L-2023-128, License Amendment Request to Revise TS 5.5.17, Pre-Stressed Concrete Containment Tendon Surveillance Program
| ML23262B018 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 09/19/2023 |
| From: | Strand D Point Beach |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| L-2023-128 | |
| Download: ML23262B018 (1) | |
Text
September 19, 2023 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 RE: Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 Renewed Facility Operating Licenses DPR-24 and DPR-27 NEXTera~
ENERGY.
~
L-2023-128 10 CFR 50.90 License Amendment Request to Revise TS 5.5.17, Pre-Stressed Concrete Containment Tendon Surveillance Program NextEra Energy Point Beach, LLC (NextEra) hereby requests amendments to Renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP)
Units 1 and 2, respectively. NextEra proposes to revise Technical Specification (TS) 5.5.17, "Pre-Stressed Concrete Containment Tendon Surveillance Program," for consistency with the requirements of 10 CFR 50.55a, "Codes and standards."
Specifically, the proposed changes replace the reference to Regulatory Guide (RG) 1.35 with a reference to Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel (B&PV) Code as contained in NUREG-1431, Revision 5, "Standard Technical Specifications - Westinghouse Plants." NextEra also proposes to delete the provisions of Surveillance Requirement 3.0.2 in TS 5.5.17.
The requirements of 10 CFR 50.55a were amended by the NRC in Federal Register 61 FR 41303 to incorporate by reference, Subsections IWE and IWL of Section XI of the ASME B&PV Code. Because of these changes to 10 CFR 50.55a, RG 1.35 was subsequently withdrawn by the NRC in August 2015 in Federal Register 80 FR 52067. Federal Register 80 FR 52067 stated that although RG 1.35 is withdrawn, its use in existing licenses is still valid, and changes to the licenses can be accomplished using other regulatory products.
This License Amendment Request is the product NextEra proposes to use to change TS 5.5.17 to replace RG 1.35 with a reference to ASME Section XI, Subsection IWL to align PBNP TS with 10 CFR 50.55a and the containment inservice inspection program.
This request contains four Attachments:
- 1. Attachment 1 provides a description and evaluation of the proposed license amendment request
- 2. Attachment 2 provides a marked-up TS Page
- 3. Attachment 3 provides a revised TS Page
- 4. Attachment 4 provides a marked-up TS Bases Page (Information only)
NextEra Energy Point Beach, LLC 6610 Nuclear Road, Two Rivers, WI 54241
Point Beach Nuclear Plant, Units 1 and 2 Docket Nos. 50-266 and 50-301 L-2023-XXX Page 2 of 2 NextEra has determined that the information for the proposed amendment does not involve a significant hazards consideration, authorize a significant change in the types or total amounts of effluent release, or result in any significant increase in individual or cumulative occupational radiation exposure. Therefore, the proposed amendment meets the categorical exclusion requirements of 10 CFR 51.22(c)(9) and pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
The Point Beach Onsite Review Group has reviewed the enclosed amendment request.
In accordance with 10 CFR 50.91(b)(1), a copy of this license amendment request is being forwarded to the designee for the State of Wisconsin.
This letter contains no NRC commitments.
Approval of the proposed amendments is requested within six months from the date of this submittal with implementation within 60 days following issuance of the amendment. Expedited review of the proposed amendment is requested as the next inspections for Unit 1 and for Unit 2 are due by June 17, 2024, and an alternative request pursuant to the requirements of 10 CFR 50.55a(z)(1) is planned for submittal and approval prior to the next inspections due by June 17, 2024.
Should you have any questions regarding this submittal, please contact Mr. Kenneth Mack, Licensing Manager, at 561-904-3635.
I declare under penalty of perjury that the foregoing is true and correct.
- l4~
Executed on the/ J day of September 2023.
