ML21175A119

Request to Amend License SNM-42, Chapter 5, Nuclear Criticality Safety

ML21175A119
Person / Time
Site:	BWX Technologies
Issue date:	06/11/2021
From:	Freudenberger R BWXT
To:	Document Control Desk, Office of Nuclear Material Safety and Safeguards
References
21-030
Download: ML21175A119 (28)

Text

BWX Technologies, Inc.

June 11, 2021 21-030 ATTN: Document Control Desk Director, Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Reference:

License No. SNM-42, Docket 70-27

Subject:

Request to Amend License SNM-42, Chapter 5, Nuclear Criticality Safety

Dear Sir or Madam:

BWXT Nuclear Operations Group, Inc. - Lynchburg (BWXT NOG-L), requests approval for an amendment to Chapter 5 of the SNM-42 License Application in accordance with 10 CFR 70.34.

Enclosure 1 to this letter provides a summary of the changes to Chapter 5, Nuclear Criticality Safety, of the SNM-42 License Application and BWXT NOG-L's justification for the changes. contains the proposed revision of Chapter 5 of the SNM-42 License Application.

The revised Chapter 5 has extensive changes and as a result revision bars are not included.

BWXT NOG-L also requests the removal of License Conditions S-3, S-4, and S-11. The justification is provided at the end of Enclosure 1.

If you have questions or require additional information, please contact Chris Terry, Manager of Licensing and Safety Analysis, at ctterry@bwxt.com or 434-522-5202.

Sincerely, Richard J. Freudenberger Manager, Environment, Safety, Health and Safeguards BWXT Nuclear Operations Group, Inc. - Lynchburg Enclosures cc: NRC, James Downs NRC, Region 11 NRC, Resident Inspector P.O. Box 785 Lynchburg, VA 24505 USA t: + 1.434.522.6000 f: + 1.434.522.6805 www.bwxt.com POWERING TRANSFORMATION

ENCLOSURE 1 Summary of Changes to Chapter 5 of the SNM-42 License Application P.O. Box 785 Lynchburg, VA 24505 USA t:

• 1.434.522.6000 f: + 1.434.522.6805 www.bwxt.com POWERING TRANS FD RMATI ON

Summary of Changes to Chapter 5 of the SNM-42 License Applicatio~- . . . ]

Overview of Chapter 5 Changes The goal of this revision of Chapter 5, Nuclear Criticality Safety, is to remove items that predate the implementation of Subpart H of 10 CFR 70 (hereinafter referred to as "Subpart H") and remove explanatory text where it does not add value nor enhance safety. The amended Chapter 5 is intended to establish commitments on what will be done and reduce the discussion of the "why" and "how". If requirements are covered in other chapters of the License, it is not necessary to repeat them in Chapter 5. Examples include: training, audits, inspections, Integrated Safety Analysis (ISA) and management measures.

Major Changes to Chapter 5

1. Deletion of the discussion of Double Contingency as a "shall" and requiring the control of two parameters or the use of Defense in Depth for multiple controls on one parameter and replacing it with a commitment to follow the Double Contingency Principle as stated in ANSI/ANS-8.1-2014 which is endorsed in Regulatory Guide 3. 71, Rev. 3.

2. Elimination of the kett limit and Margin of Subcriticality (MoS) for the Limiting Condition of Operation (LCO).

• This was removed since it is a legacy item which does not provide actual safety since both the normal and credible upset conditions must be demonstrated as subcritical, and the Margin of Safety and the Mos do not correlate. This is demonstrated by the following example:

System A has a kett of 0.5. It is a sphere of High Enriched Uranium (HEU) metal hanging by a string over a large water tank. System 8 has a kett of 0.99. It is an infinite lattice of natural enrichment fuel rods optimally spaced in water. In System A, if the string fails and the metal sphere falls into the water, the system is supercritical. In System 8, nothing can be done with the rods to make them critical since it is an infinite array and is

, optimally spaced. System A has a lower normal kett, but is less safe than System 8.

3. Removal of the requirement for determining the Safety and Failure limits for each controlled parameter in an analysis.

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• This was a legacy requirement which predates Subpart H. It was established to demonstrate the sensitivity of a parameter was understood. Calculation of the Failure Limit for a parameter has little to no value with the implementation of the ISA, since no a

credible upset can result in critical condition and the* risk of a criticality accident must be highly unlikely.

4. Removal of the explanation in Section 5.2.3 for why Low Enriched Uranium (LEU) systems can have a higher kett limits (lower MoS).

• This was a legacy item which was included when a different MoS was requested for LEU materiaL This is explanatory, and is being removed for that reason.

5. Deletion of legacy limits and requirements which predate Subpart H. These include:

• Liquid effluent processing

• Dry Low Level Waste handling

6. Elimination of the Appendix to Chapter 5 (developed in 1995), but incorporating the key concepts into the body of Chapter 5.

• This appendix was added in 1995 as a compromise with the NRC prior to Subpart H.

The means and methods of controls are included in the new Section 5.2. The wording has been streamlined.

7. Three of the four methodologies described in Section 5.2.2 were removed since the objective is to focus on the "what" and not the "how" and "why."

• Removal of the discussion of the Solid Angle technique and the Lattice Density technique since they are well documented in the literature.

• Change of "Law of Substitution" to "Substitution Methodology". This is retained in the chapter since it is not a commonly used technique.

• The Water Box method was removed. The technique will be documented and used in analyses. It will be the responsibility of the analyst to demonstrate the system is subcritical under normal and credible abnormal conditions. The methodology used by the analyst must demonstrate subcriticality, and the demonstration must be approved by the reviewer. This change provides flexibility in how the system is shown to be safe.

8. Removal of the portion of Section 5.2.5 which detailed the approach used for analyzing and controlling moderation in operations.

• This is a legacy item that described how a system would be analyzed. With the implementation of Subpart H, the discussion is not needed.

9. A change was made to allow for use of historical operational data to set the bounds of credible ranges on a parameter or upset condition.

