ML20236F789

Amends 128 & 113 to Licenses NPF-11 & NPF-18,respectively, Revising TS Section 6, Administrative Controls, to Incorporate Revised Organizational Titles & Delete Unit 1 FOL Conditon 2.C.(30)(a)

ML20236F789
Person / Time
Site:	LaSalle
Issue date:	06/26/1998
From:	Skay D NRC (Affiliation Not Assigned)
To:
Shared Package
ML20236F792	List: ML20236F789 ML20236F794
References
NUDOCS 9807020344
Download: ML20236F789 (43)

Text

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y UNITED STATES NUCLEAR REGULATORY COMMISSION E

E WAsHIMToN, D.C. 20066-0001

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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASALLE COUNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 128 License No. NPF-11 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee) dated May 27,1997, as supplemented on August 1,1997, and March 24, 1998, 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 ChapterI; B.

The facility will operate in conformity wi% the application, the provisions of the Act, and the 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 set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraphs 2.C.(2) and 2.C.(30)(a) of Facility Operating License' No. NPF-11 are hereby amended to read as follows:

' License page 12 is attached, for convenience, for the composite license to reflect this change.

9807020344 900626 PDR ADOCK 05000373 P

PDR

T I

l-2-

2.C.(2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.128, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

2.C.(30)(a) DELETED 3.

This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION I

MY Donna M. Skay, Project nager Project Directorate lil-2 Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation Attachments.

1. License page 12-l

2. Changes to the Technical Specifications l

Date of Issuance: June 26, 1998 I

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__.-__m______-.--_

i y.

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 performed design and construction activities for the La Salle County Station, and it shall be

. completed to the satisfaction of NRC.

(30)

NUREG-0737 Conditions (Section 22.2)

The licensee shall complete the following conditions to the satisfaction of the NRC. These conditions reference the appropriate 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)

DELETED (b)

Nuclear Steam Sucolv System Vendor Review of Procedures (l.C.7. SER)

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

(c)

Independent Safety Enoineerina Group (l.B.1.2. SER)

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

(d)

Control Room Desian Review (l.D.1. SER. SSER #2)

The licensee shall correct the design deficiencies identified in Appendix C of Supplement No.1 to the Safety Evaluation Report, NUREG-0519 on the schedule prescribed therein.

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Amendment No.128

_ _ _ _ - _ _ _ _ - _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - - _ _ - _ - _ _ - _ _ _ _ _ _ _ _. - _ _ - - - _ _ _ _ _ _ _ = - _ _ _ _

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ATTACHMENT TO LICENSE AMENDMENT NO. 12R FACILITY OPERATING LICENSE NO. NPF-11 DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a verticalline indicating the area of change.

REMOVE INSERT XIV XIV XVil XVil 1-4 1-4 5-4 5-4 6-1 6-1 6-2 6-2 6-2a 6-13 6-13 6-14 6-14 6-15 6-15 6-17 6-17 6-18 6-18 6-21 6-21 6-24 6-24 6-25 6-25 6-26 6-26 6-27 6-27

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JEEX BASES SECTION EA_G_E 1

l 3/4.5 EMERGENCY CORE COOLING SYSTEMS l

3/4.5.1 and 3/4.5.2 ECCS-0PERATING and SHUTDOWN.............. B 3/4 5-I 3/4.5.3 SUPPRESSION CHAMBER................................

B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity...................... B 3/4 6-1 Primary Containment Air Locks...................... B 3/4 6-la Drywell and Suppression Chamber Internal Pressure.. B 3/4 6-2 Drywell Average Air Temperature.................... B 3/4 6-2 Drywell and Suppression Chamber Purge System....... B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS...........................

B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES...............

B 3/4 6-4 3/4.6.4 VACUUM RELIEF...................................... B 3/4 6-4a 3/4.6.5 SECONDARY CONTAINMENT.............................. B 3/4 6-5 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL............. B 3/4 6-5 LA SALLE - UNIT I XIV Amendment No.128

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JRQEX DESIGN FEATURES SECTION Egg l

5.1 SITE l

Exclusion Area..................................................

5-1 Low Popul at i on Zone.............................................

5-1 Si te Boundary for Gaseous Effl uents.............................

5-1 Si te Boundary for Liquid Effl uents..............................

5-1 5.2 CONTAINMENT Configuration...................................................

5-1 Design Temperature and Pressure.................................

5-1 Secondary Containment...........................................

5-1 5.3 REACTOR CORE F u el A s s embl i e s.................................................

5-4 Control Rod Assemblies..........................................

5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature.................................

5-4 Vo1ume..........................................................

5-4 5.5 DELETED..........................................................

5-4 i

5.6 FUEL STORAGE Criticality.....................................................

5-5 Drainage........................................................

5-5 Capacity........................................................

5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT..............................

5-5 i

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LA SALLE - UNIT 1 XVII Amendment No.128

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s DEFINITIONS 1.20 DELETED LIMITING CONTROL ROD PATTERN 1.21 A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the j

core being on a thermcl hydra.ulic limit, i.e.,

operating on a limiting i

value for APLHCR, LHGR, or MCPR.

iTNEAR HEAT GENERATION RATE I

1.22 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit length of fuel rod.

It is the integral of the heat flux over the heat transfer area associated with the unit length. LHCR is monitored by the ratio of LHOR to its fuel specific limit, as specified in the CORE OPERATING LIMITS REPORT.

LOGIC SYSTEM FUNCTIONAL TEST 1.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e, all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit, from sensor through and including the actuated device to verify OPERABILITY. THE LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

1.24 DELETED MEMBERSfS) OF THE PUBLIC 1.25 MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the l

licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE. CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall l

also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification section 6.2.F.4 and (2) descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual l

Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

Lk SALLE UNIT 1 1-4 Amendment No. 128 l

DESIGN FEATURES g

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO ) as fuel material. The 2

bundles may contain water rods or water boxes. Limited substitutions of Eircalloy or IIRLO or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

CONTROL ROD ASSEMBLIES l

5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B C) and/or 4

hafnium metal. The control rod assembly shall have a nominal axial absorber f

length of 143 inches.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

In accordance with the code requirements specified in Section 5.2 a.

of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2.

1650 peig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3.

1500 psig from the discharge shutoff valve to the jet pumps.

c.

For a temperature of 575*F.

VOLUME l

5.4.2 The total water and steam volume of the reactor vessel and recirculation system is - 21,000 cubic feet at a nominal T,of 533*F.