Sincerely, Dianne Strand General Manager Regulatory Affairs cc:
USNRC Regional Administrator, Region Ill Project Manager, USNRC, Point Beach Nuclear Plant Resident Inspector, USNRC, Point Beach Nuclear Plant Public Service Commission of Wisconsin Attachments:
- 1. Description and Evaluation of Proposed License Amendment Request
- 2. Marked-up TS Page
- 3. Revised TS Page
- 4. Marked-up TS Bases Page (Information only)
ATTACHMENT 1 Description and Evaluation of Proposed License Amendment Request
Description and Evaluation of Proposed License Amendment Request
- 1.
SUMMARY
DESCRIPTION Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) hereby requests a license amendment to Units 1 and 2 renewed operating licenses DPR-24 and DPR-27, respectively.
Specifically, NextEra proposes to revise Technical Specification {TS) 5.5.17, "Pre-Stressed Concrete Containment Tendon Surveillance Program," to replace the reference to Regulatory Guide (RG) 1.35 with a reference to Section XI, Subsection IWL of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code. NextEra also proposes to delete the provisions of Surveillance Requirement (SR) 3.0.2 in TS 5.5.17.
The requirements of 10 CFR 50.55a were amended by the NRC in Federal Register 61 FR 41303 to incorporate, by reference, Subsections IWE and IWL of Section XI of the ASME B&PV Code.
Because of these changes to 10 CFR 50.55a, RG 1.35 was subsequently withdrawn by the NRC in August 2015 in 80 FR 52067. The withdrawal did not affect the Point Beach licensing bases requirement to use RG 1.35, as 80 FR 52067 stated that it did not alter any prior or existing licensing commitments based on its use. However, the withdrawal acknowledged that the guidance provided in RG 1.35 was incorporated into later revisions of Subsection IWL, or preserved in 10 CFR 50.55a. The NRC stated that as a result, RG 1.35 became redundant and was no longer needed. Therefore, NextEra proposes to revise TS 5.5.17 to replace the reference to RG 1.35 with a reference to ASME Section XI, Subsection IWL, to align Point Beach TS with 10 CFR 50.55a and the wording contained in NUREG-1431, Standard Technical Specifications -
Westinghouse Plants.
- 2. DETAILED DESCRIPTION NextEra proposes to revise TS 5.5.17 to replace the reference to RG 1.35 with a reference to Section XI, Subsection IWL of the ASME B&PV Code. The proposed TS change is shown below (deleted text is struck through while added text I italicized and bolded).
The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, 1990 Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC.
Additionally, NextEra also proposes to delete the applicability of SR 3.0.2 to the tendon surveillance inspection frequencies as shown below (deleted text is struck through).
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.
A marked-up TS page and a revised TS page are provided in Attachments 2 and 3, respectively.
The TS Bases will be revised for consistency with the proposed TS change and the wording contained in NUREG-1431. The Bases for TS 3.6.1, "Containment," is also marked up to replace the reference to RG 1.35 with a reference to Section XI, Subsection IWL of the ASME
B&PV Code. The marked-up TS Bases page is included in Attachment 4 for information only and will be updated in accordance with TS 5.5.13, "Technical Specifications (TS) Bases Control Program."
- 3. TECHNICAL EVALUATION The Point Beach containment is a reinforced concrete structure with a cylindrical wall, a flat foundation mat, and a shallow dome roof. The cylinder wall is prestressed with a post tensioning system in the vertical and horizontal directions, and the dome roof is prestressed utilizing a three way post tensioning system. The inside surface of the containment is lined with a carbon steel liner to ensure a high degree of leak tightness during operating and accident conditions.
Containment inservice inspection requirements originated with the issuance of RG 1.35, "lnservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," February 1973. The final revision, Revision 3, was issued in July 1990. TS 5.5.17, "Pre-Stressed Concrete Containment Tendon Surveillance Program," states in part, 'The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, 1990." This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments to ensure containment structural integrity.
