minor 

because the facilities would represent new design and construction, with complete documentation, and the NRC has participated with DOE and their contractors in Phase IA and 

Phase lB-1 of the program. Regulatory transition of other TWRS activities might involve older 

facilities with less documentation and could result in more significant requirements. 

D27. SCOPE OF NRC REGULATION 

Issue 

DOE has discussed the scope of NRC regulation, including facilities beyond the immediate 

TWRS/WTP facilities. 

Discussion 

HLW has been stored in large underground storage tanks at the Hanford site since 1944. 

The existing tank waste, as well as new waste added to the tanks farms, is regulated by the 

TPA's RCRA enforcement provisions. The TPA, initially issued in 1989, is an enforceable 

agreement among DOE, Washington State Department of Ecology, and EPA for achieving environmental compliance at the Hanford site. Under the TPA, the 149 single-shell tanks must 

be emptied by 2018, and the 28 double-shell tanks must be emptied by 2028. All processing 

must be completed by 2028.236 The TPA was modified in 1996 to permit private companies to 

234 Code of Federal Regulations, 7tl•e 10, Energy, Section 76.35(b), "Certification of Gaseous Diffusion Plants, 

Contents of Application." 

235 Buhl, A.R., T. Murley, G. Edgar, and D. Silverman. "NRC Regulation of Doe Facilties." Nuclear News, p. 32. 

May 1997. 

236 Environmental Protection Agency (US)(EPA) and State of Washingtong Department of Ecology. "Hanford Federal Facility Agreement and Consent Order" (-ri-Party Agreement between DOE,EPA, Ecology).

NUREG-1747 252
perform remediation of the tank waste in response to a DOE initiative to encourage industry to 

use innovative approaches to remediate the tank waste.237 

The DOE TWRS-P program was developed to address the remediation of the tank waste 

component of the overall TWRS program using a privatized contractor. Within the TWRS-P 

program, the DOE regulatory program had been structured to facilitate the possible future 

transition to NRC regulation. An MOU between DOE and NRC also had been prepared to 

ensure that the DOE regulatory program is consistent with NRC's regulatory approach and to 

allow NRC to gain sufficient experience and knowledge of the TWRS-P activities to be able to 

license the TWRS-P facility (i.e., the vitrification facility) in the future.' 

As initially envisioned by DOE, both vitrification of tank waste and storage of tank waste would 

be privatized, and the NRC would regulate all TWRS activities of the privatized contractor(s). This was described as Phase II of TWRS-P. At present, the NRC has no direct involvement in 

overseeing tank-related activities. Although NRC has licensing authority over any facilities 

expressly authorized for the long-term storage and disposal of defense HLW, previous NRC 

legal determinations have indicated that the storage of HLW in the tank farms does not 

constitute long-term storage and is not subject to NRC licensing. Thus, legislation would likely be needed to grant the NRC the authority to regulate the tank farms (legislation is also needed 

for NRC to regulate the vitrification facility-it could be part of the same legislative package). DOE, using its own rules, Orders, standards and other directives currently regulates the tank 

farms. 

From the perspective of DOE, focusing financial resources and NRC regulatory attention on the 

vitrification facility alone could result in insufficient consideration of the considerable 

environmental and public risk posed by continued operation of the tanks with their associated 

leakage and potential for collapse and explosion. Conversely, having both the tank farms and 

the vitrification facility regulated by the NRC would ensure that NRC's regulatory approach and 

principles and a consistent regulatory framework are applied to both facilities. In evaluating regulatory requirements and decisions in the context of the entire TWRS complex, the NRC 

could balance the risks posed by continued operation of the tank farms against the risks posed by operation of the vitrification facility, thereby selecting the alternative that minimizes the net 

risk to the environment and to the public. The NRC staff considers the term "consistent 

regulatory framework" to mean similar baseline requirements, similar risk based performance criteria, and similar regulatory procedures (license and amendment reviews, inspections, and 

enforcement). 

A consideration with respect to having the NRC regulate the tank farms is that regulation would 

most likely be done using a certification process. (The requirement for using a licensing 

process or a certification process would likely be specified in the legislation needed to grant the 

NRC the authority to regulate the tank farms.) To implement certification, a new section (Part) in the NRC regulations would need to be developed to specify the certification process and the 

237 Department of Energy (U.S)(DOE). EIS-01 89-SA23, "Supplement Analysis for the Tank Waste Remediation 

System." DOE: Richland, Washington. May 1998. 

2W "Memorandum of Understanding between the U.S. Nuclear Regulatory Commission and U.S. Department of 

Energy." January 29, 1997.

253 NUREG-1747
standards to be applied to the tank farms. The specific requirements in this new Part could be 

different than the requirements to be applied to the vitrification facility (i.e., 10 CFR Part 70). 

The tank farms are currently et government-owned, contractor-operated facility. If the tank 

farms were to be transitioned to NRC regulation, the licensee/certificate holder could be DOE, the contractor, or both. DOE: Ihas previously expressed interest in transitioning operation of the 

tank farms to a privatized contractor, similar to that originally planned for the vitrification facility. This was described as Phase II of TWRS privatization. The licensees/certificate holders for the 

tank farms and the vitrification facility might be different, and separate licenses/certifications 

could be required for NRC regulation of these facilities. Thus, the NRC regulatory focus would 

tend to center on whether the activities occurring at each individual facility are meeting the 

commitments in each safety analysis report and the conditions of each license/certificate. 

