ML20054L117

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Amend 1 to License NPF-11 Authorizing Operation of Facility at Full Power
ML20054L117
Person / Time
Site: LaSalle Constellation icon.png
Issue date: 06/15/1982
From: Harold Denton
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20054L115 List:
References
NUDOCS 8207070165
Download: ML20054L117 (17)
v • d • e


Text

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  • n NUCLEAR REGULATORY COMMISSION

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COMMONWEALTH EDISON COMPANY

. DOCKET NO. 50-373 LA_SALLE COUNTY STATION, UNIT 1 FACILITY OPERATING LICENSE License No. NPF-11 Amendment No. 1

1. The Nuclear Regulatory Cmmission (the Cmmission or the NRC) having found that:

A. The application for a license filed by the Commonwealth Edison Company complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I, and all required notifications to other agencies or bodies have been duly made; B. Construction of the La Salle County Station, Unit 1 (the facility),

has been substantially cmpleted in confomity with Construction Permit No. CPPR-99 and the application, as amended, the provisions of the Act, and the regulations of the Commission; C. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Cmmission; D. There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted without endangering the health

! and safety of the public, and (ii) that such activities will be con-l ducted in compliance with the Cmmission's regulations set forth in 10 CFR Chapter I; E. The Commonwealth Edison Company is technically qualified to engage in the activities authorized by this operating license in accordance with the Coratission's regulations set forth in 10 CFR Chapter I;

, F. The Commnwealth Edison Company has satisfied the applicable provisions of 10 CFR Part 140, " Financial Protection Requirements and Indemnity Agreements," of the Cmmission's regulations; G. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; 8207070165 820615 PDR ADOCK 05000373 g P PDR

. A. _ _ .. ._. .._

,' 4 H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Facility Operating License No. NPF-11, subject to the conditions for protection of the environment set forth herein, is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. The receipt, possession, and use of source, by-product and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and ?C.

2. Based on the foregoing findings regarding this facility, Facility Operating License NPF-ll is hereby issued to the Comnonwealth Edison Company (the licensee) to read as follows:

A. This license applies to the La Salle County Station, Unit 1, a boiling water nuclear reactor and associated equipment, owned by the Common-wealth Edison Company. The facility is located in Brookfield Township, La Salle County, Illinois, and is described in the licensee's " Final Safety Analysis Report," as supplemented and amended, and in the licensee's Environmental Report, as supplemented and amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1) Commonwealth Edison Company, pursuant to Section 103 of the Act and 10 CFR Part 50, " Licensing of Production and Utilization Facilities," to possess, use, and operate the facility at the designated location in Brookfield Township, La Salle County, Illinois, in accordance with the procedures and limitations set forth in this license; i

(2) Commonwealth Edison Company, pursuant to the Act and 10 CFR Part s 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; (3) Commonwealth Edison Company, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in anounts as required; 1

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(4) Commonwealth Edison Company, pursuant to the Act and 10 CFR Parts  ;

30, 40, and 70, to receive, possess, and use in amounts as required I any byproduct, source or special nuclear material without restriction  ;

to chemical or physical fom, for sample analysis or instrument 1 calibration or associated with radioactive apparatus or components;  !

and (5) Commonwealth Edison Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is stbject to the additional conditions specified or incorporated below:

(1) Maximum Power Level The licensee is authorized to operate the facility at full reactor core power (3323 megawatts thermal) in accordance with the conditions specified herein and in Attachment 1 to this license. The preoperational tests, startup tests and other items identified in Attachment 1 to this license shall be completed as specified.

Attachment 1 is an integral part of this license.

l (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environ-I mental Protection Plan contained in Appendix B attached hereto are hereby incorporated in this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Resolution of Rebar Damage and Adequacy of Off-gas Building Roof Prior to power operation following initial criticality and zero power physics testing, the licensee shall complete its assessment of the rebar damaged due to drilling and coring in concrete and the structural adequacy of the of f-gas building roof, and shall submit the results to the NRC staff for review and approval .

