ML20141L037

Proposed Tech Specs,Incorporating Restructured Station Organization,Change to Submittal Frequency of Radiological Effluent Release Rept & Other Administrative Changes

ML20141L037
Person / Time
Site:	LaSalle
Issue date:	05/27/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20141L032	List: ML20141L031 ML20141L037
References
NUDOCS 9706020258
Download: ML20141L037 (70)

Text

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ATTACHMENTB PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS I

'I N PF-11 NPF-18 11 1-4 1-4 I

, 6-1* 5-1*  ;

( 5-2* 5-2*

5-4 6-4 6-1 Insert A 6-1 Insert A l 6-2 Inserts B and C 6-2 Inserts B and C l l 6-2a DELETE entire page 6-2a DELETE entire page $

l 6-3* 6-3  ;

l 6-4* 6-4* u 6-13 replace with Insert D 6-13 replace with Insert D 6-14 6-14 l 6-15 6-15 l 6-16* 6-16*

6-17 6-17 6-18 6-18 l 6-19* 6-19' .

6-20* 6-20* $

l 6-20a* 6-20a*

l 6-21 6-21 l 6-22* 6-22 l 6-23' 6-23 6-24 Insert E 6-24 Insert E 6-25 6-25*

l 6-25a*

I 6-26 6-26 6-27 6-27 6-28* 6-28

• No change to this page, it is included for information only to provide clarity for other proposed changes, i

l B-1 9706020258 970527 PDR ADOCK 05000373 P PDR

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j ATTACHMENT B

PROPOSED AMENDMENTS TO THE j LICENSEITECHNICAL SPECIFICATIONS

)e CHANGE

SUMMARY

j Technical Description j Specification item Number Chance Description j

i Unit 1 License 1 The Shift Technical Advisor (STA) function l Condition shall be fulfilled by a fully-trained on-shift j 2.C.(30)(a). technical advisor to Shift Manager.

1 Definition 1.27 2 Change " Semi-Annual Radioactive Effluent 3 Release Reports" to " Annual Radioactive l Effluent Release Reports".

j 5.5 2 Delete line item 5.5.1, which is the text

describing the location of the j Meteorological Tower.

l 6.1. A.2 1 Replace "The Station Manager" with "The j individual filling the ANSI N18.1-1971

Section 4.2.1 position of Plant Manager j (" Plant Manager"),".

$ 6.1.B 3 Delete the requirement to annually issue a

} management directive.

l 6.1.C 1 Delete the discussion of the ANSI

qualifications of the Systems Engineering i Supervisor and Operations Manager.

! 6.1.C.1 1 Change " Shift Supervisor" to " Shift Manager".

4 6.1.C.4 1 Replace "the Shift Supervisor, the Station Control Room Engineer" with "one Shift

! Manager, one Unit Supervisor, the Shift j Technical Advisor,".

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ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS Technical Description Specification item Number Chance Description l

6.1.C.5 1 Change " Site Quality Verification Director ,

and Station Manager" with " Safety I j Assessment / Site Quality Verification l Manager and Plant Manager",

l l 6.1.C.6 1 Revise the description of the Shift l Technical Advisor; Delete the description of specific STA work practices.

6.1.C.6 1 Change " Shift Supervisor" to " Shift Manager".

Figure 6.1-3 1 Supersede with proposed Figure.

6.1.1.3 1 Change " Shift Supervisor" to " Shift Manager".

l 6.2.C 1 Replace " Station Manager" with " Plant Manager".

6.2.C.2 1 Replace " Station Manager" with " Plant Manager".

6.2.C.2 3 Replace " Sections 4.2 and 4.4" with

" Sections 4.2, 4.3, 4.4, 4.5.1, o r 4.6,".

6.2.E 3 Change " Specification 6.2.A.4" to

" Specification"6.2.A.d".