Further more, separate NRC licensing project managers would likely be assigned for the tank 

farms and vitrification facility, and separate docket files would likely be opened in each case. 

However, to the extent that both facilities were licensed/certified by a single safety regulator, 

greater consistency of safety, decision making can be expected. 

Cases exist or will exist on DOE sites where NRC regulates one facility and DOE regulates other facilities, although these have much simpler interfaces than at TWRS. The TMI-2 

Independent Spent Fuel Storage Installation (at the DOE Idaho Site) is an example of a case 

where NRC will license and regulate a new facility that will receive waste from an existing facility, specifically the Test Area North (TAN) facility. In this case, DOE regulates current TAN 

facility spent fuel pool storage operations and will regulate dry storage cask loading operations performed at the TAN facility. NRC regulation will begin when the fuel leaves the TAN facility boundary in an approved transportation cask.2 9 In another case, at the Portsmouth and 

Paducah GDPs, certain operations have been certified by NRC while others remain under DOE 

control. This shared site arrangement required that regulatory boundaries be established to 

ensure a clear understanding of responsibilities for regulatory oversight and the rules governing 

particular portions of the plants. 24° 

Preliminary NRC Staff Assessment 

The NRC staff anticipates that the scope of NRC regulation-the TWRS/WTP facilities, tank 

farms, other Hanford facilities--will be defined in the enabling legislation and this will address 

this issue. A single regulatory entity offers consistency and the potential to consider total risks 

from tank waste management. However, absent specific design and quantitative analysis, it is 

not possible to assess balancing total risk for the tank waste/HLW related systems at Hanford 

and achieving adequate assurances of safety is likely to be based upon individual facility 

analyses. 

239 Department of Energy (U.S.)(DOE). Rev. 0. Section 1.2, "License Application for the Idaho National Engineering 

Laboratory Three Mile Island Unit Two Independent Spent Fuel Storage Installation." DOE: Idaho Operation Office. 

October 1996. 

240 Buhl, A.R., T. Murley, G. Edgar, and D.Silverman. "NRC Regulation of DOE Facilities." Nuclear News. p. 32. 

May 1997.

NUREG-1747 254
D28. LOCATION OF THE REGULATOR

Issue 

DOE believes the regulator should be located near the Hanford site. 

Discussion 

The location of the regulator is a potential issue, especially prior to the start of operations. Ideally, the regulator should be close to the regulated. As a practical matter, since the regulator 

usually has more than one licensee, this is normally not possible. However, the issue of NRC 

assumption of regulatory authority for DOE facilities, or facilities on DOE land, is new. 

As a regulator, the NRC usually stations a resident inspector locally, has a lead field office, and 

maintains a technical review staff at the NRC Headquarters. The resident inspector and the 

field office focus on daily and more immediate issues, including inspections and enforcement, 

and routine meetings with the facility staff on operations. The Headquarters staff focuses on 

licensing and evaluation issues, and includes more specific area technical specialists. For 

regulatory activities associated with the proposed TWRS/WTP facilities, the NRC would 

probably maintain several FTEs as residents at the Hanford site, have a dedicated team of 

several FTEs at the Region IV Office, and maintain a staff of around 15 specialists at the NRC 

Headquarters. 

Preliminary NRC Staff Assessment 

The NRC staff anticipates that NRC regulation of TWRS/WTP facilities would involve some 

staff FTEs at the Hanford site, some staff FTEs at the Region IV Office, and some staff FTE's 

at the NRC Headquarters. The actual staffing levels would be determined by Congress via the 

enabling legislation and the scope of NRC regulation it requires. 

D29. MULTI-STEP LICENSING 

Issue 

The program for TWRS/WTP is split into several phases, and a multi-step licensing process 

may be more appropriate. 

Discussion 

NRC has adopted one-step licensing for 10 CFR Part 70 licenses. TWRS-P had contractually established a two-step process with separate authorization of contraction and operation. Regulatory activity also occurred during preliminary design of TWRS-P, and is likely to occur 

during the completion of the design under the M&O contractual arrangement. The NRC does

255 NUREG-1747
not have a well-defined approach for regulating preliminary design work although such activities 

have occurred in the past and are occurring now.241 

The RU expects preliminary design to be the basis of the Contractor's PSAR in the 

Construction Authorization Report (CAR). The NRC would normally expect their 10 CFR 

Part 70 license applications to be supported by a more advanced design. If the NRC were to 

assume regulatory authority after the start of construction but before the start of operations, the 

Contractor may have to upgrade the PSAR to a level consistent with NRC's expectations before 

this transition. However, the NRC regulatory approach is flexible and can adapt to a two-step approach with a submittal for a construction permit and a subsequent submittal of a license 

application for operations. 

In the revised 10 CFR Part 70,242 NRC is requiring that a licensee effectively perform a second 

step in the one-step licensing process by requiring that a preliminary hazards analysis be 

performed on the preliminary design prior to construction of new facilities. NRC notes that this 

could be viewed as an additional licensing step if approval was required, but it is not clear 

whether this step would include public participation. Without public participation it could seem 

to be an NRC approval of work already done on the preliminary design.243 

Preliminary NRC Staff Assessment 

The revised 10 CFR Part 70 allows multi-step licensing. An amendment process would allow 

the constructed TWRS/WTP facility to obtain approval for changes, different flow capacities, and different radionuclide inventories. This is not an issue. 

D30. DIFFERENCES IN 10 CFR PART 20 AND 10 CFR PART 835 

Issue 

DOE believes differences in 110 CFR Part 20244 and 10 CFR Part 835245 need to be understood 

and reconciled to the extent necessary. 

