ML013390222

Standard Format for Operating License Amendment Requests

ML013390222
Person / Time
Site:	Nuclear Energy Institute
Issue date:	08/24/2001
From:	Hendricks L Nuclear Energy Institute
To:	Office of Nuclear Reactor Regulation, Nuclear Energy Institute
References
Download: ML013390222 (27)

Text

0 NEI NUCLEAR ENERGY INSTITUTE Lynnette Hendricks DIRECTOR, LICENSING NUCLEAR GENERATION August 24, 2001 TO: Administrative Points of Contact

SUBJECT:

Standard Format for Operating License Amendment Requests The final version of the NEI white paper entitled "Standard Format for Operating License Amendment Requests from Commercial Reactor Licensees" is enclosed for your information and voluntary use. A copy of the paper will be forwarded to the NRC for information.

If you have questions or comments, please contact Mike Schoppman at 202-739-8011 (mas@nei.org).

Sincerely, Lynnette Hendricks Enclosure c: Suzanne C. Black, U.S. Nuclear Regulatory Commission (NRR/DSSA)

NEI Licensing Action Task Force i

[Licensee's letterhead]

[Date] U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

[Plant / Unit Name(s)] Docket No(s) [50-., 50-__J [Title], [Brief description

& affected Tech Spec(s), if any]2

REFERENCES:

[as necessary]

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, [license holder] hereby requests the following amendment:

[brief summary of the proposed amendment and the results of the corresponding "no significant hazards determination"].

[License holder] requests approval of the proposed amendment by [date + justification].3 Once approved, the amendment shall be implemented within [ ] days.' [Include or attach a listing of formal licensee commitments that would derive from NRC's approval of the proposed amendment.]

[If the proposed amendment is "risk informed," include a statement that the guidance in Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," has been followed.]

I The title used by many licensees is "License Amendment Request (LAR)." Other licensees use "Proposed License Amendment (PLA)." These and other equivalent terms are acceptable titles. 2 Usually, but not always, the proposed amendment is intended to revise particular Technical Specifications.

However, there will be some cases that result in a license amendment, but do not change any Technical Specification (for example, an amendment to incorporate a change to the updated FSAR or Technical Specification Bases that requires prior NRC approval in accordance with 10 CFR 50.59 (c)(2)).

I Provide justification in the cover letter for the "need date." For example, approval by that date is necessary to prepare for the scheduled startup following a refueling outage. 4 A 60-120 day implementation period is typical. If additional implementation time is needed, provide justification in the cover letter, e.g., significant procedure changes are required to support implementation.

i NEI White Paper August 2001 STANDARD FORMAT FOR OPERATING LICENSE AMENDMENT REQUESTS FROM COMMERCIAL REACTOR LICENSEES This paper outlines a standardized format that licensees may use on a voluntary basis to prepare a proposed plant-specific amendment to the Operating License for a commercial reactor. Italicized information in brackets represents amendment specific information that should be provided by the licensee.

This paper uses footnotes to explain certain concepts.

Thus, they are part of the White Paper, not part of the LAR format. The White Paper includes:

-Licensee's cover letter to NRC -Licensee's standard affidavit, or standard language from 28 USC 1746 -Licensee's evaluation of the proposed amendment (including attachments)

-An Appendix that provides a sample LAR (for illustration)

If you have any questions or require additional information, please contact [Mr. IMrs. IMs., licensee's point of contact for the NRC Office of Nuclear Reactor Regulation]

at [telephone number].

Sincerely, [Signature]

[Name] [Title]

Enclosures:

-Notarized affidavits

-Licensee's evaluation of the proposed change(s)Attachments:

1. Proposed Technical Specification Changes (mark-up)

2. [Proposed Technical Specification pages (retyped)

'- OPTIONAL]

3. List of Regulatory Commitments

4. [Changes to TS Bases pages -OPTIONAL]cc: [Region -I [NRR Project Manager] [Plant/ Unit Resident Inspector(s)]

[State contact] SAs an alternative to a notarized "oath & affirmation" affidavit, a licensee may substitute the following statement in the cover letter: "1 declare under penalty of perjury that the foregoing is true and correct. Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.

6 This attachment is OPTIONAL.

Many licensees defer submittal of retyped Technical Specification pages until the end of the process, thereby permitting the incorporation of any revisions derived from responses to NRC Requests for Additional Information or other sources.

a AFFIDAVIT

[Licensee's standard affidavit.