• This change provides the ability to use process data, with proper consideration for the data applicability. Previously, this data could only be used to supplement experimental data. In a manufacturing environment, it is often not possible to conduct experiments to establish ranges of a parameter. Operational data is substituted for experimental data since it was collected from the process. It does not have the controls and design to be considered an experiment, but is representative of the system.

Section Changes for Chapter 5 The specific section changes to Chapter 5 are extensive and are addressed on the subsequent pages based on the section or subsection number of the currently approved Chapter 5.

Changes to Section 5.1 and 5.1.1 Section 5.1 and 5.1.1 were restructured and made more concise in the new Section 5.1.

Summary of the new Section 5.1 NOG-L management is responsible for the safety of operations involving SNM. They are responsible for establishing and maintaining a Nuclear Criticality Safety (NCS) program which prevents a criticality accident, complies with the Double Contingency Principle and the Process Analysis requirement of ANSI/ANS-8.1-2014, and the NCS performance requirements of 10 CFR 70.61. The role and responsibility of the NCS manager is defined along with the NCS responsibilities of the area managers. NCS analyses by qualified NCS staff is required as well as documentation of the NCS analyses. Oversight of the Specialized Nuclear Criticality Safety 2

Training Program and the NCS portion of General Employee Safety Training were moved to the new Section 5.5 on NCS Training.

Several of the items listed in the current Section 5.1.1 have been removed and are discussed below:

Current Chapter 5 Amended Chapter 5 (a) preventing an inadvertent nuclear Included in the new 5.1 criticality, (b) protecting against the occurrence of an Removed since it is covered in Chapter 3, identified accident sequence in the ISA

• IsA.

Summary that could lead to an inadvertent nuclear criticality, (c) complying with the NCS performance Included in the new 5.1 requirements of 10 CFR 70.61, (d) establishing and maintaining NCS safety Safety parameters are required in Chapter 3, parameters and procedures, ISA and procedures for NCS are necessary to implement the requirements of the License.

(e) establishing and maintaining NCS safety Removed since it is covered in Chapter 3, limits and NCS operating limits for IROFS, ISA.

(f) conducting NCS evaluations to assure NCS evaluations are included under the that under normal and credible abnormal NCS Manager's responsibilities (item 3) and conditions, all nuclear processes will remain at the end of the new Section 5.1. The subcritical and maintain an approved margin requirement for subcriticality is required by of subcriticality for safety, 10 CFR 70.61 which is committed to in the new Section 5.1 '

(g) establishing and maintaining NCS Removed since it is covered in Chapter 3, IROFS, based on current NCS evaluations, ISA.

(h) providing training in emergency Removed since it is covered in Chapter 8.

procedures in response to an inadvertent nuclear criticality as described in Chapter 8, (i) complying with NCS baseline design Since this is described in Chapter 11, it was criteria requirements in 10 CFR 70.64(a) as removed from here.

described in Chapter 11, U) complying with the NCS ISA Summary Removed since it is covered in Chapter 3, requirements in 10 CFR 70.65(b) as ISA.

described in Chapter 3, and (k) complying with the NCS ISA Summary Since this is described in Chapter 11, it was change process requirements in 10 CFR removed from here.

70. 72 as described in Chapter 11.

In the current Section 5.1.1, there is a lengthy discussion of Double Contingency and Defense in Depth. This was put in place in the 1990's prior to Subpart H. It changed the ANSI/ANS-8.1 wording for the Double Contingency Principle from "should" to "shall", and implied "classic"

)

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double contingency is control of two parameters. This required the development of Defense in Depth for cases where multiple controls on a single parameter were used. Since the NRC has endorsed ANSI/ANS-8.1-2014 without an exception on the double contingency statement, the section was removed and the commitment to follow the Double Contingency Principle and the Process Analysis requirement from ANSI/ANS-8.1-2014 were added to the new Section 5.1.

Changes to Section 5.1.2 The commitments on NCS Procedures and Postings have been moved to the end of the new Section 5.2. One paragraph was not brought forward. That paragraph is:

"Storage vessels such as cans, buckets, etc., which contain special nuclear material will be labeled as to the type and amount of material. In-process material, i.e., materials being processed for use in a finished product, and scrap (10 CFR 74.4) will be handled with knowledge of type and quantity of material whenever practicable. When the type or quantity is not known, such material shall be handled in favorable geometry or volume until the material can be assayed. Dry Waste material that is contaminated with low levels of uranium may be classified by operating personnel as Dry Low Level Waste in accordance with written guidelines as established in a site-wide Quality Work Instruction. Dry Low Level Waste may be collected in appropriately labeled 55-gallon type containers. Fifty-five gallon containers to which Dry Low Level Waste has been added during any day, shall be assayed for U-235 content on that day or at a frequency approved by the Nuclear Criticality Safety Manager and specified in the above site-wide Quality Work Instruction. The U-235 content of 55-gallon Dry Low Level Waste containers shall not exceed 300 grams."

Labeling of containers is under the control of Nuclear Material Controls and the Fundamental Nuclear Materials Control Plan. It is not appropriate in Chapter 5. The controls for handling material and scrap are established in NCS analyses and documented in the ISA. The specifying of how material, scrap, or Dry Low Level Waste is handled is not appropriate in Chapter 5. This was a legacy item from before Subpart H. Removal of this paragraph does not impact the level of safety in the facility.

Changes to Section 5.1.3 Section 5.1.3 was replaced in the new Section 5.4 with a pointer to Chapter 11.5.1.

Changes to Section 5.1.4

• Section 5.1.4.1 was moved to the new Section 5.5.1. The list of items covered in General Employee Safety Training was reduced. The last three bullets are covered in the Specialized training (Section 5.5.2). General Employee Safety Training covers the safety basics for facility access. Before a person can handle SNM, Specialized NCS training must be taken. The three bulleted items better fit into the specialized training.

• Section 5.1.4.2 was moved to the new Section 5.5.2.

• In Sections 5.1.4.1 and 5.1.4.2 the requirement for the training to be developed by a Training Specialist was removed. Specifying who must develop the training does not add value. The training is developed under the oversight of NCS.