1 5.5 DELETED LA SALLE - UNIT 1 5-4 Amendment No.128

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6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION A.

Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organiza-tions shall include the positions for activities affecting the safety of the nuclear power plant.

1.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descrip-tions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Manual.

2.

The individual filling the ANSI N18.1-1971 Section 4.2.1 position of Flant Manager (" Plant Manager"), shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

3.

The Chief Nuclear Officer (CNO) shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

4.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

B.

The Shift Manager shall be responsible for directing and commanding the overall operation of the facility on his shift. The primary manage-ment responsibility of the Shift Manager shall be for safe operation of the nuclear facility on his shift under all conditions.

'C.

The shift manning for the station shall be as shown in Figure 6.1-3.

LA SALLE - UNIT 1 6-1 Amendment No.128

's ADMINISTRATIVE CONTROLS 1.

At lesst one licsnssd Roactor Oparator shall b3 in the control room when fuel is in the reactor. In addition, while the reactor is in OPERATIONAL CONDITION 1,2 or 3, at least one licensed Senior Reactor Operator who has been designated by the Shift Manager to assume the control room direction responsibility shall be in the Control l

Room.

2.

A radiation protection technician

• shall be on site when fuel is in the reactor.

3.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

4.

DELETED 5.

The Independent Safety Engineering Group (ISEG) shall function to examine unit operating characteristics, NRC issuances, industry advisories, Ucenseo Event Reports and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving unit safety.

The ISEG shall be composed of at least three, dedicated, full-time engineers of multi-disciplines located on site and shall be augmented on a part-time basis by personnel from other parts of the Commonwealth Edison Company organization to provide expertise not represented in the group. The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification # that these activities are performed correctly and that human errors are reduced as much as practical. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities or other means of improving unit safety to the Manager of Quality and Safety Assessment and the Plant l

Manager.

6.

The Shift Technical Advisor shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

• The radiation protection technician position may be less than the minimum requirement for a period of time not to exceed two hours in order to accommodate unexpected absence provided immediate action is taken to fill the required position.

1. Not responsible for sign-off feature.

LA SALLE - UNIT 1 6-2 Amendment No. 128

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l FIGURE 6.1-3 MINIMUM SHIFT CREW COMPOSITION"*3 POSITION 3 MINIMUM CREW NUMBER EACH UNIT IN ONE UNIT IN EACH UNIT IN i

CONDITION 1,2, OR 3 CONDITION 1,2, OR 3, CONDITION 4 OR 5 AND ONE UNIT IN '

OR DEFUELED CONDITION 4 OR 5 OR DEFUELED SM 1

1 1

SRO 1

1 None RO 3

3 2

AO 3

3 3

STA")

1 1

None (a)

This table reflects the total requirements for shift staffing of both units.

With the exception of the Shift Manager, the shift crew composition may be one less than the minimum requirements of Figure 6.13 for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to accommodate unexpected absence of on duty shift crew members, provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Figure 6.1-3. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(b)

Table Notation:

.SM Shift Manager with a Senior Reactor Operator license for each unit whose reactor contains fuel.

SRO Individual with a Senior Reactor Operator license for each unit whose reactor contains fuel.

- During CORE ALTERATIONS on either unit a licensed SRO or licensed SRO limited to fuel handling, who has no other concurrent responsibilities, must be present to observe and directly supervise this operation.

RO An Individual with a Reactor Operator license or a Senior Reactor Operator license for unit assi,gned. At least one RO shall be assigned to each unit whose reactor contains fuel.

Individuals acting as relief operators shall hold a license for both units. Otherwise, for each unit, provide a relief operator who holds a license for the unit assigned.

AO At least one auxiliary operator shall be assigned to each unit whose reactor contains fuel.

STA Shift Technical Advisor.

(c)

While either unit is in CONDITION 1,2, or 3, an individual with a valid SRO license shall be designated to assume the control room command function. With both Units in CONDITION 4 or 5, an individual with a valid SRO or RO license shall be designated to assume the control room command function, i

(d)- The STA position shall be filled by an individual who meets the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

l LA SALLE-UNIT 1 6-13

~ Amendment No. 128 1

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.IA'SALLE - UNIT.1 6-14 Amendment No. 128

---___________:____-_____----______-__--____-L_--__---____-__--____-.__:_____----..

O, ADMINISTRATIVE CONTROLS l

6 6.1.1 HIGH RADIATION AREAS 1

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6.1.1.1 Pursuant to Paragraph 20.203(c)(5) of 10 CFR 20, in lieu of the f

" control device" or " alarm signal" required by paragraph 20.203(c)(2) of i

10 CFR 20, each hiyh radiation area in which the intensity of radiation is greater than 100 mrom/hr* but less than 1000 mram/hr* shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

Individuals l

qualified in radiation protection procedures, or personnel continuously escorted I

by such individuals, may be exempt from the RWP issuance requirement during the I

performance of their assigned duties in high radiation areas in which the j

j intensity of radiation As greater than 100 mres/hr* but less than 1000 mrom/hr*,

provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the followings A radiation monitoring device which continuously indicates the a.

radiation dose in the area, b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

c.

A health physics qualified individual, i.e.,

qualified in radiation protection procedures, with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Health Physicist in the Radiation Work Permit (RWP).

6.1.1.2 In addition to the requirements of 6.1.1.1, above, for areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrom*, the computer shall be programmed to permit entry through locked doors for any individual requiring access to any such High-High Radiation Areas for the time that access is required.

6.1.1.3 Keys to manually open computer controlled High Radiation Area doors and High-High Radiation Area doors shall be maintained under the Administra-tion control of the shift Manager on duty and/or the Health Physicist.

l 6.1.1.4 High-High Radiation areas, as defined in 6.1.1.2 above, not equipped with the computerized card readers shall be maintained in accordance with l

10 CFR 20.203 c.2 (iii), locked except during periods when access to the area l

1s required with positive control over each individual entry, or 10 CFR 20.203.c.4.

In the case of a High Radiation Area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be substituted.

Doors shall remain locked except during periods of access by personnel under an l

approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area. For LA SALLE UNIT 1 6-15 Amendment No.128

. ADMINISTRATIVE CGnimOLS PLANT OPERATING PROCEDURES AND PROGRAd5 (Continu2d) 3.

Radiation control procedures shall be maintained, made available to all station personnel, and adhered to.