(4) Snubbers l Prior to startup after the first refueling outage, the licensee shall provide, as necessary, a revision to the Technical Specifications to renove snubbers that are detemined to be unnecessary and replace them with rigid strut and rod assemblies.

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(5) Deferred Preoperational Deficiencies The licensee shall satisfactorily resolve those deficiencies which were deferred from the preoperational testing program on a schedule that shall assure that the capability of a system required to be operable by Technical Specification is not degraded.

(6) Surveillance of Tendons (Section 3.8.1*, SSER #3)

Prior to exceeding 5 percent power, the licensee shall supply the predicted lift-off forces required to complete Tables 4.6.1.5-1 and 4.6.1.5-2 of the Technical Specifications.

(7) Masonry Wall (Section 3.8.3, SER ,SSER #2)

(a) The licensee's present modifications shall not preclude the option of implementing additional modifications if directed by future staff review of the licensee's design criteria.

(b) Prior to startup after the first refueling outage, the differences between the staff's interim criteria and the criteria used by the licensee shall be resolved, and the modifications that might result from such a resolution shall be implemented.

(8) Inservice Testing of Pumps and Valves (Section 3.9.6, SER)

Pursuant to 10 CFR Part 50.55a, the relief that the licensee has requested from the ptsnp and valve testing requirements of 10 CFR Part 50, Section 55.55 (g)(2) and (g)(4){i) is granted for that portion of the initial 120-month period during which the staff completes its review.

(9) Dynamic Qualification (3.10, SER, SSER #1, SSER #2)

(a) Prior to startup after the first refueling outage, the licensee shall complete any modifications or replacement of equipment as a result of the fatigue evaluation. In the interim, the licensee shall document the occurrence of every safety relief valve actuation into the suppression pool, the associated cumulative damage factors calculated for typical representative equipment and kept up-to-date, and report to NRC any malfunction of equipment that occurs due to any safety relief discharge.

  • The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report and/or its supplements wherein the license condition is discussed.

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(b) Prior to startup after the first refueling outage, the licensee shall replace or modify the NSSS equipment (intermediate range monitor, C51-K-601 A/H and two-inch air-operated globe valve, Cll-F0ll) if the results of the requalification tests indicate either change is required.

(10) Environmental Qualifications (Section 3.11, SER, SSER #1, SSER #2)

(a) No later than June 30, 1982, the licensee shall be in compliance with the provisions of NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", Revision 1, dated July 1981, for safety-related electrical equipment exposed to a harsh envi ro nment.

( b) No later than June 30, 1982, couplete and auditable records shall be available and maintained at a central location which describe the environnental qualification methods used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with NUREG-0588, Revision 1, dated July 1981. Such records shall be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified to document complete compliance.

(c) By June 30, 1982, the licensee shall complete the corrective actions stipulated in Appendix C to Supplement No. 2 of the Safety Evaluation Report.

(11) Seismic and Loss-of-Coolant Accident Loads (Section 4.2.3.4, SER, SSER #1, SSER #2)

(a) By July 30, 1982, the licensee shall submit to NRC a couplete description of the analytical methods, along with all analytical results, with regard to fuel assembly liftoff.

(b) Prior to startup after the first refueling outage, the fuel assembly liftoff issue must be satisfactorily resolved.

(12 ) Surveillance of Control Blade (Section 4.2.3.14, SER)

Within 30 days after plant startup following the first refueling outage, the licensee shall submit a written response on item 3 of the IE Bulletin No. 79-26, Revision 1.

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(13) Scram Discharge Volume (Sections 4.6.2, SER and 6.3.2.3, SSER #2)

Prior to startup after the first refueling outage, the' licensee shall incorporate the following additional modifications into the scram discharge volume system:

( a) Redundant vent and drain valves, and (b) Diverse and redundant scram instrumentation for each instrumented volume, including both delta pressure sensors and float sensors.

(14) Low Pressure in Pump Discharge of the Control Rod Drive (Section 4.6.2, SSER #2) .

Prior to startup after the first refueling outage, the licensee shall install instrumentation for an automatic scram that would shut down the reactor in the event of low control rod drive pump discharge pressure to be activated during startup and refueling modes only.