• Section 5.1.4.3 was moved to the new Section 5.5.3. The last two sentences in the last paragraph of the section were not brought forward. These two sentences justify why the 4

supervisor can judge the operator's understanding of NCS limits and to have the operator retrained. These items are implicit to the job of any supervisor. These statements do not enhance the effectiveness of the training.

Changes to Section 5.1.5

• Section 5.1.5 was moved to the new Section 5.6.

• The revision of Regulatory Guide 3.71 was updated.

Changes to Section 5.1.6 This entire section was deleted. This section is a legacy discussion which predated Subpart H.

The safety basis for liquid effluents is documented through the ISA process.

Changes to Section 5.2 The introductory paragraphs for this section were revised to remove the explanatory text since the text did not add value to the commitments of the Chapter. The essence of the introductory paragraphs was moved to the new Section 5.3.

Changes to Section 5.2.1 This section was moved to the new Section 5.3.1. In the last paragraph, Area of Applicability was changed to Validation Applicability to be consistent with ANSI/ANS-8.24-2017.

Additionally, the list of operations covered was expanded.

Changes to Section 5.2.2 This section was moved to the new Section 5.3. The sentence after "orthogonal projection" was removed and replaced a statement that "Isolation may also be demonstrated using other techniques."

Changes to Section 5.2.2.1 and 5.2.2.2 These sections were removed since they are well documented in the literature and add no value to the chapter. In the new Section 5.3, the criteria for using critical data and subcritical limits is included.

Changes to Section 5.2.2.3 This section was included in the new Section 5.3. Unlike Solid Angle and other historical techniques, the Substitution Methodology is unique and is documented in this section to provide the basis for its use.

Changes to Section 5.2.2.4 This section was not brought forward. As discussed earlier, the analyst must demonstrate the system to be subcritical under normal and credible abnormal conditions. The method used is chosen by the analyst and must be approved by the reviewer. Not listing the details of the methodology in the License does not reduce the safety of the system since the requirements to demonstrate subcriticality remain.

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Changes to Section 5.2.3

• This section was moved to the new Section 5.3.1. Part of the first paragraph was removed since it was an explanation of the difference in the Margin of Subcriticality and Margin of Safety. The second paragraph requiring the calculation of the Safety and Failure Limits for each controlled parameter was removed. As discussed above, it was established to demonstrate the sensitivity of a parameter was understood. Calculation of the Failure Limit for a parameter has little to no value with the implementation of the ISA since no credible upset can result in a critical condition and the risk of a criticality must be highly unlikely. The paragraph explaining why LEU systems can have c;1 smaller MoS was removed. It was explanatory in nature and does not add impact the effectiveness of the License.

• The normal (LCO) kett limits and corresponding MoS were removed since they were legacy items which does not provide actual safety. Both the normal and credible upset conditions must be demonstrated as subcritical, and the Margin of Safety and the Margin of Subcriticality do not correlate. A lower kett does not indicate a system is "safer." This is demonstrated by the following example: '

System A has a kett of 0.5. It is a sphere of HEU metal hanging by a string over a large water tank. System B has a kett of 0.99. It is an infinite lattice of natural enrichment fuel rods optimally spaced in water. In System A, if the string fails and the metal sphere falls into the water, the system is supercritical. In System B, nothing can be done with the rods to make them critical since it is an infinite array and is optimally spaced. System A has a lower normal kett, but is less safe than System 8.

Changes to Section 5.2.4

\

This section was moved to the new section 5.3.1. The text was edited for:clarity and the criteria for use of data and limits from handbooks and standards was clarified. Critical data from experiments, handbooks and technical publications may be used subject to the listed percentage limits. These values do not have to be demonstrated to meet the kett limits in the chapter since the method is based on margin in physical parameters and not in eigenvalues.

Subcritical limits may be used as stated subject to the constraints of the originating document (e.g. standards).

Changes to Section 5.2.5

• The safety factors for single units are described in the new Section 5.3. The section on analyzing systems with H/X or moderation upsets has been deleted. This section discussed how to analyze a system. The approach in the revised chapter is to limit discussion of how 1

to perform an analysis and focus on what the analysis must demonstrate, namely that the system is subcritical under normal and credible abnormal conditions and that criticality is highly unlikely. Removing this text does not change this basic commitment.

• The paragraph on the philosophy of the NCS postings and integrating NCS controls into the process design was deleted. The hierarchy of NCS controls is discussed in the new Section 5.2. Additionally, this section states that "The administrative limits and controls are provided to the operating areas on Nuclear Criticality Safety postings or in operating procedures or both."

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• The paragraph discussing reliance on structural integrity was deleted. The paragraph predates Subpart H and is inherent in the ISA methodology. The discussion of how the controls are maintained are documented as management measures.

• The two paragraphs on neutron poison were deleted. The paragraph predates Subpart H.

If the NCS analysis requires the use of fixed neutron poison, it is documented in the ISA along with the appropriate management measures ..

LTC Limits

• The limits from Section 5.2,6, 5.2. 7, and 5.2.8 were retained and moved to the new Section 1

5.7. , ,*

• Section 5.2.6 was moved to the new Section 5.7.5.

• Section 5.2.7 was moved to the new Section 5.7.6. Section 5.2.7.5 was modified to change,

" ... evaluation to be subcritical by at least 5% (k-effective < 0.95" to" ... Nuclear Criticality Safety evaluation fo be safe under specific conditions of disassembly". This is in the new S,ection 5.2.7.4.

• Section 5.2.8 was moved to the new Section 5.7.1.

• Section 5.2.9.1 was moved to 5.7.2.

• Section 5.2.10 was moved to the new Section 5.7.3.

• Section 5.2.12.1 and 5.2.12.2 were combined into the new Section 5. 7.4.

• Section 5.2.11.1 was moved to the new Section 5.2.6.11 *

• Section 5.2.11.2 was deleted since it is covered in the new Section 5.7.5.6.