This attachment is not required if the licensee chooses to use the standard language from 28 USC 1746 in the cover letter -see footnote 5.]

LICENSEE'S EVALUATION

Subject:

[Brief title. Identify which Technical Specification section(s) will be changed.]

1. DESCRIPTION

2. PROPOSED CHANGE 3. BACKGROUND

4. TECHNICAL ANALYSIS 5. REGULATORY SAFETY ANALYSIS 1.1 No Significant Hazards Consideration 1.2 Applicable Regulatory Requirements/Criteria

6. ENVIRONMENTAL CONSIDERATION

7. REFERENCES i

0

1.0 DESCRIPTION

This letter is a request to amend Operating License(s)

[license number(s)]

for [plant / unit name(s)].

The proposed change(s) would revise the Operating License(s) to [provide a brief general description of the proposed amendment, the reason for the amendment, and any timing constraints

-reserve details for BACKGROUND, Section 3.0.]' 2.0 PROPOSED CHANGE [Prot..ide a detailed description of individual changes addressed within the proposed amendment.

The degree of detail should be consistent with the degree of complexity of the proposed amendment.]

in summary, [provide a paragraph containing a few descriptive sentences suitable for use by NRC in the Federal Register notice that will be published to seek public comment on the proposed amendment.

Avoid slang words or undefined abbreviations or acronyms.

This summary may duplicate wording in the licensee's cover letter and should bound the detailed changes being proposed.]

[Describe associated changes to the Technical Specification Bases, if any.] 3.0 BACKGROUND

[Include:

"* System description(s).

"* Applicable references to updated FSAR text and figures.

"* Discussion of conditions that the proposed amendment is intended to resolve.

"° An explanation of the circumstances that establish a need for the proposed amendment(s), for example, historical information, prior communication or correspondence with NRC staff, relevant reference documents, etc.] 7 In some cases, the amendment will affect only the Operating License, for example, when a license amendment is necessary in accordance with 10 CFR 50.59(c)(2).

However, in most cases, the amendment will affect one or more Technical Specifications.

4.0 TECHNICAL ANALYSIS [Include:

"* A detailed explanation of the proposed amendment.

"* A detailed description of analytical methods, applicable standards, data, and results.

"* Technical details in support of safety arguments.

"* The impact on updated FSAR accident analyses.

"* Briefly summarize the preceding arguments at the end of this section.]

[The Technical Analysis section should be written such that it may be used with minimal modification in the NRC staffs Safety Evaluation (SE).] [If the proposed amendment is risk-informed, include information in accordance with the Regulatory Guide series 1.174 -1.178 on "risk-informed decision-making." These five Regulatory Guides address plant-specific changes to the licensing basis, inservice testing, graded Quality Assurance, Technical Specifications, and inservice inspection, respectively).]

5.0 REGULATORY SAFETY ANALYSIS 5.1 No Significant Hazards Consideration

[Licensee name] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response:

No. [Explanation/

basis for the response.]

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response:

No.

I[Explanation

/ basis for the response.]

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety? Response:

No. [Explanation/

basis for the response.]

Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based on the above, [licinsec ,iame] concludes that the proposed amendment(s) present no si.gnificant hazards consideration under the standards set forth in '10 CI'1 R 50.92(c), and, accordingly, a finding of"no significant hazards c!-nsidcration" is justified.

5.2 Applicable Regulatory Requirements/Criteria

[This section describes in detail how the licensee's technical analysis, which may or may not include risk information, satisfies all applicable regulatory requirements and criteria.

Any formal commitments to administrative controls needed to ensure compliance should be included in this section. The Regulatory Analysis provides a basis that the NRC staff may use to find the proposed amendment acceptable.

It should be written such that it may be used with minimal modification in the NRC staffs Safety Evaluation (SE).] [To assist the NRC staff, the licensee may choose to include an optional table of applicable regulatory requirements

/criteria.]

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 6.0 ENVIRONMENTAL CONSIDERATION

[The identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review is the subject of 10 CFR 51.22. The categories of actions deemed "categorical exclusions" are specified by 10 CFR 51.22(c).

The licensee's consideration of environmental factors should include sufficient detail to support a finding of categorical exclusion.