241 In the licensing of the US-DOE Clinch River Breeder Reactor Plan in the late 70's when regular meetings on 

parts of the design basis and safety analysis were presented to the NRC in Germantown at biweekly meetings thus 

obtaining valuable prelicensing feedback. The MOX facility is anticipated to be a two-step licensing effort with considerable, upfront involvement and interactions between the NRC and the DOE contractors. The NRC Staff have 

also been involved in the TWRS Program itself from 1997 through 2000. 

242 Code of Regulations, TitlelO, Energy, Revised Part 70, "Licensing of Special Nuclear Material." Issued for public viewing in July 1998. 

243 BNFL Inc. "Contractors Initial Salety Analysis Report," BNFL Inc.: Richland, Washington. 1997. 

244 Code of Federal Regulations, Tit/l 10, Energy, Part 20, "Standards for Protection Against Radiation." 

245 Code of Federal Regulations, Title 10, Energy, Part 835, "Occupational Radiation Protection."

NUREG-1747 256
Discussion

DOE promulgated primary standards for radiation protection in December 1993 when 10 CFR 

Part 835 was issued as a final rule.2' The new rule was intended to codify current radiological 

safety requirements contained in DOE orders and directives; however, for various reasons, not 

all aspects of radiological safety were included in the rule. For example, respiratory protection 

was not covered under the rule because another directive, the DOE Radiological Control 

Manual247, already contained respiratory protection requirements. At the time 10 CFR Part 835 

was issued, all DOE contractors were required to incorporate compliance to the DOE 

Radiological Control Manual in their contracts. (This requirements has since been modified.) 

For the most part, DOE and NRC radiation protection regulations are similar, but a few 

differences exist. The most notable difference is that requirements found in 10 CFR Part 20 

are more comprehensive than the 10 CFR Part 835 requirements, but where differences in 

dose values exist, the DOE requirements are usually more conservative. Areas found in 10 

CFR Part 20 and not covered by 10 CFR Part 835 include: 

1. Respiratory protection including respirator protection factors. 

2. Sealed radioactive source control. 

3. Radioactive waste disposal. 

4. Packaging receipt. 

The TWRS-P Contractor was required to comply with all requirements in 10 CFR Part 835, and 

any additional requirements it elected to include in the Safety Requirements Document (SRD). 

The additional safety criteria 2'4 address areas covered in 10 CFR Part 20, but not covered by 

10 CFR Part 835. The NRC's Special Projects Branch reviewed the applicable sections of the 

TWRS-P Contractor's SRD and concluded that, in general, the SRD radiation protection 

standards are equivalent to 10 CFR Part 20, and if properly implemented, the TWRS-P 

Contractor's Radiation Protection Program (RPP) could be compliant with the NRC regulations. 

However, the NRC staff did not review the TWRS-P Contractor's SRD for specific compliance 

to 10 CFR Part 20. The RU agreed with this assessment. 

Both agencies also agreed that the key to developing a 10 CFR Part 20 compliant RPP lies in 

the details of the implementation of the SRD Safety Criteria for radiation protection. The 

TWRS-P Contractor had not developed the implementation details. It was expected that the 

level of detail needed to better evaluate compliance with 10 CFR Part 20 would become 

available at the PSAR stage when the TWRS-P Contractor would have submitted a revised 

246 Department of Energy (U.S.). Federal Register. Vol. 588, No. 238. 

247 Department of Energy (U.S)(DOE). EH-0256T, Rev. 1, "Radiological Control Manual." DOE: Richland, 

Washington. DATE: 

248 Safety Criteria are standards that the TWRS-P contractor included in the Safety Requirements Document. 

These Safety Criteria are generally based on regulations or industry standards, but can be created ad hoc.

257 NUREG-1747
RPP for construction. It is expected that a similar approach will be taken by the new 

contractors working under M&O arrangements. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this a transition issue. 

D31. COST OF CONTRACTOR DOCUMENTATION 

Issue 

DOE expects there may be some costs associated with changing documentation to meet NRC 

expectations after regulatory transition. 

Discussion 

Regulation by the NRC may require additional documentation, different formats, more 

specificity, or additional analyses. DOE used the example of the GDPs since the regulatory oversight of the GDPs.recently transitioned to the NRC from DOE; DOE evaluated the costs 

involved in the GDP documentation changes. A source at DOE's Oak Ridge Office put the cost 

of revised paperwork alone (independent of any physical change at the facilities) at $50 million 

plus for revision of procedures, the SAR, and other documents for the GDPs. It should be 

noted, however, these costs were for old plants in which existing documentation was lacking and many of the documents had to be created or required major updates, and 

training/certification of personnel was necessary. In addition, the GDPs were not in compliance with the DOE regulations and requirements at the time of transition, which contributed to the 

costs involved. It would be expected that if the TWRS/WTP contractors were in compliance with the DOE regulatory framework at the time of transition to NRC regulation, the costs for 

revision of new plant documen1ts (all of which would likely be available in electronic formats) would be substantially less. Given the use of M&O contracts, DOE would fund these 

documentation requirements and changes. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this a significant transition issue. 

D32. THE ROLE OF DOE ORDERS VERSUS NRC RULES 

Issue 

DOE is concerned that there may be misunderstandings regarding the applicability and 

enforceability of the DOE Orders and their relationship with DOE rules. 