• Section 5.2.11.3 was deleted since it is covered by the Certificates of Compliance (CoC's) for shipping containers.

• Section 5.2.11.4 was deleted since it is redundant. Commitments akeady exist to handle all SNM according to approved procedures. *

• Section 5.2.12.3 was deleted since it is covered by the CoC's for shi~ping containers.

Changes to the Appendix to Chapter 5

• The Appendix was written prior to Subpart H. The introduction and general discussion were not retained. The means and methods were simplified and included in the new Section 5.2.

The discussion of the Safety Limit, Limiting Condition of Operation, and the Routine Operating Limit were moved to the new Section 5.2. The Failure Limit definition was removed, as discussed earlier, and replaced with a statement that "A system is considered to be critical at the point when the kett value is 1."

• Section V of the Appendix was not retained. This section is inherent in the ISA methodology arid the establishment of IROFS and their supporting management measures.

Removal of License Conditions S-3 and S-4 Redacted License Conditions S-3 and S-4 are:

S-:-3 The volume in the Vault shall be no larger than shall be specifically shown to be critically safe by the licensee.

S-4 In no more than may be in transit ,within each cubicle at any one time.

These conditions were imposed as part of the approval of a license amendment to construct a new storage vault in 1989. These two conditions are legacy items which predate Subpart H.

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Conditions S-3 and S-4 are redundant since under Subpart H, the normal and upset conditions must be shown to be subcritical and it must be demonstrated that criticality is highly unlikely.

They can be removed without decreasing the effectiveness of the commitments in the license nor the safety of the operations.

Incorporation of License Condition S-11 Redacted License Condition S-11 states:

Systems involving clusters shall be deemed to include only workstations containing one or more machined and assembled clusters by themselves or in conjunction with other components that are not clusters. This shall apply to clad operations only.

This condition has been included in Section 5.3.1.3 as:

I The workstations shall be in an area where clusters or subcomponents of clusters are handled.

The wording covers the intent of the first sentence, but the second sentence was deleted. The second sentence was deleted since the Sectioning Facility which is attached to the Recovery area and is an unclad area could have clusters. The Sectioning Facility is used for destructive evaluation of preassemblies, subassemblies and clusters. This is an area where clusters are normally handled so it is covered by the commitment. When the condition was established, the NRC reviewer was concerned that a cluster could be modeled in an unclad area, like Recovery, to mask a higher kett condition in an unclad area .. The commitment as stated in 5.3.1.3 precludes this since clusters are not handled in this area.

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ENCLOSURE 2 Revision of SNM-42 License Application, Chapter 5 - Nuclear Criticality Safety P.O. Box 785 Lynchburg, VA 24505 USA t:

• 1.434.522.6000 f:

• 1.434.522.6805 www.bwxt.com POWERING TRAN SFO RMATI ON

SNM-42 CHAPTERS NUCLEAR CRITICALITY SAFETY

CHAPTERS NUCLEAR CRITICALITY SAFETY TABLE OF CONTENTS Chapter Title Page 5.1 Nuclear Criticality Safety Program 5-1 5.2 Nuclear Criticality Safety Control Philosophy 5-2 5.3 Nuclear Criticality Safety Analysis Techniques 5-4 5.3.1 Neutron Transport Computer Codes for Calculation ofkeff 5-6 5.3.1.1 Neutron Transport Computer Codes for Calculation ofkeff 5-7 5 .3 .1.2 Validation Applicability 5-8 5.3 .1.3 Margin of Subcriticality 5-8 5.4 Nuclear Criticality Safety Audits and Inspections 5-9 5.5 Nuclear Criticality Safety Training 5-9 5.5.1 General Employee Safety Training 5-9 5.5.2 Specialized Criticality Safety Training 5-10 5.5.3 Evaluation of Training 5-10 5.6 Criticality Accident Alarm System 5-10 5.7 LTC Specific Requirements 5-11

5. 7 .1 Mixed Uranium and Plutonium Limits at LTC 5-11

5. 7 .2 Hot Cells in Building B 5-11 5.7.3 Storage Tubes Inside LTC Buildings 5-11

5. 7.4 Outside Storage at LTC 5-11 5.7.5 Unirradiated Commercial (PWR and BWR) Fuel Assemblies at LTC 5-12 5.7.6 Irradiated Commercial (PWR and BWR) Fuel Assemblies at LTC 5-13 Date: xx/xx/xx Page: 5-i

5.1 Nuclear Criticality Safety Program NOG-L management has overall responsibility for the safety of all operations involving SNM. NOG-L management is committed to implementing a Nuclear Criticality Safety (NCS) program which incorporates the following objectives:

1. preventing an inadvertent nuclear criticality,

2. complying with the Double Contingency Principle as stated in ANSI/ANS-8.1-2014,

3. complying with the Process Analysis requirement of ANSI/ANS-8.1-2014,

4. complying with the NCS performance requirements of 10 CFR 70.61.

NOG-L management has established a Nuclear Criticality Safety program to implement these objectives. The program is led by the manager of Nuclear Criticality Safety who meets the requirements stated in Chapter 2. The Nuclear Criticality Safety manager has the overall authority and responsibility for the implementation of the Nuclear Criticality Safety program for the site. The Nuclear Criticality Safety manager is responsible for:

1. maintaining computational methods and practices,

2. determining the need for Nuclear Criticality Safety evaluations,

3. performing evaluations, and ensuring NCS controls are properly implemented,

4. maintaining Nuclear Criticality Safety inspection and audit programs,

5. training of the NCS staff to perform their duties,

6. overseeing of the Specialized Nuclear Criticality Safety Training Program and the NCS portion of the General Employee Safety Training The NCS manager has the authority to terminate any operation deemed to be unsafe or contrary to license conditions or good safety practices.

The responsibilities of the Nuclear Criticality Safety manager do not relieve area management of their responsibility for ensuring that operations are conducted in compliance with Nuclear Criticality Safety requirements. Decisions of the Nuclear Criticality Safety manager are not to be compromised by direct pressures of time or production.

