ML20063N620
| ML20063N620 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  (DPR-24-A-063, DPR-24-A-63, DPR-27-A-068, DPR-27-A-68) |
| Issue date: | 08/31/1982 |
| From: | Clark R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20063C546 | List: |
| References | |
| TAC-49449, TAC-49450, NUDOCS 8210040420 | |
| Download: ML20063N620 (16) | |
Text
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UNITED STATES
/V b,
NUCLEAR REGULATORY COMMISSION g
WASHINGTON, D. C. 20555 s.;...../
WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-266 POINT BEACH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No. DPR-24 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated February 17,1977, compl4s with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be -
conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
DESIGNAiED ORIGINAL Cortified By pakon88! ole 8 san P
6 a4 '
A i i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indic.-tad in the attachment to this license
' amendment, and paragraph 3.B of Facility Operating License No. DPR-24 is hereby amended to read as follows:
(B) Technical Specifications 4
The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 20 days from the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
( k[,
f~~
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 31, 1982 I
- a s
ATTACHMENT TO LICENSE AMENDMENTS AMEND? TENT NO. 63 TO FACILITY OPERATING LICENSE NO. OPR-24 AttENDMENT NO. 68 TO FACILITY OPERATING LICENSE NO. DPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A as follows:
Remove Pages Insert Pages
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15.i 15.1 15.4.2-1 15.4.2-1 15.4.2-1c 15.4.2-1c 15.4.2-1d 15.4.2-2 15.4.2-2 15.4.2-3 15.4.2-4 15.4.2-5 15.6.9-7 15.6.9-7 c.
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r TABLE OF CONTENTS Section Title Page 15 TECHNICAL SPECIFICATIONS AND BASES 15.1 Definitions 15.1-1 1
15.2.0 Safety Limits and Limiting Safety System Settings 15.2.1-1 15.2.1 Safety Limit, Reactor Core 15.2.1 15.2.2 Safety Limit, Reactor. Coolant System Pressure 15.2.2-1 15.2.3 Limiting Safety System Settings, Protective In,strumentation.
15.2.3-1 15.3 Limiting Conditions for Operation 15.3-0 15.3.0 Ceneral Considerations 15.3.0-1 15.3.1 Reactor Coolant System 15.3.1-1 15.3.2 Chemical and Volume Control System 15.3.2-1
- 15. 3.)
Emergency Core Cooling System, Auxiliary CoolfE3 '
Systems, Air Recirculation Fan Coolers, and Containment Spray 15.3.3-1 15.3.4 Steam and Power Conversion System 15.3.4-1 15.3.5 Instrumentation System 15.3.5-1 15.3.6 Containment System 15.3.6-1 15.3.7 Auxiliary Electrical Systems 15.3.7-1 15.3.8 Refueling 15.3.8-1 15.3.9 Effluent Releases 15.3.9-1 15.3.10 Control Rod and Power Distribution Limits 15.3.10-1 4
15.3.11 Movable In-Core Instrumentation 15.3.11-1 15.3.12 Control Room Emergency Filtration 15.3.12-1 15.3.13 Shock Suppressors (Snubbers) 15.3.13-1 15.3.14 Fire Protection System 15.3.14-1 15.3.15 overpressure Mitigating System 15.3.15-1 15.3.16 Reactor Coolant System Pressure Isolation Valves 15.3.16-1 15.4 Surveillance Requirements 15.4-1 15.4.1 Operational Safety Review 15.4.1-1 15.4.2 In-Servica Inspection of Safety Class Components 15.4.2-1 l
15.4.3 Primary System Testing Following Opening 15.4.3-1 15.4.4 Containment Tests 15.4.4-1 15.4.5 Emergency Core Cooling System and Containment Cooling System Tests 15.4.5-1 15.4.6 Emergency Power System Periodic Tests 15.4.6-1 15.4.7 Main Steam Stop Valves 15.4.7-1 15.4.8 Auxiliary Feedwater System 15.4.8-1 15.4.9 Reactivity Anomalies 15.4.9-1 15.4.10 Operational Environmental Monitoring 15.4.10-1 15.4.11 Control Room Emergency Filtration 15.4.11-1 15.4.12 Miscellaneous Radioactive Materials Sources 15.4.12-1 15.4.13 Shock Suppressors (Snubbers) 15.4.13-1 15.4.14 Surveillance of Auxiliary Building Crane 15.4.14-1 15.4.15 Fire Protection System 15.4.15-1 15.4.16 Reactor Coolant System Pressure Isolation Valves Leakage Tests 15.4.16-1 15-1 Unit 1 - Olddt 4ptf120/ 1981, /50, 63 Unit 2 - /ddf Ap/II 20/ 1981,/5(, 68
e s
15.4.2 IN-SERVICE INSPECTION OF SAFETY CLASS COMPONENTS Applicability Applies to in-service inspection of Safety Class Components.