(15) Containment Long Term Program Load Specifications (6.2.1.1, SSER #2)

Prior to October 1,1982, the licensee shall submit its confirmatory assessment of the containment design adequacy for pool dynamic loads (chugging, vent lateral and diaphragm reverse pressure) developed in conjuction with the Long Term Program and reported in N UREG-0808.

(16) Pressure Interlocks on Yalves Interfacing at Low and High Pressure (Section 6.3.4, SSER #2)

Prior to startup after the first refueling outage, the licensec shall implement isolation protection against overpressurization .

of the low pressure energency core cooling systems (RHR/LPCI .

1 and LPCS) at the high and low pressure interface containing

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a check valve and a closed motor-operated valve.

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! (17) Compliance with Regulatory Guide 1.97 (Sections 7.5.2, SER)

By July 1,1982, the licensee shall submit a proposed implementation schedule for meeting Regulatory Guide 1.97 Revision 2, dated December 1980.

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(18) Additional Instrumentation and Control Concerns (Section 7.7.3.4, SSEA #1)

The licensee shall resolve' the following cencerns to the NRC staff's satisfaction prior to,startup after the first refueling outage:

( a) whether common electrical power sources or sensor malfunctions may cause multiple control systems failures, and (b) whether high. energy line breaks will result in unacceptable consequential control system failures.

(19) Low and/or Degraded Grid Voltage (Section 8.2.2.2, SER)

Prior to startup after the first refueling outage, the licensee shall install a second level of undervoltage protection (20) Reliability of Diesel-Generators (Sections 8.3.1.1, SER and 9.6.3.4, SER)

Prior to startup after the first refueling outage, the licensee shall implement the following design modifications with respect to diesel-generator reliability:

( 2avy duty turbocharger gear drive assecbly be installed on the diesel-generators.

(b) A prelube pump, powered from a reliable direct current power supply, be installed in the system to operate in parallel with the engine-driven lube. oil pump, or an alternative acceptable to the NRC shall,5e installed to preclude dry-starting of the diesel-engine.

(c) Controls and monitoring instrumentation be > removed from the engine and engine skid, except instruments qualified for this location. The non-qualified control and monitoring instruments shall be installed on a free standing floor mounted panel and located on a vibration free floor area.

- If the ficor is not vibration free, the panel shall be equipped wi th victation mounts.

(21) Direct Current Power Systems (Section 8.3.1.2, SER)

Prior to -tartup after the first refueling outage for the 125 and 250-volt direct current systems for Divisions is and 2 and the 125-volt Division 3 direct current system, the follbwing additional instrumentation shall be provided in the control, rohm: (1) Battery current (ammeter-charge / discharge), (2) Battery ' charger output 3'

voltage (voltmeter), (3) Battery charger output current (ammeter),

(4) Battery high discharge rate alarm, and (5) Battery charger trouble alarm. In the interim, the licensee shall implement approved procedures to monitor battery current, battery charger output voltage, and battery charger output current at the local panels at least once per eight hour shift.

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(22) Reactor Containment Electrical Penetrations (Section 8.4.1, 3ER)

Prior to startup after the first refueling outage, a redund:nt fault current device (circuit breakers or fuses) shall be provided on each penetrating circuit that would limit a fault current surge to be less than the surge for which the penetratisn is qualified except for low energy (milliamps) instrument systems.

(23) Separation of Class lE and Non-Class IE Cable Trays (Section 8.4.6.1, SER, SSER #1, SSER #2)

Prior to startup after the first refueling outage, the licensee shall provide adequate reparation or barriers between Class IE and adjacent non-Class lE cable trays.

(24) cire' Protection Program (Section 9.5, SER, SSER #2, SSER #3)

(a) The licensee shall maintain in ef fect and fully implement all provisions of the approved Fire Protection Plan. In addition, the licensee shall maintain the fire protection program set forth in Appendix R to 10 CFR Part 50, except for the following deviation::

(i) Hydrostatic hose tests in accordance with NEPA 1962-1979, and (ii) No automatic fire detection systems in areas 2K/3K and 5B4.