These procedures shall show permissible radiation exposure and shall be consistent with the requirements of 10 CFR 20.

This radiation protection program shall be organized to meet the requirements of 10 CFR 20.

C.

TECHNICAL REVIEW AND CONTROL Procedures required by specification 6.2.A and 6.2.3 and other procedures g

which affect nuclear safety, as determined by the Plant Manager, and l

changes thereto, other than editorial or typographical changes, shall be reviewed as follows prior to beplementation except as noted in specification 6.2.D:

1.

Each procedure or procedure change shall be independently reviewed by a qualified individual knowledgeable in the area affected other than the individual who prepared the procedure or procedure change. This review shall include a determination of whether or not additional cross-disciplinary reviews are necessary.

If deemed necessary, the reviews shall be performed by the qualified review personnel of the appropriate discipline (s).

2.

Individuals performing these reviews shall meet the applicable experience requirements of ANSI N18.1-1971, sections 4.2, 4.3, 4.4, 4.5.1, or 4.6, and be approved by the Plant Manager.

3.

Applicable Administrative Procedures recommended by Regulatory Guide 1.33, Plant Emergency Operating Procedures, and changes thereto shall be submitted to the Onsite Review and Investigative Function for review and approval prior to implementation.

4.

Review of the procedure or procedure change will include a determination of whether or not an unreviewed safety question is involved. This determination will be based on the review of a written safety evaluation prepared by a qualified individual or documentation that a safety evaluation is not required. Onsite Review, Offsite Review and Commission approval of items involving unreviewed safety questions shall be obtained prior to station approval for implementation.

5.

The Department Head approval authority shall be specified in station procedures.

6.

Written records of reviews performed in accordance with this specification shall be prepared and maintained in accordance with specification 6.5.

7.

Editorial and Typographical changes shall be made in accordance with station procedures.

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LA SALLE - UNIT 1 6-17 Amendment No.128

e

' ADMINISTRATIVE CONTROLS D.

Temporary chang 30 to proceduros 6.2.A cnd 6.2.5 cbova cay be mada previded:

l 1.

The intent of the original procedure is not altered.

j 2.

The change is approved by two members of the plant management staff, at l

least one of whom holds a Senior Reactor Operator's License on the unit affected.

3.

The change is documented, reviewed and approved in accordance with specification 6.2.C. within 14 days of implacentation.

E.

Drills of the emergency procedures described in Specification 6.2.A.d shall be conducted at frequencies as specified in the Generating Stations Emergency Plan (GSEP). These drills will be planned so that during the course of the year, communication links are tested and outside agencies are contacted.

I F.

The following programs shall be established, implemented, and maintained 1.

Primary coolant Sources outside Primary Containment i

A program to reduce leakage from those portions of systems outside primary containment that could contain highly radioactive fluida during a serious transient or accident to as low as practical levels. The systems include LPCS, HPCS, RHR/LPCI, RCIC, hydrogen recombiner, process sampling, containment monitoring, and standby gas treatment systems.

The program shall include the followings a.

Preventive maintenance and periodic visual inspection requirements, and b.

Integrated leak test requirements for each system at refueling cycle intervals or less.

2.

In-Plant Radiation Monitorino

{

l A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

l This program shall include the following:

I a.

Training of personnel, b.

Procedures for monitoring, and i

Provisions for maintenance of sampling and analysis equipment.

c.

3.

Post-accident Samolina A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulate in plant gaseous effluents, and containment atmosphere samples under accident conditions.

The program shall include the followings a.

Training of personnel, b.

Procedures for sampling and analysis, c.

Provisions for maintenance of sampling and analysis equipment.

i l

LA SALLE - UNIT 1 6-18 Amendment No. 128 1

a ADMINISTRATIVE CONTROLS 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pursuant to Specification 2.1.1, 2.1.2 and 2.1.3 and critical reactor operation shall not be resumed until authorized the NRC. The conditions of shutdown shall be promptly reported to the Site ce President or his designated alternate. The incident shall be reviewed by the Onsite and Offsite Review and Investigative Functions and a separate Licensee Event Report for each occurrence shall be prepared in accordance with Section 50.73 to 10 CFR Part 50.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Site Vice President and the Director of Safety Review shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6.5 PLANT OPERATING RECORDS A.

Records and/or logs relative to the following items shall be kept in a manner convenient for review and shall be retained for at least 5 years:

1.

Records of normal plant operation, including power levels and periods of operation at each power level; 2.

Records of principal maintenance and activities, including inspection and repair, regarding principal items of equipment pertaining to nuclear safety; 3.

Records and reports _of reportable events; 4.

Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (failing to meet see Section 4 of these specifications are being met. All equipment surveillance re)quirements and the corrective action taken shall be recorded; 5.

Records of changes to' operating procedures; 6.

Shift Manager logs; and i

7.

Byproduct material inventory records and source leak test results.

l i

B.

Records and/or logs relative to the following items shall be recorded in a manner convenient for review and shall be retained for the life of the plant:

1.

Substitution or replacement of principal items of equipment pertaining i

to nuclear safety; 2.

Changes made to the plant as it is described in the SAR; i

l 3.

Records of new and spent fuel inventory and assembly histories; 4.

Updated, corrected, and as-built drawings of the plant; 1

LA SALLE - UNIT 1 6-21 Amendment No. 128

e ADMINISTRATIVE CONTROLS' perfermed prior to cxceeding th3 liait, reculto of cnnlysio whilo limit was exceeded and results of one analysis after the radiciodine activity was reduced to less than limit. Each result should include date and time of sampling and the radiciodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the ILmit was exceeded; (4) Graph of the I-131 concentration and one other radiciodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

3.

Annual Radioloalcal Environmental operatino Report

• The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period.

The material provided shall be consistent with the objectives outlinea in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

4.

Annual Radioactive Effluent Release Report **

The Annual Radioactive Effluents Release Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the oDCM and PCP and (2) in conformance with 10 CFR 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

5.

Monthlv Doeratino Report Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basis to the addresses specified in 10 CFR 50.4 no later than the 15th of each month following the calendar month covered by the report.

l A single submittal may be made for a multi-unit station.

A single submittal may be made for a multi-unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

LA SALLE UNIT 1 6-24 Amendment No.128

ADMINISTRATIVE CONTROLS Monthly Goeratino Reoort (Continued)

A report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

6.

Core Ooeratino Limits Reoort a.