The NCS staff positions and minimum qualifications are provided in Chapter 2. NCS analyses shall be performed by a qualified NCS staff member and peer reviewed by an appropriately qualified NCS staff member. All NCS analyses shall be documented.

5.2 Nuclear Criticality Safety Control Philosophy The risk of a criticality is minimized by controlling parameters within specified limits.

There are twelve parameters of criticality control. These controlled parameters are defined as follows:

Date: xx/xx/xx Page: 5-1

1. Favorable Geometry Control uses limiting dimensions of a piece of equipment or SNM arrangement to increasing neutron leakage by limiting dimensions of a piece of equipment or fuel arrangement.

2. Spacing Control decreases the neutron interaction by separating SNM.

3. Volume Control uses fixed volumes to limit the amount of SNM present and increase leakage.

4. Fixed Neutron Absorber Control uses a solid neutron absorber (poison) to reduce neutron multiplication.

5. Piece Count Control limits fuel mass and/or geometry by controlling the number of containers or components with known amounts of SNM and/or fixed geometries.

6. Mass Control limits the amount of SNM at a given location.

7. Moderation Control limits or excludes either interstitial (mixed with the SNM) or interspersed (between SNM units) moderating materials.

8. Concentration Control limits the mass ofSNM per unit volume.

9. Material Specification Control is a control based on consideration of the physical or chemical composition of material (e.g. metal, nitrate) and may also include the shape and composition of other materials (e.g. ATR element).

10. Uranium Enrichment Control utilizes the inherent differences in critical attributes (critical dimensions, mass, etc.) of uranium at different enrichments of 235 U.

11. Soluble Neutron Absorber Control uses a neutron absorber in soluble form to increase neutron absorption in a system.

12. Reflector Control limits the reflection of neutrons to a SNM system from adjacent materials.

The bounding assumptions for controlled parameters may take credit for physical properties and behaviors, experimental data, or historical operational data. Historical operational data may be used to establish the parameter range with appropriate consideration of the data applicability. Parameters that are not controlled shall be considered at their most reactive, credible values.

Three parameter limits are used in NCS analyses. These limits are defined below:

1. The Safety Limit (SL) is the value of the controlled parameter that will not be exceeded unless more than one unlikely, independent, and concurrent change in process conditions (contingency) has occurred.

2. The Limiting Condition of Operation (LCO) is the value of the controlled parameter that will not be exceeded unless a contingency has occurred.

3. The Routine Operating Limit (ROL) is the implementing value that is the same or more restrictive than the LCO and helps ensure that a violation of the LCO is unlikely.

Date: xx/xx/xx Page: 5-2

The margins of subcriticality for the Safety Limit and the Limiting Condition of Operation are provided in Table 5.2. The Routine Operating Limit does not have an associated margin of subcriticality.

Items Relied On For Safety (IROFS) are defined and implemented to maintain each identified controlled parameter within its specified Limiting Condition of Operation, thereby ensuring the subcriticality of a system or process. The four types of controls, listed in order of preference, are defined as follows:

1. Passive Engineered controls use fixed design features or devices that take advantage of natural forces such as gravity, ambient pressure, etc. No human intervention is required except for maintenance and inspection.

2. Active Engineered controls use add-on, active hardware (e.g., electrical, mechanical, hydraulic, etc.) to sense parameters and automatically secure the system to a safe condition. No human intervention is required during operation.

3. Enhanced Administrative controls rely on human judgment, training, and personal responsibility for implementation and are augmented by warning devices (visual or audible) which require human action according to a procedure.

4. Simple Administrative controls rely on human judgment, training, and personal responsibility for implementation each time the control function is needed.

The availability and reliability of the IROFS established for criticality safety are assured through management measures which are described in Chapter 11. The characteristics of IROFS are described in Chapter 3.

Activities at the site involving special nuclear material are conducted according to the limits and controls established by Nuclear Criticality Safety. The administrative limits and controls are provided to the operating areas on Nuclear Criticality Safety postings or in operating procedures or both.

Nuclear Criticality Safety postings describe the administrative limits and controls for a particular area, operation, work station, or storage location as appropriate to provide workers a ready reference for verifying compliance and safe operation.

Nuclear Criticality Safety postings generally include the following information as a minimum:

• Type of material permitted.

• Form of material.

• Allowable quantity (number of containers, pieces, weight, or volume).

• Spacing of fuel units, if required.

• Restriction on the presence of moderators, if required.

Date: xx/xx/xx Page: 5-3

5.3 Nuclear Criticality Safety Analysis Techniques The safety of operations can be demonstrated using many different methods including

• subcritical values in standards endorsed in Regulatory Guide 3.71, Revision 3,

• subcritical or critical values, from widely accepted industry handbooks, experimental data, or peer-reviewed publications,

• use of industry-accepted hand-calculation methods subject to the limitations of those methods,

• and use of discrete ordinates and Monte Carlo computer codes to calculate the effective multiplication factor (kea-).

Subcritical values stated in standards or handbooks may be used as stated in the standard or handbook. Use of these subcritical values is subject to the constraints specified in the standard or handbook.

Critical experiment data or critical data from handbooks and publications may be used with the application of appropriate safety factors. The limits derived from the critical values shall be no greater than:

• 90% of the critical dimension for cylinder diameters,

• 85% of the critical dimension for slab thicknesses,

• 75% of the critical spherical volume,

• 45% of the critical mass if double batching is credible,

• 75% of the critical mass if double batching is not credible,

• 45% of the minimum critical concentration.

The subcritical values derived from safety factors as described above do not have to meet the margins of subcriticality specified in Table 5.2.

Historically a method referred to as "Law of Substitution" (now called Substitution Methodology) has been used to account for interaction between different items. This methodology is described below since it is not commonly used in industry.

Substitution Methodology The substitution methodology allows intermixing of different units without explicitly calculating the intermixed array. The methodology requires establishing the required horizontal edge-to-edge spacing necessary for each of the units in an infinite planar array under the worst credible level of interspersed moderation. The minimum allowed horizontal edge-to-edge spacing is 12 inches between fissile units, subject to the following:

Date: xx/xx/xx Page: 5-4

• The unit must be modeled with any associated non-controlled parameters at the optimum, credible conditions.