(b) Prior to startup after the first refueling outage, the licensee shall provide fire protection systems in fire sreas 2C/3C, 4C3 and 6E.

(c) Prior to startup after the first refueling outage, the licensee with respect to fire doors shall implement one of the foll owi ng:

(i) Perform an engineering review of the manufacturer's certified doors and door frames by a nationally recognized laboratory to certify that the door and door frames provide the required fire resistanc3 rating, or (ii) Test a replicate "as installed" door assembly oy a nationally recognized laboratory to determine the door rating, or (iii) Replace manufacturer's labeled doors and door franes with UL rated i tems.

( d) Prior to startup after the first refueling outage, the licensee sitall demonstrate the adequacy of its fire protection for record sto rage.

(25) Radiation / Chemistry Technicians on the Backshift (Section 13.1, SSER#2)

(a) Prior to initial criticality, all Radiation / Chemistry Technicians on the backshift shall be trained per the La Salle's Training Qualification Guide. All such Technicians shall also have satisfactorily completed the following energency response training:

(i) Tasks to be performed during the first 60 minutes of a serious energency on the backshift; (ii) Post-accident sanpling and analysis for the first three hours of an emergency; (iii) In-plant radiation surveys during an accident; (iv) Use and interpretation of both portable and fixed area radiation monitoring equipment, such as the Eberline PING-3 and SAM-2; (v) Interpretation of critical effluent monitoring data fo'r assisting the Chief Engineer during the first hour of an accident (i.e.,

station vent monitor and standby gas treatment monitor);

(vi) First aid and bioassay techniques; and (vii) Use of respiratory equipment during emergency situations.

(b) By June 1,1983, the licensee shall have Radiation / Chemistry Technicians onshift for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day who meet ANSI N18.1-1971 or who are qualified in accordance vd th a NRC approved alternative program.

(26) Industrial Security (Section 13.6, SER, SSER#3)

The licensee shall maintain in effect and fully implement all provisions of the Commission's approved physical security plan, guard training and qualification plan, and contingency plan, including anendments made pursuant to the authority of 10 CFR 50.54(p). The approved plans which contain safeguard information are collectively entitled: "La Salle County Station Security Plan Units 1 and 2," Revision 11, dated December 24,1981; "La Salle County Station Guard Training and Qualification Plan," submi tted by their letter dated August 16, 1979, as revised in August, 1980; and "La Salle Nuclear Power Station Contingency Plan, dated March,1980, as revised by pa9es dated June,1980.

The licensee is exempt from the commitment to fully implement those portions of the Security Plan as described in Items 1 and 2 in the licensee's letter dated April 1,1982, provided that the conpensatory measures delineated in the above referenced letter are in place.

Compensatory measures as descrioed in Item 3 in the April 1,1982 letter are approved with full implementation of the security plan conmi ttents to be accomplished no later than July 1,1982.

The 1icensee is exempted from the provisions of 10 CFR 73.55(d)(9),

but shall meet all other commitments of the Pnysical Security Plan and the following additional items.

(a) Change all keys, locks, and combinations and related equipment used to control access to protected areas and vital areas at least every 12 months.

(b) Issue keys,1ocks, combinations, and other access control devices to protected and vital areas only to those individuals who possess access authorization to those areas.

(c) Change keys, locks, combinations, and related equipment to which an individual had access within 5 days and immediately for card keys after access authorization is withdrawn due to lack of trustworthiness, reliability, or inadequate work performance.

(27) Initial Test Program (Section 14, SER)

The licensee shall conduct the post-fuel-loading initial test program (set forth in Section 14 of the licensee's Final Safety Analysis Report, as amended) without making any major modifications of this program unless modifications have been identified and have received prior NRC approval. Major modifications are defined as:

(a) Elimination of any test identified in Section 14 of the licensee's Final Safety Analysis Report, as amended as being essential; (b) Modification of test objectives, methods or acceptance criteria for any test identified in Section 14 of the licensee's Final Safety Analysis Report, as amended as being essential; (c) Performance of any test at a power level different from that described in the program; and

( d) Failure to complete any tests included in the described program l (planned or scheduled for power levels up to the authorized powar l evel ) .