For -the majority of I changes, it is clear that the environment will not be affected (e.g., extending a surveillance interval).

Therefore, a simple statement (see below) is sufficient.

If appropriate, the licensee can provide more detailed information to strengthen the justification of categorical exclusion.]

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cvanulative occupational radiation exposure.

Accordingly, the proposed amen, .ient meets the eligibility criterion for categorical exclusion set forth in 10 CFR ýE.22(c)(9).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statenr ent or environmental assessment need be prepared in connection with the proposec amendment.

[and/or] The proposed amendment is confined to (i) changes to surety, insurance, and/or indemnity requirements, or (ii) changes to recordkeeping, reporting, or administrative procedures or requirements.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(10).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

[Identify and number all references used to prepare the proposed amendment.

Each reference should be cited at least once in this attachment (Licensee's Evaluation).

If a reference is needed to understand, review, or approve the proposed amendment, it should be considered for inclusion as an enclosure and identified with a suitable enclosure number or letter.] [If precedent can be identified, the licensee should reference the affected power plant(s) and amendment number(s), and briefly discuss how the precedent applies to the specific circumstances of the proposed amendment.

If there are any differences between identified precedent and the proposed amendment, the licensee should explain the differences and describe their impact on the acceptability of the proposed amendment.

Precedent, by itself, does not demonstrate the acceptability of a proposed amendment, but it does give the NRC staff information about how they have treated similar changes in the past. This may simplify the NRC staffs review.]

0 Attachment 1 Proposed Technical Specification Changes (mark-up) 0 Attachment 2 Attachment 2 (OPTIONAL)

[Proposed Technical Specification Pages (retyped)]

I I Attachment 3 List of Regulatory Commitments The following table identifies those actions committed to by [Licensee]

in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Please direct questions regarding these commitments to [name of licensee c&ntact].i REGULATORY COMMITMENT DUE DATE i i at Attachment 4 [Changes to Technical Specification Bases Pages -OPTIONAL]

APPENDIX A SAMPLE LICENSE AMENDMENT REQUEST (LAR)t a

Subject:

Application For Amendment To Technical Specification (TS) 3.9.4, "Containment Penetrations," to Allow Open Penetrations During Refueling Operations if Appropriate Administrative Controls are Established

1.0 DESCRIPTION

The proposed amendment would revise the [LICENSEE]

plant's Technical Specifications (TS) for Limiting Condition for Operation (LCO) 3.9.4, "Containment Penetrations," to allow all containment penetrations that provide direct access from the containment atmosphere to the outside atmosphere to be open during refueling activities if appropriate administrative controls are established.

The current TS causes some outage activities to be interrupted while fuel handling activities or core alterations are in progress.

The proposed revision would allow for more efficient performance of outage work while continuing to provide an acceptable barrier against the release of fission product radioactivity to the outside atmosphere during core alterations or fuel handling activities inside containment.

The proposed changes are consistent with NRC-approved Industry/Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-312, Revision 1 (Reference 1). Surveillance Requirement (SR) 3.9.4.1 is also revised to delete the exception for containment penetrations P-63 and P-98, because valve-specific exceptions would not be necessary.

2.0 PROPOSED CHANGE Specifically the proposed changes would revise the following:

2.1 TS 3.9.4(c)-NOTE The NOTE in Limiting Condition for Operation (LCO) 3.9.4, Item (c), is revised to allow un-isolating containment penetration flow path(s) under administrative controls during operations involving core alterations or fuel movement inside containment.

2.2 SR 3.9.4.1 Surveillance Requirement (SR) 3.9.4.1 is revised to delete the exception for containment penetrations P-63 and P-98 based on the revision to the NOTE (item 2.1 above). 3.0 BACKGROUND

[Include system descriptions and updated final safety analysis report (UFSAR) references, as needed.] Current TS 3.9.4 permits opening only containment penetrations P-63 (service air valves KA V-039 and KA V-1i18) and P-98 (breathing air valves KB V-001 and KB V-002) under administrative controls during operations involving core alterations or fuel movement inside containment.

Amendment No. 107 (Reference

2) approved the use of administrative controls for these penetrations based on the finding that direct access could be established through these penetrations if the associated air compressor was turned off or failed with its respective containment isolation valves open. The proposed change would allow any containment penetration flow path that provides direct access from the containment atmosphere to the outside atmosphere to be open during operations involving core alterations or fuel movement inside containment if appropriate administrative controls are established and maintained.