Discussion 

In the transition of regulatory oversight from DOE to the NRC for the GDPs, the NRC presumed that the GDPs were required to strictly comply with DOE Orders. However, that was not the 

case. Thus, the role of DOE Orders in the conduct of DOE business required clarification. The 

term "Order" or more precisely, "Safety Order" is a misnomer and leads one to a basic

NUREG-1747

258
misconception in understanding the DOE regulatory process. The terminology problems can be 

exacerbated because the NRC regulatory structure also includes the term "Order" but it is used 

in the more conventional sense. NRC Orders have the force of law and are issued to licensees 

as mandatory requirements. According to DOE and contrary to what the term "Order" implies, 

DOE Safety Orders are not, on their own, legal requirements and are not necessarily 

mandatory. 

DOE Safety Orders were not promulgated according to the Administrative Procedures Act 

requirements and are not legally binding. The Administrative Procedures Act requires public 

noticing and a comment period, among other things, before an Order is promulgated. Absent 

this process, Orders can be incorporated into a contract to be administratively enforceable 

under the terms and remedies provided in the contract. Without a statutory or contractual 

requirement, implementation of a particular Order typically involved an agreement by DOE to 

compensate the contractor for any additional burden associated with Order compliance. The 

Price-Anderson Act Amendments of 1988 authorized DOE to impose civil and criminal penalties 

for violations of enforceable orders. However, in practice, there were no enforceable orders 

and few civil or criminal penalties were ever imposed. 249 

From the DOE perspective, the need for defining "DOE Orders" during the DOE/NRC transition 

process became apparent during discussions related to transition of the GDPs to NRC 

oversight. The GDPs were constructed in the early 1950s and operated through multiple DOE 

regulatory regimes. During near term (1990) regulation, the plants were subjected to 

aggressive DOE inspections (Tiger Teams) which identified noncompliance with DOE Safety 

Orders, recommended privatized operations, and recommended NRC oversight. Even before 

1992, when DOE was directed to lease the GDPs to USEC, an upgrade program was initiated 

and new DOE Safety Orders were issued.2 ° 

Order-related lessons learned from the GDP transition are not generally applicable to TWRS. 

DOE Order compliance was specifically excluded from the TWRS-P contract.25' Both the DOE 

and NRC have been involved in developing the TWRS-P regulatory (Integrated Safety 

Management Plan [ISMP] and SRD) and design (Integrated Safety Analysis [ISA] and SAR) 

bases. The regulatory basis for the new contracts has not been defined yet. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this an issue. 

249 Defense Nuclear Facilties Safety Board. Tech 5. 

250 Department of Energy (U.S)(DOE). ORO-2051, "Regulatory Oversight Program." DOE: Oak Ridge, Tennessee. 

September 1997. 

21 Department of Energy (U.S.)(DOE) Rev. 1, "Radiological, Nuclear, and Process Safety Regulation of TWRS 

Privatization Contractors." DOE: Richland, Washington. January 1998.

259 NUREG-1747
D33. THE ROLE OF THE NATIONAL ENVIRONMENTAL POLICY ACT IN REGULATORY 

TRANSITION 

Issue 

DOE believes the potential effects of the National Environmental Policy Act (NEPA) need to be 

considered for regulatory transition. 

Discussion 

Presently, DOE has responsibility for ensuring compliance of the TWRS/WTP activities to 

NEPA. As part of its compliance activities, DOE has issued an Environmental Impact 

Statement (EIS) that includes the TWRS/WTP activities. However, when TWRS-P regulatory oversight transitions to the NRC, NRC will also have to comply with NEPA requirements 

pertinent to its activities. 

DOE's implementing regulations for NEPA are found in 10 CFR Part 1021. Similarly, the NRC 

implementing regulations are found in 10 CFR Part 51 and provide for the adoption of other 

Federal agencies' NEPA documentation, as well as the participation of the NRC in the NEPA 

process of other Federal agencies as a lead or cooperating agency. 

Both DOE and NRC rules penmit the flexibility of being lead agencies or cooperating agencies 

and to work along with other Federal entities on NEPA documentation. In the past, the NRC 

has not viewed the task of NEPA coordination and integration as a large challenge, and 

normally works with other agencies when preparing EISs. 

Since the TWRS/WTP project is a major Federal action, DOE had prepared an EIS in the early 1990s for it and related activities. The EIS considers the actual cleanup activities and 

operations of the TWRS-P facilities and includes construction, operation, and decontamination 

and decommissioning activities. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this an issue because of the existing NEPA coverage and the 

similarities between the two agencies in their approaches to NEPA. It is possible that an 

updated or supplemental EIS/lEnvironmental Assessment (EA) may become necessary, based 

upon the final approaches selected by the TWRS/WTP contractors. 

D34. ROLE OF THE ACRS OR ACNW IN REGULATORY OVERSIGHT TRANSITION 

Issue 

DOE is concerned by the potential effects of the Advisory Committee on Reactor Safeguards (ACRS) and/or the Advisory Committee on Nuclear Waste (ACNW) on the transition of 

TWRS/WTP regulation from DOE to NRC.

NUREG-1747 260
Discussion

The ACRS was established in 1957 by revision of the AEA of 1954 and provides advice to the 

NRC on potential hazards of proposed or existing nuclear reactor facilities and the adequacy of 

proposed safety standards. The AEA also requires that the ACRS advise the Commission with 

respect to the safety of operating reactors and perform such other duties as the Commission 

may request. Consistent with the Energy Reorganization Act of 1974, the Committee will 

review any material related to the safety of nuclear facilities and activities of the DOE 

specifically requested by DOE. Upon request, the ACRS has also provided advice to the 

DNFSB and the U.S. Navy. In addition, the ACRS, on its own initiative, performs reviews of 

specific generic matters on nuclear facility safety-related items. Activities of the ACRS are 

conducted in accordance with the Federal Advisory Committee Act, which provides for public 

attendance and participation in ACRS meetings. 