A final rule amending 1 O CFR 50.55a, "Codes and standards," was issued by the NRC in 61 FR 41303 requiring licensees to implement the requirements of the ASME B&PV Code,Section XI, Subsections IWE and IWL. Subsection IWE delineates the requirements for inservice inspection of Class MC (metallic containment) components and the metallic liner of Class CC (concrete containment) components. Subsection IWL delineates the requirements for inservice inspection of concrete containments.
Because of these changes to 10 CFR 50.55a, which incorporated, by reference, the requirements of ASME Section XI, Subsection IWL, RG 1.35 was rendered obsolete.
Therefore, in August 2015, the NRC withdrew RG 1.35. The withdrawal did not affect the licensing bases of Point Beach, as 80 FR 52067 stated that it did not alter any prior or existing licensing commitments based on its use. However, the withdrawal acknowledged that the guidance provided in RG 1.35 was incorporated into later revisions of Subsection IWL, or preserved in 1 O CFR 50.55a. The NRC stated that as a result, RG 1.35 became redundant and was no longer needed.
10 CFR 50.55a(g)(5)(ii) requires that if a revised inservice inspection program for a facility conflicts with the TSs for the facility, the licensee must apply to the NRC for an amendment to restore conformance between the TSs and the revised program. Therefore, NextEra proposes to revise TS 5.5.17 to replace the reference to RG 1.35 with a reference to ASME Section XI, Subsection IWL, to align the Point Beach TS with 1 O CFR 50.55a and the containment inservice inspection program. This change provides clarity and reduces the potential for human error by eliminating multiple, redundant governing standards for containment tendon surveillance.
Additionally, NextEra proposes to delete the applicability of SR 3.0.2 to the tendon surveillance inspection frequencies. SR 3.0.2 states that each SR is met if the Surveillance is performed within 1.25 times the interval specified in the frequency, as measured from the previous performance or as measured from the time a specified condition of the frequency is met.
Since the tendon inspection frequencies will be in accordance with ASME Section XI, Subsection IWL, which specifies requirements for extending inspection frequencies, the provisions of SR 3.0.2 are no longer needed for the Pre-Stressed Concrete Containment Tendon Surveillance Program.
The proposed change adopts wording that is identical to NUREG-1431 for the containment tendon surveillance program.
- 4. REGULATORY EVALUATION 4.1 Applicable Regulatory Requirements/Criteria Section 50.55a(g)(4) of 10 CFR requires, in part, that throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section XI of editions and addenda of the ASME B&PV Code that become effective subsequent to editions specified in paragraphs (g)(2) and (g)(3) of 10 CFR 50.55a and that are incorporated by reference in paragraph (a)(1)(ii) of 10 CFR 50.55a, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
Section 50.55a(g)(5)(i) of 1 O CFR requires that the inservice inspection program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of 10 CFR 50.55a.
Section 50.55a(g)(5)(ii) of 1 O CFR requires that if a revised inservice inspection program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program.
4.2 Precedent The proposed amendments are similar to changes previously approved by the NRC for use at the following stations:
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Amendment 240 (Unit 1) and 214 (Unit 2), January 30, 2001 (ADAMS Accession Number ML003776835)
LaSalle County Station, Units 1 and 2, Amendments 148 (Unit 1) and 134 (Unit 2),
August 16, 2001, (ADAMS Accession Number ML012060445)
Wolf Creek Generating Station, Amendment 152, March 17, 2004, (ADAMS Accession Number ML040820934)
Three Mile Island Nuclear Station Unit 1, Amendment 251, July 13, 2004 (ADAMS Accession Number ML042010070)
Joseph M. Farley Nuclear Plant, Units 1 and 2, Amendments 172 (Unit 1) and 165 (Unit 2), April 14, 2006, (ADAMS Accession Number ML060830380)
Vogtle Electric Generating Station, Units 1 and 2, Amendments 14 7 (Unit 1) and 127 (Unit 2), December 12, 2006, (ADAMS Accession Number ML062970484)
Braidwood Station Units 1 and 2, Amendments 158 (Unit 1) and 158 (Unit 2), and Byron Station Units 1 and 2, Amendments 163 (Unit 1) and 163 (Unit 2), March 26, 2009, (ADAMS Accession Number ML090610133)
Millstone Power Station Unit 2, Amendment 341, September 29, 2020, (ADAMS Accession Number ML20237H995) 4.3 No Significant Hazards Consideration Analysis Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC (NextEra) is requesting amendments to Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant, Units 1 and 2, respectively. The proposed amendments would revise Technical Specification (TS) 5.5.17, "Pre-Stressed Concrete Containment Tendon Surveillance Program," to replace the reference to Regulatory Guide (RG) 1.35 with a reference to Section XI, Subsection IWL of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code as contained in NUREG-1431, Revision 5, "Standard Technical Specifications - Westinghouse Plants."