Note that TS 3.6.3, "Containment Isolation Valves," currently has a provision similar to the one proposed here for temporarily opening containment penetration flow paths in MODES 1 through 4 under administrative controls.

Administrative controls ensure that, in the event of a fuel handling accident (FHA) inside containment, open penetrations will be promptly closed. Such administrative controls include the awareness of the temporary flow path conditions and the designation of individuals to isolate the flow paths in the event of an FHA. The proposed change is consistent with NRC approved Industry/TSTF Standard TS Change Traveler, TSTF-312, Revision 1. Justification for TSTF 312 is that (1) the dose consequences for the design basis FHA indicate acceptable radiological consequences, and (2) the licensee will implement administrative procedures that ensure open containment penetrations can and will be promptly closed in the event of a fuel handling accident (FHA). The [LICENSEE]

company's dose calculations will document the time to close the penetrations.

4.0 TECHNICAL ANALYSIS During core alterations or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity to the environment due to a FHA inside containment will be restricted to required limits when the TS requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in TS 3.6.1, "Containment." The requirements of TS 3.9.4, "Containment Penetrations," ensure that the consequences of a postulated FHA inside containment during core alterations or fuel handling activities remain within acceptable limits. The LCO establishes containment penetration closure requirements, which limit the potential escape paths for fission products by ensuring that there is at least one integral barrier to the release of radioactive material.

LCO 3.9.4 requires a minimum of one door in the emergency airlock to be closed and at least one door in the personnel airlock to be capable of being closed and requires that each penetration providing direct access from the containment atmosphere to the outside atmosphere either be closed by a manual or automatic isolation valve, blind flange, or "equivalent." As discussed in the TS Bases, Background section for TS 3.9.4, "equivalent" isolation methods must be approved and may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for containment penetrations during fuel movement.

The proposed TS change to allow the containment penetration flow path(s) to remain open while using administrative controls fully implements NRC approved TS traveler TSTF-312, Revision 1. Further, this approach is consistent with the administrative controls currently allowed by [LICENSEE]TS for more restrictive, higher operational modes. Current provisions in TS 3.6.3, "Containment Isolation Valves," allow penetration flow paths to be un-isolated I

1 under administrative controls in MODES I through 4. The controls include a designated operator having continuous communication with the Control Room who can isolate the open valve in the event of an accident.

This allowance has been determined to be an acceptable means to allow the opening of flow paths in consideration of the administrative controls that minimize the impact of an accident.

These modes are more significant than during refueling operations due to the RCS energy and potential to provide a significant motive force for the expulsion of radionuclides, subsequent to a design basis accident.

A similar allowance is acceptable for penetrations that are open during fuel movement or core alterations provided appropriate administrative controls are utilized.

During core alterations or fuel movement activities inside containment, the potential for a FHA resulting in containment pressurization is negligible since the reactor is shutdown.

Therefore, allowing penetration flow path(s) that have direct access from the containment atmosphere to the outside atmosphere to be unisolated is acceptable during refueling operations provided appropriate administrative controls are used. These proposed controls include an awareness of the open penetration and designation of individual(s) readily available for closing the open penetration in the event of a FHA inside containment.

The [LICENSEE]

design basis FHA is defined as the dropping of a spent fuel assembly onto the spent fuel pool fuel storage area or inside containment.

Both analyses assume the rupture of the cladding of all the fuel rods in the assembly.

These FHA events are postulated for safety system design purposes even though many administrative controls and physical limitations are imposed on fuel handling operations.

Section 15.7.4 of the UFSAR (Reference

3) discusses the consequences of a postulated FHA inside containment.

The results from the current analysis of a FHA indicate an exclusion area boundary thyroid dose of 64.1 REM and whole body dose of 0.177 REM. These results are well within the 10 CFR 100 (Reference

4) offsite dose limits of 300 REM and 25 REM, respectively, and less than the guideline values of Standard Review Plan, Section 15.7.4, Revision 1 (Reference 5). The FHA analysis presented in the USAR considered dropping a single irradiated fuel assembly in either the containment or fuel building.