(28) Assurance of Proper Design and Construction (Section 17.4, SSER #2)

Prior to exceeding 5 percent of full power, the licensee shall l

have conducted an independent review of the mechanical and structural l design of the loop C residual heat removal system, excluding I

all branch piping less than 3 inches, in the functioning mode

of the low pressure injection system using loads resulting from l

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the actuation of the automatic depressurization system in conjunction with the operating basis earthquake to verify that this system has been designed and constructed in accordance with all pertinent NRC requirements. This verification review shall consider design, installation, inspection, testing, and any other aspects necessary to ensure conformance with the design. This review shall be performed independently of the licensee and its contractors who perfomed design and construction activities for the La Salle County Station, and it shall be completed to the satisfaction of NRC.

(29) NUREG-0737 Conditions (Section 22.2)

The licensee shall cmplete the following conditions to the satisfaction of the NRC. These conditions reference the appro-priate items in Section 22.2, "TMI Action Plan Requirements for Applicants for Operating Licenses," in the Safety Evaluation Report and Supplements 1, 2 and 3, NUREG-0519.

(a) Shift Technical Advisor (I.A.1.1, SER, SSER #2)

The Shift Technical Advisor (STA) function shall be fulfilled by the Station Control Room Engineer (SCRE) who will be a designated SRO. However, if a SCRE is not available, the licensee shall provide a fully-trained on-shif t technical advisor to the shift engineer (shift supervisor).

(b) Nuclear Steam Supply System Vendor Review of Procedures (I.C.7, SER)

Prior to beginning low-power testing, the licensee shall assure that the General Electric review of the power-ascension test procedures has been cmpleted and the General Electric recommendations have been incorporated.

( c) Independent Safety Engineering Group (I.B.1.2, SER)

The licensee shall have an on-site independent engineering group.

(d) Control Room Design Review (I.D.1, SER, SSER #1)

The licensee shall correct the design deficiencies identified in Appendix C of Supplement No. I to the Safety Evaluation Report, NUREG-0519 on the schedule prescribed therein.

(e) Training During Low-Power Testing (I.G.I, SER, SSER #2)

At least 4 weeks prior to perfoming the Special Test, Simulated Loss of Onsite and Offsite Alternating-Current Power Test, the licensee shall provide a safety analysis for the test and its procedures to NRC for review and approval.

( f) Direct Indication of Safety / Relief Valve Position (II.D.3, SER, SSER #2)

Prior to startup after the first refueling outage, the licensee shall replace the safety / relief valve position indicator to a model that meets the IEEE Standards 323-1974 and 344-1975.

(g) Instrumentation for Detection of Inadequate Core Coolng (II.F.2, SER SSER #1, SSER #2)

By July 31, 1982, the licensee shall submit a report addressing the analysis perfomed by the BWR Owners Group regarding additional instrumentation relative to inadequate core cooling and the licensee shall implement the staff's requirenents after the cmpletion of the staff's review of this report on a schedule acceptable to the staff.

(h) Proper Functioning of Heat Removal Systems (II.K.1.22, SER, SSER #2, and II .K.3.13, SER, SSER #2)

Prior to startup after the first regueling outage, the licensee shall implement the logic to restart automatically the core isolation cooling system (1) Modify Break Detection Logic to Prevent Spurious Isolation of l High Pressure Coolant Injection and Reactor Core Isolation Cooling System (II .K.3.15, SER, SSER #2 )

i Prior to startup after the first refueling outage, the licensee shall implement a circuit modification to assure

( that transients monitored by pressure instruments to sense flow in these two systems actually sense continuous high steam fl ow.

(j) Modification of Automatic Depressurization System Logic -

Feasibility for increased Diversity for Some Event Sequences (II .K.3.18, SER, SER #1, SSER #3)

(a) By October 1,1982, the licensee shall evaluate the alternative design modifications of the BWR Owners Group relative to the logic for the automatic depressurization system, submit such evaluation, and propose modifications to NRC for review and approval .