The NRC established the ACNW in 1988. Until that time, the ACNW was a subcommittee of 

the ACRS. The ACNW reports to and advises the NRC on nuclear waste disposal facilities as 

directed by the Commission. This includes 10 CFR Part 60 and 10 CFR Part 61, and other 

applicable regulations and legislative mandates such as the Nuclear Waste Policy Act, the Low

Level Radioactive Waste Policy Act, and the Uranium Mill Tailings Radiation Control Act, as 

amended. The primary emphasis of this Committee is waste disposal facilities. 

Although neither of the above descriptions of the ACRS/ACNW charters directly encompass 

TWRS/WTP activities, a DOE discussion with the Executive Director of ACRS/ACNW indicates 

that one or both of the advisory committees would likely evaluate issues related to licensing of 

TWRS/WTP. The DOE/RU had concerns that ACRS/ACNW involvement in the regulatory 

oversight transition may include the potential for delay, and DOE discussed these concerns with 

the Executive Director and ACNW Nuclear Waste Branch Chief. Both indicated that although 

the committees may ask for additional information following a presentation of the proposed 

action, there is little likelihood, in their view, of a substantial delay of the TWRS-P transition 

process. The committees meet eight times per year and submit their reports within 1 month 

following the meetings. 

An ACNW letter to NRC Chairman Jackson, dated December 23, 1997, forwarded the ACNW 

1998 Strategic Plan and Priority Issues and Activities. Listed as a second-tier priority was a 

commitment to review waste-related activities associated with the NRC's possible regulation of 

certain DOE facilities if NRC assumes responsibility for those activities as result of privatization 

or enactment of new legislation. Following their review of the September 16, 1996, NRC 

Strategic Plan Framework Document, the ACNW endorsed the concept of DOE regulation by 

NRC and expressed associated recommendations. 

Based on the above discussion, the ACNW would likely take an interest in the transition of 

regulatory responsibility associated with TWRS-P. A review of ACNW reports from July 1996 to 

the present suggests that a typical ACNW recommendation would be expected within 3 months 

of an initial presentation of the issue to the Committee. Some issues that were not time 

sensitive were not concluded for up to a year, while reports on other issues were released 

within 1 week of the presentation to the Committee. Records of recent meetings suggest that 

ACNW involvement in the TWRS-P regulatory oversight transition process would not result in 

any delays. ACNW participation in the process would occur in parallel with other 

transition/licensing activities.

261 NUREG-1747
Preliminary NRC Staff Assessment 

The NRC staff does not consider this to be a significant issue because no delays are 

anticipated. 

D35. IMPACT ON FUTURE GENERATIONS 

Issue 

DOE is interested if there would be any impacts from regulatory transition upon future 

generations and resources. 

Discussion: 

The premise for the TWRS-P vitrification plant is that it will take waste, process it, and return a 

stable vitrified waste form for, safe, long-term storage and disposal. By doing so, DOE intends 

to remediate the Hanford wasl:e tanks and, eventually in the mid-21st Century,252 return the land 

to beneficial use as it was prior to 1940. Future generations in the latter half of the 21st 

Century would benefit by greater land resources than are available now. 

As previously noted, external regulation of DOE facilities, including TWRS/WTP facilities, is 

expected to improve the safety and protection of workers, the public, and the environment. 

This can only have a positive effect upon the future. 

Preliminary NRC Staff Assessment 

The NRC Staff does not see this as an issue for regulatory transition. Anyschedule impacts 

from regulatory transition would be minuscule compared to the long time periods involved. 

D36. PUBLIC INPUT TO REGULATORY DECISIONS 

Issue 

DOE is concerned that regulatory decisions may not continue to be open to the public and 

public input after transition of regulation to the NRC. 

Discussion 

When the Hanford reservation was established, it was under a cloak of secrecy, and decisions 

were made in wartime conditions where the imperative was the successful conclusion of the 

war. Decisions were made by the military with little input from anyone else. This unilateral form 

of decision making extended to DOE's management in the days of the cold war. Even as late 

as the 1970s, very little input was accepted from Stakeholders other than that which was 

legislated (as for the Agreement Tribes) or had an impact on Congressional funding 

(politicians). 

252 The Tn-Party Agreement specifies complete clean-up of tank wastes by 2028, following which time the tanks 

themselves and all appurtenances would require decontamination and decommissioning and disposal.

NUREG-1747

262
Now, however, the situation is different. The DOE/RU has an Openness Policy' that makes it 

clear that Stakeholder's views are important and will be addressed. The RU satisfies this policy in a number of ways, both in providing information and in listening to the views of others in open meetings and within such groups as the Hanford Advisory Board. Much of the interaction is 

fairly informal and Stakeholders have a number of ways in which they may submit comments 

and concerns. DOE has undertaken to disposition all Stakeholders' comments. 

The transfer of TWRS/WTP to NRC regulatory oversight would result in the NRC licensing or 

certifying the TWRS/WTP Contractor's facility. In any licensing action, NRC is required to 

obtain public participation under 10 CFR Part 2, the Rules of Practice. In this process, the 

interaction is more formal and judicially governed than in the DOE case. The NRC generally 

conducts interactions with the licensee in the public domain to the maximum extent possible. 