Core o)erating limits shall be established and documented in the CORE 0)ERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

h (1)

The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

(2)

The minimum Critical Power Ratio MCPR limits, and power and flow de(MCPR) scram time, dependent pendent MCPR limits for Technical Specification 3.2.3.

Effects of analyzed equipment out of service are included.

(3)

The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4.

(4)

The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6-2.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

For LaSalle County Station Unit 1, the topical reports are:

(1)

ANFB Critical Power Correlation, ANF-ll25(P) Corporation, Supplements 1 and 2, Advanced Nuclear Fuels (A) and April 1990.

(2)

Letter, Ashok C. Thadani (NRC to R.A. Cpeland (SPC),

BWR Fuel Design," July 28,g of) ULTRAFLOW

" Acceptance for Referencin Spacer on 9x9-IX/X 1993.

(3)

Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Eeactors Critical Power Met.iodology/ Advanced Nuclear Fuels Corporation for Boiling Water Reactors:

Methodology for Analysis of Assembly Channel Bowing Effects Supplem/NRC Correspondence, XN-NF-524(P)(A) Revision 2, and ent 1 Revision 2, Supplement 2, Advanced Nuclear Fuels Corporation, November 1990.

(4)

COTRANSA 2: A Computer Program for Boiling Water Reactor Transient Analysis, ANF-913(P)(A), Volume 1, Revision 1 and Volume 1 Supplements 2, 3, and 4, Advanced Nuclear Fuels Corporation, August 1990.

LA SALLE UNIT 1 6-25 Amendment No.128

s ADMINISTRATIVE CONTROLS c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis are met.

d.

The OORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory Commission Document control Desk with copies to the Regional Administrator and Resident Inspector.

B.

Deleted C.

Unique Reporting Requirements 1.

special Reports shall be submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.7 PROCESS CONTROL PROGRAM iPCPl*

{

6.7.1 The PCP shall be approved by the Commission prior to implementation.

6.7.2 Licensee Jnitiated changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.18.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s),

and l

2)

A determination that the change will maintain the overall con-i formance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.

Shall become effective upon review and acceptance by the Onsite Review and Investigative Function.

• The Process Control Program (PCP) is common cp Cai'c 1 and La salle

.x Unit 2.

LA SALLE UNIT 1

+ M.

Amendment No.128

ADMINISTRATIVE CONTROLS 6.8 0FFSITE DOSE CALCULATION MANUAL fODCMi*

6.8.1 The ODCM shall be approved by the Commission prior to implementation.

6.8.2. Licensee initiated changes to the 00CM:

Shall be documented and records of reviews performed shall be retained a.

as required by Specification 6.5.B.18.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s), and 2)

A determination that the change will maintain the level of radio-active effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after review and acceptance by the On-Site Review and Investigative Function and the approval of the Plant Manager on the date specified by the On-Site Review and Investigative Function.

Shall be submitted to the Commission in the form of a complete, legible c.

copy of the entire ODCM as a > art of or concurrent with the Annual l

Radioactive Effluent Release leport for the period of the report in which any change to the 00CM was made effective.

Each change shall be identified by markings in the margin of the affected pages clearly indicatingtheareaofthepagethatwaschanged,andshallindicatethe date (e.g., month / year) the change was implemented.

6.9 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6.9.1 Licensee initiated major chan systems (liquid, gaseous and solid):ges to the radioactive waste treatment Shall be reported to the Commission in the Monthly Operating Report for a.

the period in which the evaluation was reviewed by the Onsite Review and Investigative Function. The discussion of each change shall contain:

l '.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59; 2.

Sufficient detailed information to totally support the reason for the change without benefit or additional or supplemental information; 3.

A detailed description of the equipment,lant systems; components and processes involved and the interfaces with other p

• The OFFSITE DOSE CALCULATION MANUAL (ODCM) is common to La Salle Unit I and i

La Salle Unit 2.

t l

LA SALLE UNIT 1 6-27 Amendment No. 128

pau4 9

g UNITED STATES j

NUCLEAR REGULATORY COMMISSION l

i 2

WASHINGTON, D.C. enmas anni

\\...../

i l

COMMONWEALTH EDISON COMPANY l

DOCKET NO. 50 374 LASALLE COUNTY STATION UNIT 2 j

AMENDMENT TO FACILITY OPERATING LICENSE i

Amendment No.113

' License No. NPF-18 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment filed by the Commonwealth Edison Company (the licensee) dated May 27,1997, as supplemented on August 1,1997, and l

March 24,1998, 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; B.

The facility will operate in conformity with the application, the provisions of the Act, and the 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 ectivities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:

l

. (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.114 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION I

med Y-Donna M. Skay, Project M ager Project Directorate ill-2 Division of Reactor Projects - Ill/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: June 26,1998 l

a

ATTACHMENT TO LICENSE AMENDMENT NO. 113 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a verticalline indicating the area of change.

REMOVE INSERT XIV XIV XVil XVil 1-4 1-4 3/4 3-39 3/4 3-39 5-4 5-4 6-1 6-1 6-2 6-2 6-2a 6-3 6-3 6-13 6-13 6-14 6-14 6-15 6-15 6-17 6-17 6-18 6-18 6-21 6-21 6-22 6-22 6-23 6-23 6-24 6-24 6-26 6-26 6-27 6-27 6-28 6-28 I

e

INDEX BASES SECTION PAGE l

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 and 3/4.5.2 ECCS-OPERATING and SHUTDOWN..............

B 3/4 5-1 3/4.5.3-SUPPRESSION CHAMBER................................

B 3/4 5-4 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity......................

B 3/4 6-1 Primary Containment Air Locks......................

B 3/4 6-2 Drywell and Suppression Chamber Internal Pressure.. B 3/4 6-2a j

Drywell Average Air Temperature....................

B 3/4 6-2a Drywell and Suppression Chamber Purge System.......

B 3/4 6-2a 3/4.6.2 DEPRESSURIZATION SYSTEMS...........................

B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES...............

B 3/4 6-4 j

3/4.6.4 VACUUM RELIEF......................................

B 3/4 6-4a 3/4.6.5 SECONDARY CONTAINMENT..............................

B 3/4 6-5 i

3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL.............

B 3/4 6-5 l

1 LA SALLE - UNIT 2 XIV Amendment No.113

l l

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INDEX DESIGN FEATURES SECTION PAGE 5.1 SITE Exc l u s i o n A re a........................................................ 5 - 1 Low Population Zone.....................................................