• Reflective boundary conditions must be placed on the X & Y faces, no closer than 6 inches :from the surface of the SNM unit (establishes minimum 12 inch edge-to-edge horizontal spacing requirement).

• The floor must be modeled as minimum 12-inch-thick concrete.

• A minimum of 1200 cm of air (or interspersed moderation) must be modeled above the floor, unless there is intervening material that could increase reactivity.

o In this case, the more restrictive of the 1200 cm or the actual material and height is used.

o If there are multiple levels above the unit, the interaction of the levels must be addressed by explicit vertical modeling OR the interaction must be bounded by use of a reflector such as thick concrete on the intervening face of the model.

• The normal (LCO) condition of the controlled parameters must meet the LCO keff limit, see Table 5.2.

• All credible interspersed moderation levels must be evaluated. This cannot result in a situation that exceeds the Safety Limit keff (see Table 5.2).

• The spacing between units is increased until the Safety Limit keff is met.

The process is repeated for other SNM units that will be intermixed. The required spacing between two different units is the larger of the two.

Neutron Isolation A unit containing fuel may be considered isolated from another unit if the separation (edge-to-edge of fuel) is greater than the larger of the following distances:

a. Twelve feet, or

b. The greatest distance across an orthographic projection of either array on a plane perpendicular to a line joining their centers.

Isolation may also be demonstrated using other techniques.

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5.3.1 Neutron Transport Computer Codes for Calculation ofkeff When using computer codes to establish NCS limits, the controlled parameters are varied to determine the sensitivity ofkeffto changes in the controlled parameter. The amount the controlled parameters are varied is based on the contingency or contingencies under evaluation.

NCS calculations are performed using validated computer codes, techniques and cross section data sets. The computer codes are run on configuration-:-controlled software and hardware. All such analyses are reviewed by an independent qualified staff member.

Validation of computer codes and cross section sets is done per ANSI/ANS-8.24-2017. A computer code will be initially verified then validated. Reverification of the computer code system will occur at least annually or after revision to the computer code system. The application of new computer codes or additional benchmark data will be reviewed and approved by a qualified NCS staff member.

5.3 .1.1 Calculational Margin and Margin of Subcriticality The calculational margin includes the allowances for the bias (keff minus the experimental keff value or the ratio of the calculational keff to the experimental ketr minus 1) and the bias uncertainty as well as uncertainties associated with interpolation, extrapolation and trending. The bias uncertainty accounts for the uncertainties in benchmarks, the calculational models and the calculational method. An acceptable keff is determined by:

1 - Af<MoS - Af<:c:M ~ kcalc(analysis) + 2 CTcalc(analysis)

(USL = 1 - Af<Mos - Af<:c:M)

The preferred form is:

1 - &MoS ~ kcalc(analysis) + 2 crcalc(analysis) + &cM where: &Mos is the Margin ofSubcriticality (MoS -listed in Table 5.2),

&cM is the Calculational Margin, kcalc(anaiysis) is the calculated ketr of a system being evaluated as part of Nuclear Criticality Safety analysis, crcalc(anlaysis) is the uncertainty on that calculation.

Techniques for Establishing the Calculational Margin:

The Non-Parametric Method (NPM) is primarily applied when the underlying distribution of the data is not known or cannot be verified. The confidence level that a fraction of the population is above an observed value is:

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where:

q is the desired population fraction (0.95) n is the number of benchmarks in the data set m is the rank order indexing from the smallest sample to the largest (m=l for the smallest sample, m=2 for the second smallest, etc.).

For the smallest value in the sample, the equation reduces to:

P= 1- q n = 1- 0.95 n If there are more than 58 benchmarks, a rank order (m) greater than 1 can be used provided the selected rank order yields a p equal to or greater than 0.95 which would assure at least a 95/95 confidence level.

Non-parametric methods are the preferred method to establish the calculational margin. The non-parametric approach used is based on the specified rank order calculated ketr value. This method has three terms that define the calculational margin (&cM).

&cM = jbiasj + CTcalc + Af<:NPM where: bias = kca1c-kexp , or kca1Jkexp - 1 kca1c used is the specified rank order calculated ketr of the benchmarks used forthe validation, kexp is the reported experimental ketr for the same configuration, kNPM is the margin accounting for the amount of experimental data.

Since no credit is taken for a positive bias, if the specified rank order calculated ketr of the benchmarks is greater than the experimental value, the bias is set to zero and the equation becomes:

Af<:cM = 0 + O'calc + Af<:NPM (for kexp~caic)

The non-parametric margin (&NPM) is based on the degree of confidence for 95% of the population and is shown below.

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Table 5.1 Degree of Confidences and Non-Parametric Margin Degree of Confidence Non-Parametric Number of for 95% of the Margin Experiments (n) population (13) (AkNPM)

>0.9 n>45 0.00

>0.8 32 <n <44 0.01

>0.7 24<n <31 0.02

>0.6 18 <n <23 0.03

>0.5 14<n<17 0.04

>0.4 10<n<13 0.05

<0.4 n<lO Insufficient data

~ - Percent confidence that a :fraction of the population is above the lowest point.

Other statistical methods such as Lower Tolerance Band (95/95 or greater) or Lower Tolerance Limit (95/95 or greater) may be used if the data meets the assumptions of the methodology. When methods that employ trending are used, trends may indicate ketI values greater than unity for some parameter ranges. In ranges where the trended ketivalue exceeds unity, additional margin shall be applied equal to the amount of the positive bias as a function of the trending parameter. For methods that use average values, it is possible to have average ketI's that exceed unity. In those cases, the additional margin shall be applied. The margin shall be the amount of the positive bias.

5 .3 .1.2 Validation Applicability:

The Validation Applicability includes:

1. Unclad fuel of different chemical forms in both solid and solution form,

2. Clad fuel components,

3. The full range of enrichments, HIX, and average energy, and

4. A large number of moderators and reflectors.

The Validation Applicability covers the operations existing in the facility.