The postulated FHA inside containment with the personnel airlock doors remaining open, remains the limiting FHA when considering open penetration flow paths. In the event of a FHA inside containment with open penetrations, transmission of radionuclides to the outside environment is unlikely.

This is because the dispersion of radioactive material through the containment will not be driven by any pressure differential resulting from the accident, but only due to the course of containment air circulation.

The administrative controls for prompt closure of the containment penetration flow paths would minimize the potential spread of radioactive isotopes from the containment to the outside environment.

Therefore, following a FHA inside containment, the lack of containment pressurization provides sufficient time to manually isolate the penetration flow paths to minimize dose consequences.

The consequences of a FHA inside containment with open penetration flow paths are bounded by the current analysis described in the USAR. This ensures that offsite dose is well within the 10 CFR 100 regulatory limits and less than the guideline values of Standard Review Plan, Section 15.7.4, Revision 1. Amendment No. 95 (Reference

6) approved leaving the containment air lock open during fuel movement and core alteration.

In that application, [LICENSEE]

recalculated the doses and revised the design basis for the FHA analysis to be consistent with Regulatory Guide 1.25 (Reference 7). In that re-analysis, credit was not taken for the containment building barriers.

The analysis calculated the doses for the 0-2 hour period at the exclusion area boundary.

The calculated doses were within the Standard Review Plan criteria of 6 REM to the whole body and 75 REM to the thyroid..

As discussed in Amendment No. 107, the potential dose consequences from a simultaneous release of the gaseous effluents through an unisolated penetration flow path and the open personnel airlock doors is the same. That is because the analysis assumes all radioactive material from the FHA is released to the environment within a two-hour period. Therefore, allowing penetration flow paths to be unisolated during core alterations or movement of irradiated fuel will not invalidate the conclusion that the potential dose consequences from a FHA will be well within the 10 CFR Part 100 limits. The NRC has historically required technical specifications to maintain containment closure during core alterations and fuel handling as a defense-in-depth measure to further limit releases.

This has been relaxed by allowing both doors to the containment personnel air lock and selected containment penetrations to be open during core alterations and fuel handling with provisions in place to quickly close one door or isolate the penetration (References 2 and 6). [LICENSEE]

proposes to extend the same provisions for penetration flow paths open during core alterations and fuel handling.

These procedural controls include: 1. Appropriate personnel will maintain an awareness of the open status of the penetration flow path during core alterations and movement of irradiated fuel assemblies within containment.

2. Specified individuals will be designated and readily available to promptly isolate open penetration flow paths in the event of a FHA inside containment.

Based on the analysis of the FHA and the administrative controls specified for the proposed allowance to unisolate containment penetration flow paths, the proposed changes are acceptable.

With respect to the proposed administrative controls, the proposed License amendment provides assurance that offsite dose levels associated with a FHA inside containment will be maintained well within the applicable regulatory limits of 10 CFR 100. 5.0 Regulatory Safety Analysis 5.1 No Significant Hazards Consideration

[LICENSEE]

has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR50.92(c) as discussed below: [provide No Significant Hazards Determination]

I 5.2 Applicable Regulatory Requirements/Criteria In the following paragraphs applicable regulatory criteria and acceptance limits as they are related to the proposed changes are discussed.

A summary of this discussion is provided in the table below.

0 5.2.1 Regulations The regulatory basis for TS 3.9.4, "Containment Penetrations," is to ensure that the primary containment is capable of containing fission product radioactivity that may be released from the reactor core following a FHA inside containment.

This ensures that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. 10 CFR Part 50, Appendix A (Reference 8), General Design Criterion (GDC) 16, "Design," requires that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as the postulated accident conditions require. [Note: the applicability of 10 CFR 50, Appendix A, is a function of the plant-specific licensing basis.] GDC 19, "Control Room," requires that adequate radiation protection shall be provided to permit access and occupancy under acLcident conditions without personnel receiving radiation exposure in excess of 5 REM whole body, or its equivalent to any part of the body for the duration of the accident." GDC 54, 'Piping Systems Penetrating Containment," requires that piping systems penetrating primary reactor containment shall be provided with leak detection, isolation, and containment capabilities having redundancy, reliability, and performance capabilities which reflect the importance to safety of isolating these piping systems. Such piping systems shall be designed with a capability to test periodically the operability of the isolation valves and associated apparatus and to determine if valve leakage is within acceptable limits. GDC 56, "Primary Containment Isolation," describes the isolation provisions that must be provided for lines that connect directly to the containment atmosphere and which penetrate primary reactor containment unless it can be demonstrated that the isolation provisions for a specific class of lines are acceptable on some other defined basis. GDC 61, "Fuel Storage and Handling and Radioactivity Control," requires that the fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

The parameters of concern and the acceptance criteria applied are based on the requirements of 10 CFR 100 with respect to the calculated radiological consequences of a FHA and GDC 61 with respect to appropriate containment, confinement, and filtering systems.