The ongoing pre-licensing activities on the MOX project are a good example. 

Preliminary NRC Staff Assessment 

The NRC staff does not consider this an issue as NRC practices and procedures usually result 

in more openness with the public than DOE regulatory actions. 

253 Department of Energy (U.S.)(DOE) RL/REG-97-04, Rev. 2, "Policy for Openness and Openness Plan." DOE: 

Richland, Washington. June 1998. "The radiological, nuclear, proces safety regulation of TWRS-P contractors 

provide that the Hanford site shall be transacted publicly and candidly....' "The Tribal Nations and Public should be 

involved in decisions concerning privatization contractor regulation. Their involvement improves our processes and 

products and helps (to) ensure safety. We welcome and encourage this involvement."

263 NUREG-1747
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NUREG-1747 264
APPENDIX E 

LISTING OF U.S. NUCLEAR REGULATORY COMMISSION 

CORRESPONDENCE 

WITH THE U.S. DEPARTMENT OF ENERGY ON 

TANK WASTE REMEDIATION SYSTEM-PRIVATIZATION 

1. 11/20/96 Lettter to T. Sheridan from NRC RE Preliminary Comments RE Initial Quality 

Assurance Program, RL/REG-96-01, Rev. 0 

2. MOU signed by Hazel O'Leary & Chairman Jackson on 1/15/97, Published in FR 3/13/97 

RE MOU: Cooperation & Support of Significant Projects &Activities 

3. MOU signed by J. Wagoner, DOE, on 1/15/97 and C.Paperiello, NRC, on 1/29/97, 

Published in FR 3/18/97 RE MOU: Cooperation & Support for Demonstration Phase 

(Phase I) of DOE Hanford TWRS Privatization Activities 

4. 2/21/97 Letter to J. D. Wagoner from C.J. Paperiello RE Baseline Guidance Documents for 

Operation of the NRC's TWRS Unit 

5. 2/27/97 Letter to T. Sheridan from NRC RE Comments Provided RE Guidance for 

Employees Concerns Management System 

6. 8/17/97 NRC Comments provided to DOE via letter RE LMAES Standards Approval 

Submission (ISMP, SRD) 

7. 10/17/97 and 10/30/97 NRC Comments provided to DOE/RU via letter RE BNFL 

Standards Approval Submission (ISMP, SRD) 

8. 2/6/98 Letter to DOE with NRC Comments RE BNFL Initial Safety Analysis Report (ISAR) 

9. 2/13/98 Letter to DOE with NRC RE LMAES Initial Safety Assessment Package (ISAP) 

10. 4/3/98 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments Provided on 

BNFL's QA Program and Implementation Plan 

11. 5/21/98 Letter to J. Wagoner, DOE, from C. Paperiello, NRC, EDO- 98-00191/NMSS

980078; RE: Co-Located Worker Standards for Accidents (Encl 1 SECY-98-38, Encl 2 

4/28/98 SRM) 

12. 6/22/98 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC RE Forwarding 2 Draft 

Documents- -Draft Part 70 and SRP for TWRS 

13. 6/29/98 Letter to D.C. Gibbs, DOE from R. Pierson, NRC, RE Issue Paper: Seismic 

Events for TWRS

265 NUREG-1747
14. 7/2/98 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Providing 2/23/99 Comm 

Paper "Modifications to MOU" 

15. 7/10/98 Letter to D.C. Gibbs, DOE from R. Pierson, NRC, RE Forwarding Comments on 

Construction Authorization Request Outline 

16. 7/23/98 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE NRC Comments on BNFL's 

Revised SRD & ISMP 

17. 10/30/98 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments on BNFL's 

Report "Radiation Protection Program for Design 

18. 11/6/98 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments on Regulatory 

Unit Disposition of earlier comments 

19. 11/19/98 Letter to G. Wagoner, DOE, from C.J. Paperiello, NRC, RE Co-Located Worker 

Standards for Accidents 

20. 12/1/98 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments on BNFL 

transmittal of Criticality Safety Program PL-W375-NS00001 

21. 12/2198 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Protocol Proposal for 

Inspection Assistance to DOE RU Inspections of TWRS-P contractor facilities and activities 

22. 12/3/98 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Issue Paper on Future US 

NRC Role for the Hanford TWRS 

23. 12/03/98 Letter to C. Gbbs, DOE, from R. Pierson, NRC, RE TWRS Comments on 

"Demonstrations on the DOE-96- 0004 Process," BNFL Rpt-W375-PRO0002, Nov 11, 

1998. 

24. 12/11/98 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments on BNFL 

Submittal of 11/19/98 and 12/98 Seismic Topical Mtg. 

25. 12/17/98 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Transmits Inspection Accompaniment Report (9B-01): Employee Concerns Program 

26. 12/31/98 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Comments on BNFL's 

TWRS Safety Criteria for Environmental Radiation Protection Program and for 

Environmental Radiation Monitoring 

27. 1/08/99 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Revised Protocol for NRC 

Inspection Assistance. 

28. 1/22/99 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE BNFL Criticality Safety 

Program.

NUREG-1747

266
29. 2/11/99 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Comments on BNFL's 

TWRS Safety Criteria for Environmental Radiation Protection Program (ERPP) and for 

Environmental Radiation Monitoring. 

30. 2/12/99 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE: Preliminary Comments 

Provided on 1/28/99 Topical Meeting-Review of Hydrogen & Explosion Info 

31. 2/19/99 Letter to D.C.Gibbs, DOE, fm R. Pierson, NRC, RE Draft SRP for the Review of a 

Licence Applicaiton for TWRS-P Project. 