If extensions to the Validation Applicability are required, the extensions will be consistent with the assumptions and limitations of the method used to establish the calculational margin or by application of additional margin, which must be justified.

5.3.1.3 Margin of Subcriticality:

The margin of subcriticality varies depending on the systems and the condition. Three categories for margins are:

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1. Low-enriched systems contain uranium enriched s 10 weight percent in 23su

2. High-enriched systems contain uranium enriched > 10 weight percent in 235 U and

3. Systems involving welded clusters in which the welded cluster is the reactivity driver of the system, must meet the following:

• Be fueled by high enriched uranium (>90 weight percent 235 U).

• Have a thermal neutron spectrum when full flooded.

• Be constructed of the same geometric style elements as those in the applicable critical experiments.

• Any significant absorbers must have been included in the applicable critical experiments.

• The workstations shall be in an area where clusters or subcomponents of clusters are handled.

A system is considered to be critical at the point when the keff value is 1. The Safety Limit and the Limiting Condition of Operation have keff which is lower than the critical value by the approved margin of subcriticality, also call the margin of subcriticality. The system keff limits and margin of subcriticality (Af<Mos) for the three different categories are listed in table below.

Table 5.2 Margin of Subcriticality for LCO and SL LCOandSL Type of System L1kMoS kecc Low-enriched systems 0.03 0.97 High-enriched systems 0.05 0.95 Systems involving welded clusters 0.025 0.975 5.4 Nuclear Criticality Safety Audits and Inspections N CS audits and inspections are defined in Chapter 11. 5 .1.

5.5 Nuclear Criticality Safety Training 5.5.1 General Employee Safety Training All individuals are given Nuclear Criticality Safety training prior to being granted unescorted access to the Restricted Areas as defined by 10 CFR 20. This includes, as a minimum, the following training:

• A discussion about the fission process and criticality.

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• A discussion about the fission process and criticality.

• A brief history of criticality accidents.

• The effects and consequences of a criticality accident at this plant.

• The importance of an immediate evacuation in case of a criticality accident.

This training shall be developed with the technical oversight of Nuclear Criticality Safety. This training is repeated annually.

5.5.2 Specialized Criticality Safety Training In addition to General Employee Safety Training, all employees who handle fissile materials are given specialized instruction annually. This program covers the general safety principles of handling fissile material and also covers the application of these principles by discussing examples of specific criticality safety limits. Specialized Nuclear Criticality Safety training shall be developed with the technical oversight of Nuclear Criticality Safety.

Specialized training is supplemented by on-the-job training and qualification of operators. This training specifically addresses the criticality safety limits contained in operating procedures and on postings for specific jobs.

5.5.3 Evaluation of Training The effectiveness of the Nuclear Criticality Safety training is judged by three methods.

First, written and/or oral tests are given to each individual who receives Specialized Nuclear Criticality Safety instruction; the test must be passed. Tests are not normally given following General Employee Safety Training.

Second, Nuclear Criticality Safety inspections of the entire plant reveal how well personnel understand the safety controls as demonstrated by of the number of Nuclear Criticality Safety violations found.

A third method of evaluating how well employees understand the safety requirements is the supervisor's close contact with the employee.

5.6 Criticality Accident Alarm System The site shall maintain a criticality accident alarm system for each area in which 700 grams or more of 235 U is possessed, 450 grams or more of plutonium, or 450 grams or more of any combination thereof. This monitoring system shall be capable of energizing a clearly audible alarm signal if an inadvertent criticality occurs. The placement of the detectors shall be determined by calculation utilizing detection criteria described in 10 CFR 70.24(a)(l),

and methodology described in Regulatory Guide 3.71, Nuclear Criticality Safety Standards for Fuels and Material Facilities, Revision 3, October 2018.

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Whenever the criticality monitoring system is out of service, in storm-watch mode, or being tested or repaired, compensatory measures shall be in place to ensure evacuation if a criticality occurs. Compensatory measures shall be specified in facility procedures, and periods when the criticality monitoring system is out of service should be minimized to the extent practical.

5. 7 LTC Specific Requirements LTC Building B is limited to 40 units, excluding the hot cells, underwater storage, in-ground storage tubes and the examination of commercial fuel assemblies. Each unit shall be separated by a minimum of 8 inch edge-to-edge and 24 inches center-to-center.

The limits are listed below:

5. 7 .1 Mixed Uranium and Plutonium Limits at LTC Fuel (other than fuel contained in irradiated fuel rods) containing or potentially containing uranium and plutonium will be handled based on units. Each unit will be limited to total fissile material based on the plutonium weight percentage as shown below:

233U (g) + 235U (g) 239pU (g) + 241pU (g) 350 + 220  ::;; l

5. 7.2 Hot Cells in Building B The hot cells shall be limited to three units, as defined in 5.7.1, in Hot Cell No. 1, provided the units are separated by a minimum of 12 inch edge-to-edge and one unit in each of the other hot cells.

5. 7 .3 Storage Tubes Inside LTC Buildings SNM in storage tubes shall be limited to the values specified in 5. 7 .1 for each tube.

Storage tubes shall be spaced a minimum of 17 inch center-to-center (except for one pair of tubes which may be spaced a minimum of 16.5 inches), are approximately 5 inches in diameter, and are totally immersed in concrete.

5. 7.4 Outside Storage at LTC Outside storage consists of underground storage, shipments and the fenced outside storage area located adjacent to Building J.

SNM in storage tubes shall be limited to the values specified in 5.7.l for each tube.

Tubes shall be spaced 20 inch center-to-center, are approximately 5 inches in diameter, and are totally immersed in concrete.