6.5. A.6 1 " Shift Manager" supersedes " Shift engineers' ".

! B-3

l ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSEITECHNICAL SPECIFICATIONS Technical Description Specification item Number Chanae Description 6.6.A.4 2 Change " Semiannual" in both the paragraph header and the first sentence to

" Annual"; Change "6 months" to " calendar year"; Replace "within 60 days after January 1 and July 1" to " prior to May 1".

6.6.A.5 3 Replace the text " Director, Office of Nuclear Reactor Regulation...the appropriate Regional Office, to arrive" to

" addressees specified in 10CFR50.4, to arrive".

6.6.A.5 2 Delete tha text "Any changes to the l OFFSITE DOSE CALCULATION MANUAL l shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition,". Change "a report of any major changes...." to "A report of any major c h a n g e s...".

6.6.C.1 3 Replace " Director of the Office of Inspection and Enforcement (Region lil)"

with " Regional Administrator of the NRC Regional Office".

6.8.2.c 2 Change " Semiannual" in Semiannual Effluent Release Report to " Annual".

Unit 2 pages 3 Change header to " Administrative 6-2, 3, 15, 22, 23, Controls".

24, 27, & 28 B-4

ATTACHMENT B

( PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS i

i insert A The individual filling the ANSI N18.1-1971 Section 4.2.1 position of Plant Manager (" Plant Manager"),

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ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS Insert B Safety Assessment / Site Quality Verification Manager and the Plant Manager.

Insert C i l

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.

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l ATTACHMENT B  !

PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS i l

l Insert D FIGURE 6,1-3 MINIMUM SHIFT CREW COMPOSITIONH4 ,

l POSITIONS MINIMUM CREW NUMBER I

EACH UNIT IN ONE UNIT IN CONDITION 1, EACH UNIT IN CONDITION 4 CONDITION 1,2, OR 3 2, OR 3, AND ONE UNIT IN OR 5 OR DEFUELED CONDITION 4 OR 5 OR DEFUELED )

l SM 1 1 1 SRO 1 1 None RO 3 3 2 AO 3 3 3 STAM 1 1 None 1

(a) This table reflects the total requirements for shift staffing of both units. l With the exception of the Shift Manager, the shift crew composition may be one less than the m6nimum requirements of Figure 8.13 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 onduty 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 pusition 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 heense for each un61 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 heense 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 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.

(d) The STA position may be filled by any individual who meets the Commission Policy Statement on Engineering Expertise on Shlft.

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l ATTACHMENT B PROPOSED AMENDMENTS TO THE

, LICENSE / TECHNICAL SPECIFICATIONS Insert E l

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. I i

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DEFINITIONS 1.20 DELETED U MITING CONTROL ROD PATTERN

, 1.21 A LIMITING CONTROL R0D PATTERN shall be a pattern which results in the i core being on a thensal hydraulic limit, i.e., operating on a limiting .

value for APLHGR, LHGR, or MCPR.

l LINEAR HEAT GENERATION RATE 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 l transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TESI <

l.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 l performed by any series of sequential overla such that the entire logic system is fested. pping or total system steps l MAXIMUM FRACTION OF LIMITING POWER DENSITY l

1.24 The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the highest value of the FLPD which exists in the core.

l MEMBERS (S) 0F THE PUBLIC l

1.25 associateMEMBER (S)d with the plant.0F THE PUBLIC This category shall does not include include all persons employees who are not o of the licensee, its contractors, or vendors. Also excluded from this categor persons who enter the site to service equipment or to make deliveries. yThis are 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 and Theparameters OFFSITE used DOSE in CALCULATION the calculation of oMANUAL (ODCM) ffsite doses resulting f radioactive gaseous and liquid effluents in the calculation of gaseous and liquid effluent monitorina Alare/Tri Setpoints, and in the conduct of the Environmental Radiological Monito ing Program. The ODCM M 1 1 the Radioactive Effluent Controls and Radiolooical l also Environmentalcontain (Mo)nitoring Programs required by Technical Specification Section included in 6.2.F.4 the Annual and (2)Radioldescri tions ofOperating ical Environmental the information and that sh Annual Radioactive Effluent Re ease Reports required by Technica Specification Sections 6.6.A.3 and 6.6.A.4.