Obiectives To provide-assurance of the continuing integrity of the safety class systems.
l Specifications A.
Steam Generator Tube Inspection Requirements 1.
Tube Inspection Entry from the hot-leg side with examination f rom the point of entry completely around the U-bend to the top support of the cold-leg is considered a tube inspection.
2.
Sample Selection and Testing Selection and testing of steam generator tubes shall be made on the following basis:
(a) One steam generator of each unit shall be inspected during inservice inspection in accordance with the following requirements:
1.
The inservice inspection may be limited to one steam generator on an alternating sequence basis. This _
examination shall include at least 6% of the tubes if the results of the first or a prior inspection indicate that both generators are performing in a comparable manner.
2.
'4 hen both steam generators are required to be examined by Table 15.4.2-1 and if the condition of the tubes in one generator is found to be more severe than in the other steam generator of a unit, the steam generator sampling sequence at the subsequent inservice inspection shall be modified to examine the steam generator with the more severe condition.
(b) The minimum sample size, inspection result classification and the associated required action shall be in conformance "ith the requirements specified in Table 15.4.2-1.
The results of each sampling examination of a steam generator shall be classified into the following three categories:
Point Beach Unit 1 15.4.2-1 Amendment No. 19, 63
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Defset is en imparfection of such severity that it exceeds the minimum acceptable tube wall thickness of 50%. A tube containing a defect is defective.
Plunning limit is the imperfection depth beyond which the tube must be removed from service, because the tube may become defective prior to ihe next scheduled inspection.
The plugging limit is 40% of the nominal tube wall thickness.
6.
Corrective Measures l
~
All tubes that leak or have degradation exceeding the plugging limit shall be plugged prior to return to power from a refueling or inservice inspection condition.*
7.
Reports g.
(a) After each inservice examination, the number of tubes l
plugged in each steam generator shall be reported to the Commission as soon as practicable.
(b) The complete results of the steam generator tube inservice I
inspection shall be included. in the Operating Report for the period in which the inspection was completed.
In addition all results in Category C-3 of Table 15.4.2-1 shall be reported to the C0= mission prior to resumption of plant operation.
(c) Reports shall include:
1 1.
Number and axtent of tubes inspected 2.
Location and percent of all thickness penetration for each indication 3.
Identification of tubes plugged (d) Reports required by Table 15.4.2 Steam Generator Tube l
Inspection - shall provide the information required,by Specification 15.4.2.A.7(b) and a description of investiga-l tions conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.
B.
In-Service Inspection of Safety Class Components Other than Steam Generator Tubes 1.
Inservice inspection of ASME Code Class 1, Class 2 and Class 3.
components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specific written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
- Point Beach Nuclear Plant Unit 1 may be operated at power with up to six tubes in one steam generator hcving degradation exceeding the plugging limit provided those tubes have been repaired by insertion of sleeves into the tubes to bridge the degraded or defective portion of the tube.
The plugging limit is 35% of the nominal sleeve wall thickcsss for tubes that have been repaired by sleeving.
Point Beach Unit 1 15.4.2-lc Amendment No. 19, 56, 63
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2.
Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specified written relief is granted by the NRC pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
3.
Containment isolation valves will be tested in accordance with Technical Specification 15.4.4 instead of Section IWV-3420, Valve Leak Rate Test.
Bases l
The proposed inspection program 1
, where practical, in compliance with the recommendations of ASME Boiler a d Pressura Vessel Code,Section XI, Summer 1971 Addenda.
It must be recogn zed, however, that equipment and techniques to perform the inspection are still in development. It is recognized, however, that examinations in certain areas are necessary and therefore a schedule is proposed that includes areas and frequencies that are believed practical at this time for this reactor.