(b) Prior to startup after the first refueling outage, the licensee shall implement the approved alternative logic  !

modification of the automatic depressurization system.

(k) Restart of Core Spray and Low Pressure Core Injection System (ll .K.3.21, SER, SSER #2 )

Prior to startup after the first refueling outage, the licensee shall provide an auto start for the high pressure core spray.

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(1) Automatic Switchover of Reactor Core Isolation Cooling System Suction--Verify Procedures and Modify Design (II.K.3.22, SER)

Prior to startup after the first refueling outage, the licensee shall imiilement the automatic switchover of the reactor core isolation cooling system suction from the condensate storage tank to the suppression pool when the condensate storage tank level is low. ,

(m) Upgrade Emergency Support Facilities (III.A.l.2, SER, SSER #1)

The licensee shall complete its Emergency Response Facilities as follows:

(1) Safety Parameter Display System October 1, 1982 (ii) Emergency Operations

. Facili ty October 1,1982 6

(iii) Technical Support Center * '

October 1,1982 D. Exemptions from certain requirements of Appendices G, H, J, and $50.55(a) to 10 CFR Part 50 and 10 CFR Part 73 are described in the Safety Evaluation

! Report and Supplement No.1, No. 2 and No. 3 to the Safety Evaluation Report.

In addition, an exemption was requested until the completion of the first refueling from the requirements of 10 CFR $70.24 and an exemption from 10 CFR Part 50, Appendix E from perfonning a full scale exercise within

! one year before issuance of an operating license, both exemptions are described in Supplement No. 2 of the Safety Evaluation Report. These exemptions are authorized by law and will not endanger life or property or the common defense and security and are otherwise in the public interest. Therefore, these exemptions are hereby granted. The facility will operate, to the extent authorized herein, in confonnity with the application, as amended, and the rules and regulations of the Commission (except as hereinafter exempted therefrom), and the provisions of the Act.

E. This license is subject to the following additional condition for the protection of the enviromient:

Before engaging in additional construction or operational activities which may result in a significant adverse environmental impact that was not evaluated or that is significantly greater than that evaluated in the Final Environmental Statement and its Addendum, the licensee shall provide a written notification to the Director of the Office of Nuclear Reactor Regulation and receive written approval from that office before proceeding wi th such activities.

F. The licensee shall report any violations of the requirements containeo in Section 2, Items C(1), C(3) through (29), and E of this license within twenty-four (24) hours by telephone and confirmed by telegram, mailgram, or facsimile transmission to the NRC Regional Administrator, Region III, or designee, not later than the first working day following the violation, with a written followup report within fourteen (14) working days.

G. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Se'ction 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H. This license is effective as of the date of issuance and shall expire September 10, 2013, pending the staff's satisfactory review of the Licensee's request for a forty (40) year license.

FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation

c' l Attachment.

I 1. Attachment 1 l

2. Appendix A - Technical

/ ' Specifications (NUREG-0861) l 3. Appendix B - Environmental l Protection Plan l

l Date of Issuance:

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E. This license is subject to the following additional condition for the protection of the envirorrnent:

Before engaging in additional construction or operational activities which may result in a significant adverse environmental impact that was not evaluated or that is significantly greater than that evaluated in the Final Environmental Statement and its Addendum, the licensee shall provide a written notification to the Director of the Office of Nuclear Reactor Regulation and receive written approval from that office before proceeding wi th such activities.

F. The licensee shall report any violations of the requirements contained in Section 2, Items C(1), C(3) through (29), and E of this license within twenty-four (24) hours by telephone and confinned by telegram, mailgram, or facsimile transmission to the NRC Regional Administrator, Region III, or designee, not later than the first working day following the violation, with a written followup report within fourteen (14) working days.

G. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

H. This license is effective as of the date of issuance and shall expire Septenber 10, 2013, pending the staff's satisfactory review of the Licensee's request for a forty (40) year license.