5.2.2 Design Bases (UFSAR) UFSAR Section 15.7.4 The [LICENSEE]

design basis FHA is defined as the dropping of a spent fuel assembly onto the spent fuel pool fuel storage area or inside containment.

Both analyses assume the rupture of the cladding of all the fuel rods in the assembly.

Section 15.7.4 of the UFSAR discusses the consequences of a postulated FHA inside containment.

I A 5.2.3 Approved Methodologies U.S. NRC Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," describes a method acceptable to the NRC staff for licensee evaluation of the potential radiological consequences of a FHA. [Note: Regulatory Guide applicability is a function of the plant-specific licensing basis.] NUREG-0800 (Reference 9), U.S. NRC Standard Review Plan, Section 15.7.4, provides guidance to the NRC staff for the review and evaluation of system design features and plant procedures provided for the mitigation of the radiological consequences of postulated fuel handling accidents.

[Note: the applicability of NUREG-0800 is a function of the plant-specific licensing basis.] 5.2.4 Analysis The method of analysis used for evaluating t~le potf ntial radiological consequences of the postulated FHA is in compliance with RegulE tory Guide 1.25 and the guidance in NUREG-0800, Section 15.7.4 and NUREG/CR-5009 (Refe ence 10). The calculated doses were within the Standard Review Plan criteria of 6 REM to the whole body and 75 REM to the thyroid.

The analysis presented in Section 15.7.4 of the [LICENSEE]

UFSAR, demonstrating the adequacy of the system design features and plant procedures provided for the mitigation of the radiological consequences of postulated FHAs, assumes no credit is taken for iodine removal by the atmosphere filtration system filters. All radioactivity released to the containment is assumed to be released to the environment at ground level over a two-hour period. 5.2.5 Conclusion The technical analysis performed by [LICENSEE]

demonstrates that the consequent doses at the exclusion area and low population zone boundaries are well within the limits of 10 CFR 100. Therefore, the proposed License amendment is in compliance with GDC 16, 19, 54, 56, and 61 as well as Regulatory Guide 1.25, NUREG/CR-5009, and the criteria contained in NUREG-0800, Section 15.7.4. 6.0 ENVIRONMENTAL CONSIDERATION

[LICENSEE]

has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20 (Reference 11), or would change an inspection or surveillance requirement.

[LICENSEE]

has evaluated the proposed change and has determined that the change does not involve, (i) a significant hazards consideration, (ii) a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

As discussed above, the proposed changes do not involve a significant hazards consideration and the analysis demonstrates that the consequences from a FHA are well within the 10 CFR 100 limits. The implementation of administrative controls precludes a significant increase in occupational radiation exposure.

Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51 (Reference 12), specifically 10 CFR 51.22(c)

(9). Therefore, pursuant 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

7.0 REFERENCES

[use a standard format for citing references]

1. TSTF 312, Revision 1 2. Operating License Amendment 107 3. UFSAR Section 15.7.4 4. 10CFR100 5. NRC Standard Review Plan, Section 15.7.4, Revision 1 6. Operating License Amendment 95 7. Regulatory Guide 1.25 8. 10 CFR 50, Appendix A, GDC 9. NUREG-0800, Standard Review Plan 10. NUREG/CR-5009 11.10 CFR 20 12. 10 CFR 51 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION CHANGES (MARK-UP)

ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION PAGES ATTACHMENT 3 LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by [LICENSE]

in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Please direct questions regarding these commitments to [licensee contact].REGULATORY COMMITMENTS Due Date ATTACHMENT 4 CHANGES TO TS BASES PAGES TSB 3.9.4 The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, the equipment hatch must be held in place by at least four bolts. Consequences of a postulated FHA inside containment during core alterations or fuel handling activities remain within 10 CFR Part 100 limits.I i