In most areas scheduled for test, a detailed pre-service mapping will be conducted using techniques which can be used for post-operation inspections. The areas indicated for inspection represent those of relatively high stress and therefore will serve to indicate potential problems before significant flaws develop there or at other areas. As more experience is gained in operation of pressurized-water reactors, the recommended time schedule and location of inspection might be altered, or should new techniques be developed, consideration will be given to incorporate these new techniques into this inspection program.
The use of conventional non-destructive, direct visual and remote visual test techniques can be applied to the inspection of al'1 primary loop comnonents except for the reactor vessel. The reactor vessel presents special problems because of the radiation levels and remote underwater accessibility to this component. Because of these limitations on access.co the reactor vessel, several steps have been incorporated into the design and manufacnuring procedures future.y{9tionfornon-destructivetesttechniqueswhichmaybeavailableinthe in prep The techniques for in-service inspection include the use of visual inspections, volumetric (ultrasonic or radiographic) and surface (dye penetrant or magnetic particle) testing of selected parts during refueling periods.
The intent of the inspection is the detection of flaws large enough to initiate fast fracture and gross leakage prior to subsequent inspection. At this time it is judged that such a flaw is substantially larger than 1/2 inch by 1 inch which is the degree of detectability. The inspection method is designed to detect flaws of this magniture.
(1) FSAR - Section 4.4 Point Beach Unit 1 15.4.2-2 Amendment No. 63
3.
Observed inadequacies in the implementation of admin-istrative or procedural controls which threaten to cause reduction of degree of redundancy.provided in reactor protection systems or engineered safety feature systems.
l 4.
Abnormal degradation of systems other than those specified in 15.6.8.2.A.3 above designed to contain radioactive material resulting from the fission process.
15.6.9.3 UNIQUE. REPORTING REQUIREMENTS The following written reports shall be submitted to the Director, Office of Nuclear Reactor Regulation, USNRC:
A.
Each integrated leak test shall be the subject of a summary technical report, including results of the local leak rate tests and isolation valve leak rate tests since the last report. The report shall include analysis and interpreta-tions of the results which demonstrate compliance with specified leak rate limits.
B.
Deleted C.
Submission of a report within 60 days after January 1 and after July 1 each year for the six-month period or fraction thereof, ending June 30 and December 31 containing:
l l
4 Point Beach Unit 1 15.6.9-7 Amendment No.19, 63
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jra ucoq'o, UNITED STATES
,o NUCLEAR REGULATORY COMMISSION
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WASHINGTON. D. C. 20555 4'~
gh WISCONSIN ELECTRIC POWER COMPANY DOCKET NO. 50-301 POINT BEACH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment. No. 68 License No. DPR-27 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Wisconsin Electric Power Company (the licensee) dated November 27, 1978, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without' endangering the health ~
and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 19 CFR Part 51 of the Commission's regulations and all applicable.equirements have been satisfied.
DESIGNATED ORIGINAL Cortified Ey
\\
. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.9 of Facility Operating License No. DPR-27 is hereby amended to read as follows:
(B) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 68 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective 20 days from the date of its issuance.
FOR THE NUCLEAR REGULAT0h? COMMISSION
' ] [o.
Robert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: August 31, 1982
ATTACHMENT TO LICENSE AMENDMENTS AMENDfiENT NO. 63 TO FACILITY OPERATING LICENSE NO. DPR-24 AttENDMENT NO. 68 TO FA'ILITY OPERATING LICENSE NO. OPR-27 DOCKET NOS. 50-266 AND 50-301 Revise Appendix A as follows:
Remove Pages Insert Pages
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15.1 15.1 15.4.2-1 15.4.2-1 15.4.2-1c 15.4.2-Ic 15.4.2-1d 15.4.2-2 15.4.2-2 15.4.2-3 15.4.2-4 15.4.2-5 15.6.9-7 15.6.9-7
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T
U TABLE OF CONTENTS Section Title Page 15 TECHNICAL SPECIFICATIONS AND BASES 15.1 Definitions 15.1-1 15.2.0 Safety Limits and Limiting safety System Settings 15.2.1-1 15.2.1 Safety Limit, Reactor Core 15.2.1-1, 15.2.2 Safety Limit, Reactor Coolant System Pressure 15.2.2-1 15.2.3 Limiting Safety System Settings, Protective Instrumentation 15.2.3-1 1 ~., 3 Limiting Conditions for Operation 15.3-0 15 1.0' General Considerations 15.3.0-1 15.3.1 Reactor Coolant System 15.3.1-1 15....