32. 2/19/99 Letter to D.C. Gibbs, DOE, fm R.C. Pierson, NRC, RE Comments Provided on 

Review of Fire Protection Info Provided by BNFL 

33. 2/24/99 Letter to D.C. Gibbs, DOE, fm R.C. Pierson RE Review of Quality Assurance 

Inspection Plan and Inspection Technical Procedure 1-101. 

34. 3/1/99 Letter to D.C. Gibbs, DOE, fm R.C. Pierson, NRC, RE Comments Provided on 

Proposed Approach for Evaluating Compliance with RESW Requirements 

35. 3/18/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE forwarding NRC 

comments on BNFL's Design Safety Features. 

36. 3/19/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Inspection 

Accompaniment Report (99-01): Training & Qualification Program 

37. 4/1/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC RE NRC comments on Draft 

RL/REG-99-05, "Guidance Review Document for the Construction Authorization Request. 

38. 4/21/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson RE Forwards CNWRA Report on 

Estimate of Hydrogen Generation from Wastes at the Proposed TWRS-P Fac. 

39. 4/23/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments forwarded on 

Methods for Assessing Consequences of Potential Accident Radiological Releases from 

the TWRS-P Fac (Preliminary). 

40. 5/10/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments Forwared on 

3/18/99 BNFL Transmittal of Seismic Documents. 

41. 5/28/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Transmittal of Point Paper 

on Process Safety at a TWRS-P Facility. 

42. 6/4/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Transmits copies of SECY

99-107 & related Secretary Requirements Memorandum 

43. 6/11/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Inspection 

Accompaniment Rpt (99-02), "Quality Assurance."

267 NUREG-1747
44. 6/14/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments provided on 

Chapter 4 of Guidance Document for Construction Authorization Request. 

45. 6/22/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE 5/25/99 RU Activity 

Report from Jan-April 99 (Triennial Report). 

46. 6/22/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments on BNFL Inc. 

Quality Assurance Plan and Implementation Program. 

47. 6/25/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE DOE 5/5/99 Ltr on 

Disposition of the NRC Comments on the BNFL Inc. Design Safety Features Submittal. 

48. 7/2/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Expenditures on 

Hanford TWRS Related Work from FY97- 2nd Qtr FY99. 

49. 7/2/99 Letter to D.C. Gibbs, DOE, from R. Pierson, NRC, RE Providing 2/23/99 Comm 

Paper "Modifications to MOU." 

50. 7/9/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Comments on BNFL Inc. 

6/14/99 Submittal "Info for June Topical Meeting. 

51. 7/15/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, DOE, RE NRC Inspection 

Accompaniment Report ('99-03): Self Assessment & Corrective Action Plan. 

52. 7/16/99 Letter to D.C. Gibbs, DOE, fromm R.C. Pierson, NRC, RE 5/99 Sellafield Report 

Sent. 

53. 7/27/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Review of SW Research 

Institute Report: "Radiological Binding on the Synthetic Biopolymer Diazoluminomelanin." 

54. 7/29/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Inspection 

Accompaniment Report (99-04): Radiation Protection Program. 

55. 8/5/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Public Concern of 

Potential Inappropriate Use of 304 Stainless Steel in Tank Farm (Ralph Davison). 

56. 8/13/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Comments on 

BNFL Report, "RPP for Design, Rev 2 & Rev 0." 

57. 8/17/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC RE Comments on Inspection 

Program for the Regulatory Oversight of the RPP-P, Rev 3 (RL/REG-98-05, 7/1/99). 

58. 8/19/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE BNFL Record Storage. 

59. 8/25/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Inspection 

Accompaniment Report 99-05, "Configuration Management Program."

NUREG-1747

268
60. 8/26/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Position on Pipe 

Damping Relative to the 1998 ASME Boiler & Pressure Vessel Code, Section III, App N, as 

Rev'sd by '99 Addenda. 

61. 11/1/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Transmittal of Report by 

CNWRA "Process Hazards & Safety Issues for TWRS-P" Volumes I & I1. 

62. 11/2/99 Letter to D.C. Gibbs, DOE from R.C. Pierson, NRC, RE Comments on Rev 4C of 

BNFL Quality Assurance Program and Implementation Plan. 

63. 11/4/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC and CNWRA 

Comments on Explosive Hazards Topical Meetings I & II, 8/24/99 & 9/28/99. 

64. 11/4/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRc, RE Quality Assurance for 

Limited Construction Authorization Acts. 

65. 11/4/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE NRC Comments on 

RL/REG-99-19, Rev. 0, "Analysis of Radiation Materials at Risk Supporting Source Term 

Development for TWRS-P Facility. 

66. 11/12/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Point/Status Paper on 

Quality Assurance. 

67. 11/12/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Transmittal of 

Comments on "Proposal to Modify the Current RPP-WTP Position with respect to Process 

Chemical Hazards for TWRS-P," 9/28/99. 

68. 11/15/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE 10/22/99 RU Response 

to NRC Point Paper on Process Safety at TWRS-P Facility. 

69. 11/23/99 Letter to D.C. Gibbs, DOE, from R.C. Pierson, NRC, RE Response to BNFL Inc.'s 

Comments on CNWRA Report on Estimates of TWRS-P Hydrogen Generation. 