FOR THE NUCLEAR REGULATORY COMMISSION Harold R. Denton, Director Office of Nuclear Reactor Regulation

Attachment:

j 1. Attachment 1

2. Appendix A - Technical Specifications (NUREG-0861)
3. Appendix B - Environmental Protection Plan Date of Issuance:

ATTACHMENT 1 TO LICENSE NPF-11 This attachment identifies certain preoperational tests, system demonstrations and other items which must be completed to the Commission's satisfaction prior to proceeding to Operational Mode 2 (initial criticality or-212tF. as appli-cable). The licensee shall not proceed beyond this Operational Mode without written confirmation from NRC that the following items have been completed in accordance with the conditions set forth below.

1. The following Preoperational Tests shall be completed, including all reviews:
a. Containment Monitoring System (PT-CM-101)
b. Drywell Pneumatic System (PT-IN-101)
c. Traversing Incore Probe (PT-NR-102) (Prior to Entering Test Condition 1)
d. Off-Gas System (PT-0G-101)
e. Primary Containment ILRT (PT-PC-101)

, f. Pipe Vibration Monitoring (PT-SI-102)

2. The following System Demonstrations shall be completed, including all reviews:
a. High Radiation Samplirg System (SD-PS-102)
b. Dynamic Effects (50-51-101)
3. Commonwealth Edison Company must install and test the microwave voice channel communications system. (373/81-14-25)
4. Commonwealth Edison Company must install radiation measurement i equipment in the EOF. (373/81-14-31)
5. Commonwealth Edison Company must assure the following commitments per TMI Action Plan Requirement III.D.3.3 are met:
a. Availability of 5 PING-3 (2A special) particulate, iodine, i and noble gas air monitoring systems mounted on carts.

l l b. Availability of Eberline Instrument Corporation SAM-2 lodine monitors with silver zeolite cartridges.

l l c. Availability of a low background, low contamination area for analyzing iodine cartridges. (373/81-00-102) 1 t

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6. Commonwealth Edison Company must assure that results from Preopera-tional Test PT-RP-101 for response time of the scram signals for the turbine stop valve and turbine control valve fast closure are added to the results obtained in Preoperational Test PT-RR-101 to obtain correct scram response times for these items. (373/80-15-12)
7. Commonwealth Edison Compar.y must conduct a site assembiy drill.

(373/80-53-03)

8. Commonwealth Edison Company must include in Startup Test Pro-cedure STP-17 requirements for measuring GAP clearances between I the process pipe and the surrounding pipe whip restraint struc-tural assemblies. (373/81-29-01)
9. Commonwealth Edison Company must develop procedures for esti-mating Noble Gas Radioiodine Release Rates required by Table II.F.1-2 per TMI Action Plan Requirement II.F.1. (373/81-00-104)
10. Commonwealth Edison Compahy must siti'sfactorily resolve those deficiencies affecting systems for which Preoperational Tests or System Demonstrations are required to be completed prit- to initial criticality. (Category 4 deficiencies as per the licensee's procedure).
11. Commonwealth Edison Company must review the Cable Pan Loading Report to verify there are no outstanding discrepancies.

(373/79-34-01)

12. Commonwealth Edison Company must calibrate high range station vent and standby gas treatment system monitors. (373/81-14-20)
13. Commonwealth Edison Company must complete modifications to the laboratory HVAC system to meet design criteria requirements of positive pressure between the counting room and surrounding areas.

(373/81-43-03) i 14. Commonwealth Edison Company must review the stop and control valves

\ closure time acceptance criteria basis to determine it meets design

>' specifications. (373/81-43-06) j 15. Commonwealth Edison Company must conduct a 100% reinspection l of high strength steel bolting. (373/81-48-06)

/

16. Commonwealth Edison Company must develop a written technical basis for the practice of installing high strength bolts without torquing requirements. (373/81-48-07)
17. Commonwealth Edison Company must complete a testing program for vibration monitoring of the Low Pressure Core Spray 1A and Residual Heat Removal System 18 pumps. (373/82-10-15 and 373/82-18-01) 2

.

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