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LA SALLE UNIT I' l-4 Amendment No.110

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5.0 DESIGN FEATURES FO ONLY- NO CH NGEb i 5.1 SITE

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EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1. i i LOW POPULATION ZONE j 5.1.2 The low population zone shall be as shown in Figure 5.'1.2-1.

j SITE B0UNDARY FOR GASE0US EFFLUENTS j 5.1. 3 The site boundary for gaser m effluents shall be as shown in Figure 5.1.1-1. '

j SITE BOUNDARY FOR LIQUID EFFLUENTS i

1 5.1. 4 The site boundary for liquid effluents shall be as shown in Figure 5.1.1-1.

5.2 CONTAINMENT CONFIGURATION -

4 l 5.2.1 The primary containment is a steel lined post-tensioned concrete i structure consisting of a drywell and suppression chamber. The drywell is a i , steel-lined post-stressed concrete vessel in the shape of a truncated cone closed i by a steel done. The drywell is above a cy;indrical steel-lined ast-stressed t ,.

concrete suppression chamber and is attached to the suppression cumber through a series of downcomer vents. The drywell has a minimum free air volume of

! 229,538 cubic feet. The suppression chamber has an air region of 164,800 to I i, 168,100 cubic feet and a water region of 128,800 to 131,90) cubic feet.

l DESIGN TEMPERATURE AND PRESSURE l 5.2.2 The primary containment is designed and shall be maintained for:

a. Maximum internal pressure 45 psig.

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b. Maximum internal temperature: drywell 340'F.

suppression chamber 275'F.

{ c. Maximum external pressure 5 psig.

s j d. Maximum floor differential pressure: 25 psid, downward.

, 5 psid, upward. -

j SECONDARY CONTAlletENT i

i j 5.2.3 The secondary containment consists of the Reactor Building, the equipment i

! access structure and a portion of the main steam tunnel and has a minimum free volume of 2,875,000 cubic feet.

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i l LA SALLE - UNIT 1 5-1 Amendment No.18  :

1 4 I _

QONLy- 56 CHANGE [

Illinois River ~

EXCLUSION AREA AND i SITE BOUNDARY FOR

=

GASEOUS AND LIQUID EFFLUENTS Effluent Discharge Point 7

20, O 00 1 Mlle Scale in Feet N Site Boundary)

(Property Line w

Waste Stabilization Pond l

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/

r, l Exclusion l

' LaSalls Lake

{ s )

i Area "'

A

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IStation Vent !' , -

I Stack i I

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Figure 5.1.1-1 LA SALLE - UNIT 1 5-2 Amendment No. 85

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DESIGN FEATURES

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} 5.3 REACTOR CORE j _ FUEL ASSEMBLIES i 5.3.1 The reactor core shall contain 764 fuel assemblies. .Each assembly l i consists o'f a matrix of Zircalloy clad fuel rods with an initial composition  !

of slightly enriched uranium dioxide. UD ,

to those fuel designs approved for use i$.BWR's. Fuel assemblies shall be limited 4

_ CONTROL R0D ASSEMBLIES 5.3.2 j The reactor core shall contair 185 cruciform shaped control rod assemblies.

j Thecontrol The control rod assemblymaterial shall shall benominal have a, boron carbide axial power aC) and/or hafnium r er length of 143 (8,bso metal. b j inches. ,

f 5.4 REACTOR COOLANT SYSTEM DESIGN PRES 5URE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a.

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

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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 psig from the discharge shutoff valve to the jet pumps.

c. For a temperature of 575'F.

VOLUME 5.4.2 systemThe total water is

• 21,000 cubicandfeetsteam volume of the reactor vessel and recireviation at a nominal 7,,, of $33*F.