5-1 Site Boundary for Gaseous Effluents.................................... 5-1 Site Bounda ry for Liquid Effluents.....................................

5-1 5.2 CONTAINMENT Configuration......................................................... 5-1 i

Design Temperature and Pressure................................... 5-1 Seconda ry Cont a i nnent.............................................

. 5-1 5.3 REACTOR CORE f uel As s embl i es.................................................. 5 - 4 Control Rod Assemblies.................

.. 5-4 5.4 REACTOR COOLANT SYSTEM Design Pressure and Temperature........................................ 5-4 Vo1ume..................................................................

5-4 1

5.5 DELETED...........................................................

. 5-4 5.6 FUEL STORAGE Criticality......................................................

5-5 Drainage.........................................................

. 5-5 Capacity...............

. 5-5 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT........

............................ 5-5 LA SALLE - UNIT 2 XVil Amendment No.113 l

l

DEFINITIONS l

LIMITING CONTROL R0D PATTERN I

1.21 A LIMITING CONTROL R00 PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting value for APLHGR, LHGR, or MCPR.

i LINEAR HEAT GENERATION RATE 1.22 LINEAR HEAT GENERATION RATE (LHGR shall be the heat length of fuel rod.

Itistheinegraloftheheatf$eneration!erunit ux over th heat transfer area associated with the unit length.

LHGR is monitored by the i

ratio of LHGR to its fuel specific limit, as specified in the CORE l

OPERATING LIMITS REPORT.

LOGIC SYSTEM FUNCTIONAL TEST 1.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, i.e, all relays and contacts, all trip units, solid state logic elements, etc. of a logic circuit from sensor through and including tfie actuated device to verify OPERABILITY.

THE LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total system steps such that the entire logic system is tested.

1.24 Deleted I

MEMBERS (S) 0F THE PUBLIC 1.25 MEMBER (S)d with the plant.0F THE PUBLIC shall include all persons who are not o associate This category coes not include employees of the licensee, its contractors, or vendors. Also excluded from this categor persons who enter the site to service equipment or to make deliveries. y are This category does include persons who use portions of the site for recreational, l

occupational, or other purposes not associated with the plant.

MINIMUM CRITICAL POWER RATIO 1.26 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.

OFFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (0DCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm / Trip Set)oints, and in the conduct of the Environmental Radiological Monitoring )rogram. The ODCM shall also contain (Monitoring Programs required by Technical Specification

1) the Radioactive Effluent Controls and Radiological Environmental Section 6.2.F.4 and (2) Radiological Environmental Operating and Annual descriptions of the information that should be included in the Annual Radioactive Effluent Release Reports required by Technical Specification Sections 6.6.A.3 and 6.6.A.4.

LA SALLE - UNIT 2 1-4 Amendment No.113

INSTRUMENTATION END-OF-CYCLE RECIRCULAT70N PUMP TR7P SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instruments-tion channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

a.

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Satpoint value.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channelo per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With the number of OPERABLE channels two or more less than required c.

by the Minimum OPERABLE Channels per Trip System requirement (s) for one trip system and:

1.

If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, otherwise, either:

1.

Increase the MINIMUM CRITICAL POWER (MCPR) Limiting Condition for Operation (LCO) to the EOC-RPT inoperable value per Speci-fication 3.2.3 within the next I hour, or 2.

Reduce THERMAL POWER to less than 30% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

e.

With both trip systems inoperable, restore at least one trip system l

to OPERABLE status within I hour, otherwise, eithers 1.

Increase the MINIMUM CRITICAL POWER (MCPR) Limiting Condition for Operation (LCO) to the EOC-RPT inoperable value per Speci-fication 3.2.3 within the next I hour, or 2.

Reduce THERMAL POWER to less than 30% RATED THERMAL POWER within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

LA SALLE - UNIT 2 3/4 3-39 Amendment No. 113

1 I

DESIGN FEATURES i

l 1,, 3 REACTOR CORE TUEL ASSEMBLIES 5.3.1 The reactor shall contain 764 fuel assemblies. Each assembly shall consist of a matrix of Eircalloy fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO ) as fuel material. The 2

bundles may contain water rods or water boxes. Limited substitutions of Eircalloy or IIRLO or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel I

assemblies shall be limited to those fuel designs that have been analyzed with I

applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B C) and/or 4

hafnium metal. The control rod assembly shall have a nominal axial absorber length of 143 inches.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of:

1.

1250 psig on the suction side of the recirculation pumps.

2.

1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3.

1500 peig from the discharge shutoff valve to the jet pumps.

c.

For a temperature of 575*F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation l

system is - 21,000 cubic feet at a nominal T,y, of 533'F.

5.5 DELETED LA SALLE - UNIT 2 5-4 Amendment No.113 l

c____-_________

6.0 ADMINISTRATIVE CONTROLS 6.1 ORGANIZATION A.

Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organiza-tions shall include the positions for activities affecting the safety of i

the nuclear power plant.

1.

Lines of authority, responsibility, and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descrip-tions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements shall be documented in the Quality Assurance Manual.

2.

The individual filling the ANSI N18.1-1971 section 4.2.1 position of Plant Manager (" Plant Manager"), shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

3.

The Chief Nuclear Officer (CNO) shall have corporate responsibility I

for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

J 4.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the I

appropriate onsite manager; however, they chall have sufficient organizational freedom to ensure their independence from operating pressures.

B.

The Shift Manager shall be responsible for directing and commanding the overall operation of the facility on his shift. The primary manage-ment responsibility of the shift Manager shall be for safe operation of the nuclear facility on his shift under all conditions.

C.

The shift manning for the station shall be as shown in Figure 6.1-3.

LA SALLE - UNIT 2 6-1 Amendment No.113

6pf@lSTRATIVE CONTROLS 1,

At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the reactor is in OPERATIONAL CONDITION 1,2 or 3, at least one licensed Senior Reactor Operator who has been designated by the Shift Manager to assume the control room direction responsibility shall be in the Control Room.

2.

A radiation protection technician

• shall be on site when fuelis in the reactor.

3.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

4.

DELETED 5.

The Independent Safety Engineering Group (ISEG) shall function to examine unit operating characteristics, NRC issuances, industry advisories, Licensee Event 3

Reports and other sources of plant design and operating experience information, including plants of similar design, which may indicate areas for improving unit safety.