70. 12/14/99 Letter to D.C. Gibbs, DOE, from M. Galloway, NRC, RE NRC Observations on 

the Topical Meetings with BNFL Info for 11/98-11/99. 

71. 12/23/99 Letter to D.C. Gibbs, DOE, from M. Galloway, NRC, RE Information Paper 

Comparing Radiological Protection Standards in 10 CFR Part 20 & 835. 

72. 1/21/00 Letter to D.C. Gibbs, DOE, from M. Leach, NRc, RE Potential Critical Tech Issues 

for Construction Authorization Resolution. 

73. 1/28/00 Letter to D.C. Gibbs, DOE, from M. Leach, NRC, RE NRC Inspection 

Accompaniment Report (01): Design Process. 

74. 2/25/00 Letter to D.C. Gibbs, DOE, from M. Leach, NRC, RE Transmittal of Comments on 

the Report and Topical Meeting on "Compliance w/Risk Goals and Project Reliability 

Database."

269 NUREG-1 747
75. 3/24/00 Letter to D.C. Gibbs, DOE, from M. Leach, NRC, RE NRC Inspection 

Accompaniment Report (70-3091/002) Training & Qualifications. 

76. 4/12/00 Letter to D.C. Gibbs, DOE, from M. Tokar, NRC, RE Additional Comments from 

March Topical Meeting. 

77. 4/27/00 Letter to D.C. Gibbs, DOE, from M. Tokar, NRC, RE NRC Response to 3/23/00 Ltr 

on Resubmittal of Aug & Sept 99 Topical Mtg Reports on Explosive Hazards. 

78. 4/25/00 Letter to D.C. Gibbs, DOE, from M. Tokar, NRC, RE Summary of Discussion of 

TWRS-P Project RPP - Differences in Regulatory Requirements and Guidance of DOE 

and NRC. 

79. 4/19/00 Letter to D.C. Gibbs, DOE, from M. Tokar, NRC, RE Comments on TWRS-P 

Project, RU Position on Conformance w/Risk Goals in DOE/RL-96-0006, RUREG-2000

08, Rev 0 3/28/00. 

80. 5/2/00 Letter to D.C. Gibbs, DOE, from M. Tokar, NRC, RE Transmittal of Point Paper on 

Materials Issues and Safety at TWRS-P Facility.

NUREG-1747 270
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION 1. REPORT NUMBER 

(2-89) (Assigned by NRC, Add VoL, Supp., Rev., N4RCM 1102. and Addendum Numbers, if any.) 3M, 3=BIBLIOGRAPHIC DATA SHEET 

(See ins•uion on the reverse) 

2. TITLE AND SUBTITLE NUREG-1747 

Overview and Summary of NRC Involvement with DOE in the Tank Waste Remediation 

System - Privatization (TWRS-P) Program 3. DATE REPORT PUBLISHED 

MONTH YEAR 

August 2001 

4. FIN OR GRANT NUMBER 

5. AUTHOR(S) 6. TYPE OF REPORT 

A.P. Murray, M.N. Baker, F.H. Burrows, S. A. Burrows, LW. Chang, G. C. Comfort, R.E. 

Shewmaker, W.R. Smith, M. Srinivasan, R.K. Struckmeyer, M. Tokar, R.G. Wescott, A. Wong Final 

7. PERIOD COVERED (Indusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. NucdearRegulatory Commission, and mailing address; ifcontrctor, 

provide name and mailng address.) 

Division of Fuel Cycle Safety and Safeguards, Office of Nuclear Material Safety and 

Safeguards, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555 

9. POSORNGORGNIATON NDADRES (f NCte NME Smesabve~ conrato, povde RCDivatnOffceor~gin~U.S Ndea Rguatoy ommsson

9. SPONSORING ORGANIZATION - NAME AND ADDRESS (if NRC, type 'Same as above",; ifCodtractor, provide NRC Division, Office or Regiony, U.S. Nuchaar ReguiatoryCommission, and maiing address.) 

Same as Above

10. SUPPLEMENTARY NOTES 

11. ABSTRACT (200 words or ess) 

In 1995, the U.S. Department of Energy (DOE) embarked on an effort to privatize the processing through vitrification of 54 

million gallons of radioactive waste that has been stored in 177 underground storage tanks at the Hanford Site. The U.S. 

Nuclear Regulatory Commission (NRC) provided assistance to DOE on the TWRS-P program, with a potential transition to 

NRC regulatory authority at a future time. In 2000, DOE terminated the privatization approach, and decided to use more traditional contracting methods. 

During their reviews, NRC staff analyzed both unmitigated and mitigated consequences from potential accident scenarios at 

the proposed facility. NRC staff's efforts identified several key areas of uncertainty, such as melter failure modes and 

frequencies, that would require further study before more refined analyses could be performed. The reviews also identified 

several open issues, including the need for significantly more detailed design information and safety analyses, and greater 

defense-in-depth. In particular, the design, at the time of termination of the privatization, was found to be very preliminary and 

corresponded to perhaps a 15 percent level of design. 

This report summarizes NRC's participation in and observations on the TWRS-P program and identifies issues from the NRC's 

perspective.

12. KEY WORDS/DESCRIPTORS (List words or phrases that wi/f assist researchers inlocating the report.) 13. AVAILABILITY STATEMENT 

Hanford, tank, waste, remediation, TWRS, DOE, HLW, vitrification, separation, safety unlimited 

14. SECURITY CLASSIFICATION 

(This Page) 

unclassified 

(This Report) 

unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 33 (2-89)
Federal Recycling Program
UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300