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5. 5 ".d::;;r--_^ ^ - -~ T: l^GE6MELETEj ~

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he meteorolonil cal tower shall be locafed asyown on Figure 5.1.1-LA SALLE - UNIT 1 5-4 Amendment No. 70

600 ADMINISTPATIVE CONTROLS

, 6.1 ORGANIZATION

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

1. Lines of authority, responsibility, and comunication 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-2 i

tions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms i -

of documentation. These requirements shall be documented in the ggg uality Assurance Manual.

2. trh station Managae shall be responsible for overall unit safe j

operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

4

' 3. The Chief Nuclear Officer (CNO) shall have corporate responsibility l 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 i nuclear safety.

4. The individual who train the operating staff and those who carry j 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.

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i B. The Shift G-"a3 shall be responsibi for directing and comanding the overall operation of the facility his, shift. The primary manage-i i

ment responsibility of the Shift ""ri*vicas shall be for safe operation of the nuclear facility on his shift under all conditi_onte A managemen r61rective signeo oy Ine site Vice President emphasizing this primary i

management responsibility and that clearly establishes the command duties

, Iof the Shift Supervisor shall be reissued to all station oersonnel on an J bnnual badd ..

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

r inoividuaT filling tne position or nuems engineering aupervisor shall l meet the minimum acceptable level for " Technical Manager" as described in Section (.2.4 of ANSI N18.1-1971. The individuals filling the position of Operations Manager shall meet the minimum acceptable level for " Plant

Manager" as described in Section 4.2.1 of ANSI N18.1-1971. -

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LA SALLE - UNIT 1 6-1 Amendwit No.107

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ADMINISTRATIVE CONTROLS I. At least one licensed Reactor Operator shall be in the control room  !

i when fuel is in the reactor. In addition, while the reactor is in  !

OPERATIONAL CONDITION I, 2 or 3, at least one licensed Senior ..

l Reactor Operator who has been designated by the ShiftEunnvhnato I i assume the control room direction responsibility shall be in the%

Control Room.

2. A radiation protection technician

• shall be on site when fuel is in l the reactor. l 1

a

3. All CORE ALTERATIONS shall be observed and directly supervised by i

either a licensed Senior Reactor Operator or Senior Reactor Operator i

' limited to Fuel Handling who has no other concurrent responsibilities 3 during this operation.

b 4. A site Fire Brigade of at least 5 members shall be maintained onsite i

g g1g) g ,

at_ all times *. The Eire Brigade shall not include dhe sh1 M

_ rsonnvisor. tne station Control Room Engineen and the 2 other members l 4 UOX9evise,' of the minimum shift crew necessary for safe shutdown of the unit and 4

g 3hgyy i

any personnel required for other essential functions during a fire

. emergency. l i Ticckgicd i j hdyise

5. The Independent Safety Engineering Group (ISEG) shall function to l examine unit operating characteristics, NRC issuances, industry l advisories, Licensee Event Reports and other sources of plant design i

i and operating experience information, including plants of similar design, which may indicate areas for improving unit safety. The ISEG l l

{ shall be composed of at least three, dedicated, full-time engineers 3 ,

of multi-disciplines located on site and shall be augmented on a 1

( part-time basis by personnel from other parts of the Commonwealth a

Edison Company organization to provide expertise not represented in the group. The ISEG shall be responsible for maintaining l b, i

j surveillance of unit activities to provide independent verificationi 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, l i

J maintenance activities, operations ac".ivities or other means of _

improvintunit safety to_1hs51;e Qua'ity Verificauon Director and) j ggg% pe Station Manager.

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i 6. ffiieStationControlRoomEngineer(SCRE)mayserveastheShi Technical Advisor (STA) during abnorma) operating and accident con-i dNSEPtT "Ch # ditions. During these conditions, the SCRE or other on duty STA shall provide technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering and plant analysis with j j stegard to the safe operation of the unit. >

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• The radiation protection technician and Fire Brigade composition may be less -

than the minimum requirements for a period of time not to exceed two hours in i order to accommodate unexpected absence provided imediate action is taken to j fill the required positions.

l' , #Not responsible for sign-off feature.