Chemical and Volume Control System 15.3.2-1 15.3.3 Emergency Core Cooling System, Auxiliary Cooltag -
Systems, Air Recirculation Fan Coolers, and-Containment Spray 15.3.3-1 4
15.3.4 Steam and Power Conversion System 15.3.4-1 15.3.5 Instrumentation System 15.3.5-1 15.3.6 Containment System 15.3.6-1 15.3.7 Auxiliary Electrical Systems 15.3.7-1 15.3.8 Refueling 15.3.8-1 15.3.9 Effluent Releases 15.3.9-1 15.3.10 Control Rod and Power Distribution Limits 15.3.10-1 15.3.11 Movable In-Core Instrumentation 15.3.11-1 15.3.12 Control Room Emergency Filtra: ion 15.3.12-1 15.3.13 Shock Suppressors (Snubbers) 15.3.13-1 15.3.14 Fire Protection System 15.3.14-1 15.3.15 overpressure Mitigating System 15.3.15-1 15.3.16 Reactor Coolant System Pressure Isolation Valves 15.3.16-1 15.4 Surveillance Requirements 15[4-1 15.4.1 Operational Safety Review 15.4.1-1 15.4.2 In-Service Inspection of bafety Class Components 15.4.2-1 l
15.4.3 Primary System Testing Following Opening 15.4.3-1 l
15.4.4 Containment Tests 15.4.4-1 l
15.4.5 Emergency Core Cooling System and Containment Cooling System Tests 15.4.5-1 15.4.6 Emergency Power System Periodic Tests 15.4.6-1 15.4.7 Main Steam Stop Valves 15.4.7-1 15.4.8 Auxiliary Feedwater System 15.4.8-1 15.4.9 Reactivity Anomalies 15.4.9-1 l
15.4.10 Operational Euvironmental Monitoring 15.4.10-1 15.4.11 Control Room Emergency Filtration 15.4.11-1 i
15.4.12 Miscellaneous Radioactive Materials Sources 15.4.12-1 15.4.13 Shock Suppressors (Snubbers) 15.4.13-1 15.4.14 Surveillance of Auxiliary Building Crane 15.4.14-1 15.4.15 Fire Protection System 15.4.15-1 15.4.16 Reactor Coolant System Pressure Isolation Valves I
Leakage Tests 15.4.16-1 15-1 Unit 1 - /ddi Aptf1 20t 1981,/5G 63
[
l Unit 2 - iddi Aptf1 29/ 1981,/gg, 68 i
15.4.2 IN-FERVICE INSPECTION OF SAFETY CLASS COMPONENTS Applicability Applies to in-service inspection of Safety Class Components.
l Obiectives To provide assurance of the continuing integrity of the safety class systems.
l Specifications A.
Steam Generator Tube Inspection Requirements 1.
Tube Inspection
~~ -
Entry from the hot-leg side with examination from the point of i entry completely around the U-bend to the top support of the cold-leg is considered a tube inspection.
2.
Sample Selection and Testing Selection and testing of steam generator tubes shall be made on the following basis:
(a) One steam generator of each unit shall be inspected during inservice inspection in accordance with the following requirements:
1.
The inservice inspection may be limited to one steam generator on an alternating sequence basis. This -
examinatdon shall include at least 6% of the tubes if the results of the first or a prior inspection indicate that both generators are performing in a comparable manner.
2'.
When both steam generators are required to be examined l
by Table 15.4.2-1 and if the condition of the tubes in one l
generator is found to be more severe than in the other steam generator of a unit, the steam generator sampling sequence at the subsequent inservice inspection shall be modified to examine the steam generator with the more severe condition.
(b)
The minimum sample size, inspection result classification-and the associated required action shall be in conformance with the requirements specified in Table'15.4.2-1.
The
(
results of each sampling examinatico of a steam generator shall be classified into the following three categories:
Point Beach Unit 2 15.4.2-1 Amendment No. I2, 68 l
l
6 Defect is an imperfection of such severity that it exceeds the minimum acceptable tube, wall thickness of 50%. A tube' containing a defect is defective.
plunning Limit is the imperfection depth beyond which the tube must be removed from service, because the tube may become defective prior to the next scheduled inspection.