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5.7.5 Unirradiated Commercial (PWR and BWR) Fuel Assemblies at LTC Unirradiated fuel assemblies will be received at a maximum of two at a time in a shipping container licensed for two assemblies, or one assembly in a shipping container licensed for one assembly. Unirradiated fuel assemblies may be handled and stored subject to the following conditions:

1. Unirradiated fuel assemblies may be stored in air in the Cask Handling Area (CHA) or in the Development Test Area.

2. Assemblies may be stored in their shipping container as received.

3. Assemblies may be stored a minimum of21-inches apart surface-to-surface.

4. Assemblies may be stored under water in the CHA pool, Pool Test Facility pool, or Development Test Area pool in racks constructed to maintain a I-foot minimum surface-to-surface separation between assemblies and any other SNM.

Assemblies may be handled and dismantled under water subject to the same requirements of the irradiated fuel in the CHA Pool.

5. No more than four unirradiated assemblies may be kept at the LTC site at one time.

6. Only one unirradiated fuel assembly shall be dismantled or reassembled at a time in the Development Test Area. The dismantling operation shall meet the following:

• Only one fuel rod may be removed from or inserted into the assembly at a time.

• Only one fuel rod may be in transit to any location at a time.

• The fuel assembly may be completely disassembled by withdrawing one fuel rod at a time from the assembly; during all stages of disassembly, the partially disassembled assembly shall be maintained within the confines of the assembly whether damaged or undamaged.

7. Associated with the dismantling operation, one storage position will be permitted for fuel rods removed from an assembly provided that:

• The assembly and associated rod storage position shall be separated from each other and from any other fissile material by a minimum of 21 inches surface-to-surface.

• The associated rod storage position shall be no larger in any dimension than the fuel assembly. There shall be one storage position for each fuel assembly to be dismantled. Rods may be stored or handled in a slab up to 4 inches thick provided the slab is separated from other fissile material by a minimum of 12 feet.

• Only one fuel rod may be removed or inserted into the associated rod storage position at a time.

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8. Fuel assemblies to be studied shall meet the following:

• Each assembly shall be of the enriched PWR type with a 15 x 15, or 17 x 17 square pin lattice not greater than 8.6 inches on a side (further identified as Babcock & Wilcox Mark B or Mark C canless assemblies).

• The maximum initial enrichment in an unirradiated fuel assembly shall not exceed 4.05 wt%.

• Damaged fuel assemblies may be examined in air. Fuel assemblies which have been damaged can be examined in water if they maintain their 8.6 inch on a side dimension.

9. Other PWR or BWR fuel assemblies which do not meet the above listed requirements may be studied, provided:

• The unirradiated, fully reflected fuel assembly (fueled with U02 only) with all control rods removed is shown by an appropriate Nuclear Criticality Safety evaluation to be safe.

• The fuel assembly is shown by an appropriate safety evaluation to be safe under specific conditions of disassembly.

10. BWR fuel assemblies may be received and studied provided:

• They are evaluated pursuant to item 9 above, or

• The BWR fuel assemblies have a maximum initial unirradiated enrichment of 4.05 wt% 235 U and have a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.

11. After examination, fuel rods, including fuel rod segments may be placed in any available hole in a fuel assembly, including instrumented and control rod guide tube positions, i.e., 225 and 285 fuel rods in Mark B and Mark C assemblies, respectively. Fuel rod segments shall have their ends sealed, and shall be encapsulated in steel tubing with ends sealed, prior to insertion into an available hole in a fuel assembly 5.7.6 Irradiated Commercial (PWR and BWR) Fuel Assemblies at LTC Irradiated fuel assemblies will be received at a maximum of two at a time in a shipping container licensed for two assemblies, or one assembly in a shipping container licensed for one assembly.

1. Irradiated fuel assemblies will be stored in the CHA pool which is limited to the following conditions:

• A maximum of four fuel assemblies or portions thereof may be in the pool at a time.

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r

• The assemblies shall be stored in racks constructed to maintain a 1-foot minimum surface-to-surface separation between assemblies and any other SNM in storage or transit. Each position in the assembly storage rack must limit contained fuel to a square not to exceed the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.

• Partially dismantled assemblies will be stored in the assembly storage rack.

• Only one assembly may be in a designated work area of the pool at any one time. There shall be a minimum of 1-foot surface-to-surface separation between the assembly in the work area and any other fissile material.

2. Dismantling of irradiated fuel assemblies is permitted in the Pool under Hot Cell No. 1 provided:

• Only one fuel rod at a time shall be removed from or inserted into the fuel assembly

• A fuel assembly can be completely dismantled by withdrawing one fuel rod at a time from the assembly; during all stages of dismantlement, the partially dismantled assembly shall be maintained within the confines of a square not exceeding the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.

3. Associated with the dismantling operation, one storage position will be permitted for fuel rods or components removed from the assembly provided that:

• The assembly and associated rod storage position shall be separated from each other and from any other fissile material by a minimum of 1 foot surface-to-surface.

• Fissile material and fuel rods or components in the associated storage positions shall be restricted to a square not exceeding the dimensions of a fresh fuel assembly or to a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.

• Only one fuel rod may be inserted or removed from the storage position at a time.

• A maximum of75 fuel rods shall be permitted in the rod storage position.

4. Fuel assemblies to be studied shall meet the following:

• Each assembly shall be of the enriched PWR type with a 15 x 15, or 17 x 17 square pin lattice not greater than 8.6 inches on a side (further identified as Babcock & Wilcox Mark B or Mark C canless assemblies).

• The maximum initial enrichment in an unirradiated fuel assembly shall not exceed 4.05 wt%.

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5. Other PWR or BWR fuel assemblies which do not meet the above listed requirements may be studied, provided:

• The unirradiated, fully reflected fuel assembly (fueled with U02 only) with all control rods removed is shown by an appropriate Nuclear Criticality Safety to be safe.

• The fuel assembly is shown by an appropriate Nuclear Criticality Safety evaluation to be safe under specific conditions of disassembly.

6. BWR fuel assemblies may be received and studied provided:

• They are evaluated pursuant item 5 above, or

• The BWR fuel assemblies have a maximum initial unirradiated enrichment of 4.05 wt% 235U and have a cross sectional area not exceeding that of a 22.5 cm (8.85 in.) diameter cylinder.

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