The requirements of 10 CFR 50.55a were amended by the NRC in Federal Register 61 FR 41303 to incorporate, by reference, Subsections IWE and IWL of Section XI of the ASME B&PV Code. Because of these changes to 10 CFR 50.55a, RG 1.35 was subsequently withdrawn by the NRC in August 2015 in 80 FR 52067. The withdrawal did not affect the Point Beach licensing bases requirement to use RG 1.35, as 80 FR 52067 stated that it did not alter any prior or existing licensing commitments based on its use. However, the withdrawal acknowledged that the guidance provided in RG 1.35 was incorporated into later revisions of Subsection IWL, or preserved in 10 CFR 50.55a. The NRC stated that as a result, RG 1.35 became redundant and was no longer needed. Therefore, NextEra proposes to revise TS 5.5.17 to replace the reference to RG 1.35 with a reference to ASME Section XI, Subsection IWL, to align Point Beach TS with 10 CFR 50.55a and the wording contained in NUREG-1431, Standard Technical Specifications - Westinghouse Plants. Additionally, since the tendon inspection frequencies will be in accordance with ASME Section XI, Subsection IWL, the provisions of SR 3.0.2 are no longer needed for the Pre-Stressed Concrete Containment Tendon Surveillance Program.
NextEra has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:
Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No The proposed amendments revise TS 5.5.17 to replace the reference to RG 1. 35 with a reference to Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code, which is incorporated by reference in 10 CFR 50.55a, "Codes and standards."
The proposed amendments provide consistency between the containment tendon surveillance program criteria in TS 5.5.17 and the regulatory requirements specified in 10 CFR 50.55a.
These regulatory requirements and the associated surveillance program ensure that the pre-stressed concrete containment tendon system is capable of maintaining the structural integrity of the containment during operating and accident conditions. The pre-stressed concrete containment tendon system is not an initiator of any accident. Therefore, this change is not related to the probability of any accident previously evaluated.
The proposed amendments ensures that the tendon surveillance program requirement in TS 5.5.17 addresses the appropriate regulatory criteria. The inspections required by the ASME Code serve to maintain containment response to accident conditions, by causing the
identification and repair of defects in the containment buildings.
The proposed amendments do not result in any reduction in the effectiveness of the existing surveillance program. The tendon surveillance program will continue to ensure that the containment structure is capable of performing its intended safety function in the event of a design basis accident.
Therefore, the proposed change does not involve a significant increase in the probability or consequence of an accident previously evaluated.
Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No The proposed amendments to TS 5.5.17 provide consistency between the containment tendon surveillance program criteria in TS 5.5.17 and the regulatory requirements specified in 10 CFR 50.55a. The proposed amendments do not result in any reduction in the effectiveness of the existing containment tendon surveillance program. The containment tendon surveillance program will continue to satisfy the requirements ASME Section XI, Subsection IWL, thus ensuring that the containment structure will perform its design safety function. This change does not introduce any new accident precursors and does not involve any alterations to plant configurations, which could initiate a new or different kind of accident.
Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.
Do the proposed changes involve a significant reduction in a margin of safety?
Response: No The proposed amendments provide consistency between the containment tendon surveillance program criteria in TS 5.5.17 and the regulatory requirements specified in 10 CFR 50.55a. The proposed amendments do not result in any reduction in the effectiveness of the existing containment tendon surveillance program. The containment tendon surveillance program will continue to satisfy the requirements of ASME Section XI, Subsection IWL, thus ensuring that the containment structure will perform its design safety function in accordance with existing margins of safety for containment integrity.
Therefore, the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, NextEra concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.4 Conclusions In conclusion, based on the considerations discussed herein, (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.
- 5. ENVIRONMENTAL CONSIDERATION The proposed amendments would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR Part 20, or would change an inspection or surveillance requirement. However, the proposed change 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 cumulative occupational radiation exposure. Accordingly, the proposed amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 1 O CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.
- 6. REFERENCES
- 1. Regulatory Guide 1.35, lnservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments, Revision 3, July 11, 1990
- 2. NUREG-1431, "Standard Technical Specifications - Westinghouse Plants," Revision 5 Marked-up TS Page (1 Page Follows)
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 5.5.17 Point Beach Reactor Coolant System (RCS) Pressure Isolation Valve (PIV)
Leakage Program A program shall be established to verify the leakage from each RCS PIV is within the limits specified below, in accordance with the Event V Order, issued April 20, 1981.
- a.
Minimum differential test pressure shall not be less than 150 psid.
- b.
Leakage rate acceptance criteria are:
- 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
- 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.
The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, 1990Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR 50.55a, except where an alternative, exemption, or relief has been authorized by the NRC.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.
5.5-18 Unit 1 - Amendment No. 2+ Unit 2 - Amendment No. 2-7 Revised TS Page (1 Page Follows)
Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 5.5.17 Point Beach Reactor Coolant System (RCS) Pressure Isolation Valve (PIV)
Leakage Program A program shall be established to verify the leakage from each RCS PIV is within the limits specified below, in accordance with the Event V Order, issued April 20, 1981.
- a.
Minimum differential test pressure shall not be less than 150 psid.
- b.
Leakage rate acceptance criteria are:
- 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
- 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
Pre-Stressed Concrete Containment Tendon Surveillance Program This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.
The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies,
and acceptance criteria shall be in accordance with Section XI, Subsection IWL of the ASME Boiler and Pressure Vessel Code and applicable addenda as _
required by 10 CFR 50.55a, except where an alternative, exemption, or relief _
has been authorized by the NRC._
The provisions of SR 3.0.3 are applicable to the Tendon Surveillance_
Program inspection frequencies.
5.5-18 Unit 1 - Amendment No.
Unit 2 - Amendment No.
Marked-up TS Bases Page (Information only)
(1 Page Follows)
BASES SURVEILLANCE REQUIREMENTS
( continued)
REFERENCES Point Beach Containment B 3.6.1 are bounded by the assumptions of the safety analysis. SR Frequencies are as required by the Containment Leakage Rate Testing Program. These periodic testing requirements verify that the containment leakage rate does not exceed the leakage rate assumed in the safety analysis.
SR 3.6.1.2 For ungrouted, post tensioned tendons, this SR ensures that the structural integrity of the containment will be maintained in accordance with the provisions of the Containment Tendon Surveillance Program.
Testing and Frequency are consistent in accordance with the recommendations of Regulatory Guide 1.35ASME Code,Section XI, Subsection IWL (Ref. 4), and applicable addenda as required by 10 CFR 50.55a.
- 1. 10 CFR 50, Appendix J, Option B.
- 2. FSAR, Chapter 14.
- 3. FSAR, Section 5.1.
- 4. ASME Code,Section XI, Subsection IWLRegulatory Guide 1.35, Revision 3.