The plugging limit is 40% of the nominal tube wall thickness.
6.
Corrective Measures l
All tubes that leak or have degradation exceeding the plugging limit shall be plugged prior to return to power from a refueling or inservice inspection condition.
7.
Reports l
'(a)b After each inservice examination, the number of tubes I
plugged in each steam generator shall be reported to the i
Commission as soon as practicable.
(b)
The complete results of the steam generator tube inservice 1
inspection shall be included in the Operating Report for the period in which the inspection was completed.
In addition all results in Category C-3 of Table 15.4.2-1 shall be reported to the Co= mission prior to resumption of plant operation.
(c) Reports shall include:
l 1.
Number and extent of tubes inspected 2.
Location and percent of all thickness penetration for each indication 3.
Identification of tubes plugged (d) Reports required by Table 15.4.2 Steam Generator Tube l
Inspection - shall provide the information required by Specification 15.4.2.A.7(b) and a description of investigations l conducted to determine cause of the tube-degradation and corrective measures taken to prevent recurrence.
B.
In-Service Inspection of Safety Class Components Other Than Steam Generator Tubes 1.
Inservice inspection of ASME Code Class 1, Class 2 and Class 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specific written relief is granted by the NRC, pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
Point Beach Unit 2 15.4.2-lc Amendment No. 12, 68
2.
Inservice testing of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g) modified by Section 50.55a(b), except where specified written relief is granted by the NRC pursuant to 10 CFR,50, Section 50.55a(g)(6)(1).
3.
Containment isolation valves will be tested in accordance with Technical Specification 15.4.4 instead of Section IWV-3420 Valve Leak Rate Test.
Bases l
The proposed inspection program is, where practical, in comgliance with the reco=mendations of ASME Boiler and Pressure Vessel Code,Section XI, Summer 1971 Addenda.
It must be recognized, however, that equipment and techniques to perform the inspection are still in development.
It is recognized, however, that examinations in certain areas are necessary and therefore a schedule is proposed that includes areas and frequencies that are believed practical at this time for this reactor.
In most areas scheduled for test, a detailed pre-service mapping will be conducted using techniques which can be used for post-operation inspections. The areas indicated for inspection represent those of relatively high stress and therefore will serve to indicate potential problems before significant flaws develop there or at other areas. As more experience is gained in operation of pressurized-water reactors, the recommended time schedule and location of inspection might be altered, or should new techniques be developed, consideration will be given to incorporate these new techniques into this inspection program.
The use of conventional non-destructive, direct visual and remote visual test techniques can be applied to the inspection of all primary loop components except for the reactor vessel. The reactor vessel presents special problems because of the radiation levels and remote underwater accessibility to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in prep future.y{jtionfornon-destructivetest techniques which may be available in the The techniques for in-service inspection include the use of visual inspections, volumetric (ultrasonic or radiographic) and surface (dye penetrant or magnetic particle) testing of selected parts during refueling periods.
The intent of the inspection is the detection of flaws large enough to initiate fast fracture and gross leakage prior to subsequent inspection. At this-time it is judged that such a flaw is substantially larger than 1/2 inch by 1 inch which is the degree of detectability. The inspection method is designed to detect flaws of this magniture.
(1) FSAR - Section 4.4 Point Beach Unit 2 15.4.2-2 Amendment No. 68
o 3.
Observed inadequacies in' the implementation of admin- -
istrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection systems or engineered safety feature systems.
4.
Abnormal deg*radation of systems other than those spacified in 15.6.8.2.A.3 above designed to contain radioactive material resulting from the fission process.
15.6.9.3 UNIQUE REPORTING REQUIREMENTS The following written reports shall be submitted to the' Director, Office of Nuclear Reactor Regulation, USNRC:
A.
Each integrated leak test shall be the subject of a summary technical report, including results of the local leak rate tests and isolation valve leak rate tests since the last report. The report shall include analysis and interpreta-tions of the results which demonstrate compliance with specified leak rate limits.
B.
Deleted C.
Submission of a report within 60 days after January 1 and after July 1 each year for the six-month period or fraction thereof, ending June 30 and December 31 containing:
Point Beach Unit 2 15.6.9-7 Amendment No. 24, 68
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