The ISEG shall be composed of at least three, dedicated, full-time engineers of multi-disciplines located on site and shall be augmented on a part time basis by personnel from other parts of the Commonwealth Edison Company organization to provide expertise not represented in the group. The ISEG shall be responsible for maintaining surveillance of unit activities to provide independent verification # that these activities are performed correctly and that human errors are reduced as much as practical. The ISEG shall make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities or other means of improving unit safety to the Manager of Quality and Safety Assessment and the Plant l

Manager.

6.

The Shift Technical Advisor shall provide advisory technical support to the Shift Manager in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

The radiation protection technician position may be less than the minimum requirement for a period of time not to exceed two hours in order to accommodate unexpected absence provided immediate action is taken to fill the required position.

1. Not responsible for sign-off feature.

4 LA SALLE - UNIT 2 6-2 Amendment No.113

~

• ADMINISTRATIVE CONTROLS 7.

The amount of evertime worked by unit staff members performing safety related functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12).

8.

The Operations Manager or Shift Operations Supervisor shall hold a Senior Reactor Operator License.

D.

Qualifications of the station management and operating staff shall meet minimum acceptable levels as described in ANSI N18.1, " Selection and Training of Nuclear Power Plant Personnel," dated March 8, 1971. The Health Physics Supervisor shall meet the requirements of radiation protec-tion manager of Regulatory Guide 1.8, September 1975. The ANSI N18.1-1971 qualification requirements for Radiation Protection Technician may also be met by either of the following alternatives:

1.

Individuals who have completed the Radiation Protection Technician training program and have accrued 1 year of working experience in the specialty, or 2.

Individuals who have completed the Radiation Protection Technician training program, but have not yet accrued 1 year of working experi-ence in the specialty, who are supervised by on-shift health physics supervision who meet the requirements of ANSI N18.1-1971 Section 4.3.2, " Supervisor Not Requiring AEC Licenses," or Section 4.4.4,

" Radiation Protection."

E.

Retraining and replacement training of Station personnel shall be in accordance with ANSI N18.1, " Selection and Training of Nuclear Power Plant Personnel", dated March 8, 1971 and Appendix "A" of 10 CFR Part 55, and shall include familiarization with relevant industry operational experience.

F.

Retraining shall be conducted at intervals not exceeding 2 years.

l LA SALLE - UNIT 2 6-3 Amendment No.113 L

=.

FIGURE 6.1-3 g

(a)m)

POSITION )

MINIMUM CREW NUMBER S

EACH UNIT IN ONE UNIT IN EACH UNIT IN CONDITION 1,2, OR 3 CONDITION 1,2. OR 3, CONDITION 4 OR S AND ONE UNIT IN OR DEFUELED CONDITION 4 OR 5 OR DEFUELED SM 1

1 1

SRO 1

1 None RO 3

3 2

AO 3

3 3

STA" 1

1 None (a)

This table reflects the total requirements for shift staffing of both units.

With the exception of the Shift Manager, the shift crew composition may be one less than the minimum requirements of Figure 6.1-3 for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Figure 6.1-3. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

(b)

Table Notation:

SM Shift Manager with a Senior Reactor Operator license for each unit whose reactor contains fuel.

SRO Individual with a Senior Reactor Operator license for each unit whose reactor contains fuel.

During CORE ALTERATIONS on either unit a licensed SRO or licensed SRO limited to fuel handling, who has no other concurrent responsibilities, must be present to observe and directly supervise this operation.

\\

RO An Individual with a Reactor Operator license or a Senior Reactor Operator license for unit assigned. At least one RO shall be assigned to each unit whose reactor contains fuel.

Individuals acting as relief operators shall hold a license for both units. Otherwise, for each 1

unit, provide a relief operator who holds a license for the unit assigned.

AO At least one auxiliary operator shall be assigned to each unit whose reactor contains fuel.

~

STA ShiftTechnical Advisor.

(c)

While either unit is in CONDITION 1,2, or 3, an individual with a valid SRO license shall be designated to assume the control room command function. With both Units in CONDITION 4 or 5, an individual with a valid SRO or RO license shall be designated to assume the control room command function.

(d). The STA position shall be filled by an individual who meets the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.

LA SALLE - UNIT 2 6-13 Amendment No.113 i

m___

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' THIS PAGE INTENTIONALLY LEFT BLANK l

I i

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LA SALLE - UNIT 2 6-14 Amendment No.113

, ADMINISTRATIVE CONTROLS 6.1.1 HICH RADIATION AREAS 6.1.1.1 Pursuant to Paragraph 20.203(c)(5) of 10 CFR 20, in lieu of the

" control device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrom/hr* but less than 1000 mrom/hr* shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

Individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas in which the intensity of radiation is greater than 100 mrem /hr* but less than 1000 mrom/hr*,

provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the followings A radiation monitoring device which continuously indicates the a.

radiation dose in the area.

b.

A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been establishrsd and personnel have been made knowledgeable of them.

c.

A health physics qualified individual, i.e.,

qualified in radiation protection procedures, with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Health Physicist in the Radiation Work Per. nit (RWP).

6.1.1.2 In addition to the requirements of 6.1.1.1, above, for areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrom*, the computer shall be programmed to permit entry through locked doors for any individual requiring access to any such High-High Radiation Areas for the time that access is required.

6.1.1.3 Keys to manually open computer controlled High Radiation Area doors and High-High Radiation Area doors shall be r.aintained under the Administra-tion control of the Shift Manager on duty and/or the Health Physicist.

l 6.1.1.4 High-High Radiation areas, as defined in 6.1.1.2 above, not equipped with the computerized card readers shall be maintained in accordance with 10 CFR 20.203 c.2 (iii), locked except during periods when access to the area is required with positive control over each individual entry, or 10 CFR 20.203.c.4.

In the case of a High Radiation Area established for a period of 30 days or less, direct surveillance to prevent unauthorized entry may be substituted.

Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work area and the maximum allowable stay time for individuals in that area.

For I

I i

l l

LA SALLE UNIT 2 6-15 Amendrent No.113 l

_ PLANT OPERATING PROCEDURES AND PROGRAMS (Continu d)

B.

Radiation control procedures shall be maintained, made available to all station personnel, and adhered to.

These procedures shall show permissible radiation exposure and shall be consistent with the requirements of 10 CFR 20.

This radiation protection program shall be organized to meet the requirements of 10 CFR 20.

C.

TECHNICAL REVIEW AND CONTROL Procedures required by Specification 6.2.A and 6.2.5 and other procedures which affect nuclear safety, as determined by the Plant Manager, and changes thereto, other than editorial or typographical changes, shall be reviewed as follows prior to Laplementation except as noted in Specification 6.2.D 1.

Each procedure or procedure change shall be independently reviewed by a qualified individual knowledgeable in the area affected other than the individual who prepared the procedure or procedure change. This review shall include a determination of whether or not additional cross-disciplinary reviews are necessary.

If deemed necessary, the reviews shall be performed by the qualified review personnel of the appropriate discipline (s).

2.

Individuals performing these reviews shall meet the applicable experience requirements of ANSI N18.1-1971, Sections 4.2, 4.3, 4.4, 4.5.1, or 4.6, and be approved by the Plant Manager.

3.

Applicable Administrative Procedures recommended by Regulatory Guide 1.33, Plant Emergency Operating Procedures, and changes thereto shall be submitted to the Onsite Review and Investigative Function for review and approval prior to implementation.

4.

Review of the procedure or procedure change will include a determination of whether or not an unreviewed safety question is involved. This determination will be based on the review of a written safety evaluation prepared by a qualified individual or documentation that a safety evaluation is not required. Onsite Review, Offsite Review and commission approval of items involving unreviewed safety questions shall be obtained prior to station approval for implementation.

5.

The Department Head approval authority shall be specified in station procedures.

6.

Written records of reviews performed in accordance with this specification shall be prepared and maintained in accordance with Specification 6.5.

7.

Editorial and Typographical changes shall be made in accordance with i

station procedures.

LA SALLE - UNIT 2.

6-17 Amendment No.113

• ADMINISTRATIVE CONTROLS D.

Tempor:ry ch ngis to procedur:s 6.2.A cnd 6.2.B cbove may be mido provided:

1.

The intent of the original procedure is not altered.

l 2.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

3.

The change is documented, reviewed and approved in accordance with Specification 6.2.C. within 14 days of implementation.

E.

Drills of the emergency procedures described in Specification 6.2.A.d shall be conducted

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at frequencies as specified in the Generating Stations Emergency Plan (GSEP). These drills will be planned so that during the course of the year, communication links are tested and outside agencies are contacted.

F.

The following programs shall be established, implemented, and maintained:

1.

Primary Coolant Sources Outside Primary Containment A program to reduce leakage from those portions of systems outside primary containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include LPCS, HPCS, RHR/LPCI, RCIC, hydrogen recombiner, process sampling, containment monitoring, and standby gas treatment systems. The program shallinclude the following:

l a.

Preventive maintenance and periodic visual inspection requirements, and b.

Integrated leak test requirements for each system at refueling cycle intervals or less.

2.

In-Plant Radiation Monitorina A program which will ensure the capability to accurately determine the airbome iodine concentration in vital areas under accident conditions. This program shall include the following:

a.

Training of personnel, b.

Procedures for monitoring, and c.

Provisions for maintenance of sampling and analysis equipment.

3.

P_o, st-accident Samplina A program which will ensure the capability to obtain and analyze reactor coolant, radioactive lodines and particulate in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shallinclude the following:

a. Training of personnel, b.

Procedures for sampling and analysis, c.

Provisions for maintenance of sampling and analysis equipment.

LA SALLE - UNIT 2 6-18 Amendment No.113 l'

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e ADMINISTRATIVE CONTROLS 6.4 ACTION TO BE TAKEN IN THE EVENT A SAFETY LIMIT IS EXCEEDED If a safety limit is exceeded, the reactor shall be shut down immediately pursuant to Specification 2.1.1, 2.1.2 and 2.1.3, and critical reactor operation shall not be resumed until authorized by the NRC.

The conditions of shutdown shall be promptly reported to the Sito Vice President or his designated alternate. The incident shall be reviewed by the Onsite and Offsite h

Review and Investigative Functions and a separate Licensee Event Report for each occurrence shall be prepared in accordance with section 50.73 to 10 CFR Part 50.

The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Site Vice President and the Director of Safety Review shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

6.5 PLANT OPERATING RECORDS A.

Records and/or logs relative to the following items shall be kept in a manner convenient for review and shall be retained for at least 5 years:

1.

Records of normal plant operation, including power levels and periods of operation at each power level; 2.

Records of principal maintenance and activities, including inspection and repair, regarding principal items of equipment pertaining to nuclear safety; 3.

Records and reports of reportable events; 4.

Records and periodic checks, inspection and/or calibrations performed to verify that the surveillance requirements (see Section 4 of these specifications) are being met.

All equipment failing to meet surveillance requirements and the corrective action taken shall be recorded; 5.

Records of changes to operating procedures; 6.

shift Manager logs; and 7.

Byproduct material inventory records and source leak test results.

B.

Records and/or logs relative to the following items shall be recorded in a manner convenient for review and shall be retained for the life of the plant:

1.

Substitution or replacement of principal items of equipment pertaining to nuclear safety; 2.

Changes made to the plant as it is described in the SAR; 3.

Records of new and spent fuel inventory and assembly histories; 4.

Updated, corrected, and as-built drawings of the plant; 5.

Records of plant radiation and contamination surveys; 6.

Records of offsite environmental monitoring surveys; LA SALLE - UNIT 2 6-21 Amendment No.113

• ADMINISTRATIVE CONTROLS PLANT OPERATING RECORDS (Continu3d) 7.

Records of radiation exposure for all plant personnel, including all contractors and visitors to the plant, in accordance with 10 CFR Part 20; 8.

Records of radioactivity in liquid and gaseous wastes released to the environment; 9.

Records of transient or operational cycling for those components that have been designed to operate safety for a limited number of transient or operational cycles (identified in Table 5.7.1-1);

10. Records of individual staff members indicating qualifications, experience, training, and retraining;

11. Inservice inspections of the reactor coolant system;

12. Minutes of meetings and results of reviews and audits performed by the offsite and onsite review and audit functions;

33. Records of reactor costs and experiments;

14. Records of Quality Assurance activities required by the QA Manual, except for those items specified in Section 6.5.A;

15. Records of reviews performed for changes made to procedures on equip-ment or reviews of tests and experiments pursuant to 10 CFR 50.59;

16. Records of the service lives of all hydraulic and mechanical snubbers required by specification 3.7.9 including the date at which the ser-vice life commences and associated installation and maintenance records;

17. Records of analyses required by the radiological environmental monitoring program;

18. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM; and

19. Records of pre-stressed concrete containment tendon surveillance.

6.6 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted LA SALLE UNIT 2 6-22 Amendment No.

113

e ADMINISTRATIVE CONTROLS 6.6 REPORTING MEOUIREMENTS (Continu d) to the director of the appropriate Regional Office of Inspection and Enforce-ment unless otherwise noted.

A.

Routine Reports 1.

Startup Report A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amend-ment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufac-tured by a different fuel sutglier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perform-ance of the plant. The report shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these vulues with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license con-ditions based on other commitments shall be included in this report.

l startup reports shall be submitted within (1) 90 days following com-pletion of the startup test program, (2) 90 days following resumption

{

cr commencement of commercial power operation, or (3) 9 months follow-l ing initial criticality, whichever is earliest. If the startup report j

does not cover all three events (i.e., initial criticality, completion of startup test presram, and resumption or commencement of commerci.1 l

power operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

2.

Annual Report I

1 A tabulation shall be submitted on an annual basis prior to March 1 of i

each year of the number of station, utility, and other personnel (including contractors) receiving exposures greater than 100 mrom/yr and their associated man rem exposure according to work and job i

functions (Notes this tabulation supplements the requirements of l

section 20.407 of 10 CFR 20),

e.g.,

reactor operations and surveil-i lance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose j

assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totaling

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less than 20% of the individual total dose n?ad not be accounted for.

In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions.

i The results of specific attivity analysis in which the primary coolant exceeded the limits of specification 3.4.5 shall be included in the Annual Report along with the following information (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) ResultG of the last isotopic analysis for radioiodine performed prior to exceedLng the limit, results of analysis while limit was exceeded and results of one analysis after j

the radioiodine activity was reduced to less than limit. Each result i

should include date and time of sampling and the radioiodine l

concentrations; (3) clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> LA SALLE UNIT 2 6-23 Amendment No.113

ADMINISTRATIVE CONTROLS prior to the first sample in which tha limit was exceeded; (4) Graph of the I-131 concentration and one other radioiodine isotope concen-tration in microcuries per gram as a function of time for the duration of the specific activity above the steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radiciodine limit.

3.

Annual Radiological Environmental Operatina Report

• The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries interpretations, and analysis of trends of the results of tneRadiciogicalEnvironmentalMonitoringProgramforthereporting period. The material provided shall be consistent with the objectives outlined in (1) the 00CM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix ! to 10 CFR Part 50.

4.

Annual Radioactive Effluent Release Report **

The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1 consistent with the objectives outlined in the ODCM and PCP and (2) )n.

i conformance with 10 CFR 50.36a and Section IV.B.1 of Append u I to 10 CFR Part 50.

5.

Monthlv Doeratina Reoort Routine reports of operating statistics and shutdown experience, including documentation of all challenges-to safety / relief valves, i

shall be submitted on a monthly basis to the addressees specified in 10 CFR 50.4 no later than the 15th of each month following the calendar month covered by the report.

A report of any major changes to the radioactive waste treatment I

systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function.

6.

Core Ooeratina Limits Reoort a.

Core o>erating limits shall be established and documented in the CORE 0)ERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

A sin le submittal may be made for a multi-unit station.

A sin le submittal may be made for a multi-unit station. The submittal shoul combine those sections that are common to all units at the station; i

however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

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i LA SALLE UNIT 2 6-24 Amendment No.113

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l gpMINISTRATIVE CONTROLS l

l Core Doeratino Limits ReDort (Continued) c.

The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as l

shutdown margin, and transient and accident analysis limits) of l

the safety analysis are met.

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d.

The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory l

commission Document Control Desk with copies to the Regional Administrator and Resident Inspector.

B.

Deleted C.

Unique Reporting Requirements 1.

Special Reports shall b*e submitted to the Regional Administrator of the NRC Regional Office within the time period specified for each report.

6.7 PROCESS CONTROL PROGRAM (PCP)*

6.7.1 The PCP shall be approved by the Commission prior to implementation.

6.7.2 Licensee initiated changes to the PCP:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.18.

This documentation shall contain:

1)

Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s),

and 2)

A determination that the change will maintain the overall con-formance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

b.

Shall become effective upon review and acceptance by the Onsite Review and Investigative Function.

• The Process control Program (PCP) is common to La Salle Unit 1 and La Salle Unit 2.

i LA SALLE UNIT 2 6-26 Amendment No.113 l

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ADMINISTRATIVE CONTROLS f

6.8 OFFSITE DOSE CALCULATION MANUAL (ODCMi*

6.8.1 The ODCM shall be approved by the Commission prior to implementation.

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6.8.2 Licensee initiated changes to the ODCM:

a.

Shall be documented and records of reviews performed shall be retained as required by Specification 6.5.B.18.

This documentation shall contain:

1) Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s),

and

2) A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.106, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations.

b.

Shall become effective after review and acceptarce by the On-31te Review and Investigative Function and the approval of the Plant Manager on the date specified by the On-Site Review and Investigative Function.

L c.

Shall be submitted to the Commission in the form of a complete,,

i legible copy of the entire ODCM as a part of or concurrent with the Annual Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made effective. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

i 6.9 MAJOR CMANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS l

6.9.1 Licensee initiated major changes to the radioactive waste treatment systems (liquid, gaseous and solid):

1 l

a.

Shall be reported to the Commission' in the Monthly Operating Report I

for the period in which the evaluation was reviewed by the Onsier Review and Investigative Function. The discussion of each change shall contain:

1.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59;

• The OFFSITE DOSE CALCULATION MANUAL (ODCM) is common to La Salle Unit I and La Salle Unit 2.

1 LA SALLE UNIT 2 6-27 Amendment No. 113 l

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ADMINISTRATIVE CONTROLS MAJOR. CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Continued) 2.

Sufficient detailed information to totally support the reason for the change without benefit or additional or supplemental information; 3.

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; 4.

An evaluation of the change which shows the predicted releases t

of radioactive materials in liquid and gaseous effluents and/or

)

quantity of solid waste that differ from those previously predicted in the license application and amendments thereto; 5.

An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto; 6.

A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period to when the changes are to be made; 7.

An estimate of the exposure to plant ~ operating personnel as a result of the change; and 8.

Documentation of the fact that the change was reviewed and found acceptable by the Onsite Review and Investigative Function.

b.

Shall becomo effective upon review and acceptance by the Onsite Review and Investigative Function.

l LA SALLE UNIT 2 6-28 Amendment No. 113