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LA SALLE - UNIT I 6-2 Amendment No. 107

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

+

ADMINISTRATIVE CONTROL $

ssure capability for performance of all STA functions:

a. The shift foreman (SRO) shall participate in the SCRE shift relief

- turnover. -

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b. During the shift, the shift engineer and the shift foreman ($R0) i shall be made aware of any significant changes in plant status in a timely manner by the SCRE. ,

c. During the shift, the shift engineer and the shift foreman (5R0) 4 shall remain abreast of the current plant status. The shift fore-man (SRD) shall return to the control room two or three times per shift, where practicable, to confer with the SCRE regarding plant status. Where not practical to return to the control room, the .

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shift foreman (SR0) shall periodically check with the SCRE for a l

plant status updata. The shift foreman (SRD) shall not abandon j . duties critical to reactor operation, unless specifically ordered j by the shift engineer.  ;

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] LA SALLE - UNIT 1 6-2a Amendment No. 78 I

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i ADMINISTRATIVE CONTROLS NFO ONLY : No c/MNM

7. The amount of overtime worked by unit staff members performing safety i

related functions shall be limited and controlled in accordance with the NRC Policy Statement on working hours (Generic Letter 82-12).

. t~

l 8. The Operations Manager or Shift Operations Supervisor shall hold a l

Senior Reactor Operator License.

. D. Qualifications of the station management and operating staff shall meet 3 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 l 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 j training program and have accrued 1 year of working experience in the

specialty, or .

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2. Individuals who have completed the Radiation Protection Technician l

training program, but have not yet accrued 1 year of working experi-i ence in the specialty, who are supervised by on-shift health physics j supervision who meet the requirements of ANSI N18.1-1971 Section i

4.3.2, " Supervisor Not Requiring AEC Licenses," or Section 4.4.4,

" Radiation Protection."

i E. Retraining and replacement training of Station personnel shall be in 1

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.

j F. Retraining shall be conducted at intervals not exceeding 2 years.

J LA SALLE - UNIT 1 6-3 Amendment No. 107 i-

i l g INISTRATIVE CONTROLS i

l G. DELETED (The Review and Investigative Function and the Audit Function are ,

i. described in the Quality Assurance Manual Topical Report CE-1-A). ,

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j INFO oNLY , l l

l No CH AtJ C4ES i '

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LA SALLE - UNIT 1 6-4 Amendment No. 107 INext once is 6-13)

(IN.5KRT "D*D

'.,, Figure 6.1-3 j

MINI.MUN SNIFT CREW CONPOSITION .

i 1 .

J WITN UNIT 2 IN CONDITION 1, 2, OR 3 i

. POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION 4

CONDITION 1, 2 and 3 CONDITION 4 and 5 ,

1 8

i SE l' 1 f l SF la None

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l l M N 1 I l M h 1

! SCRE l' ' one

\

! or, whenever a SCRE (SR0/STA) is not . included in the shift crew I i composition, the minimum shift crew composition shall be as

! follows:

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j .. WITN UNIT 2 IN CONDITION 1, 2. OR 3

'( POSITION NUMBEitOFINDIVIDUALSREQUIREDTOFILLPOSITION -

] CONDITION 1, 2 and 3 CONDITION 4 and 5

{

j SE a ga l

a

}

SF l y,,,

M 5 1 W 5, 1 i STA a y,,,

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l j WITN UNIT [IN CONDITION 4 or'5 OR DEFUELS POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION

. +

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CONDITION 1, 2 and 3 . CONDITION 4 and 5 i 8 a SE 1 l i i -

i SF 1 None i m 2 I

5 b M . 2 .

{\ STA 1 None

. LA SALLE ~ ,f13 Amendment No. 66 4

a

f

{ . -

. Figure 6.1-3(Continued)

MINIMUM SHIFT CREW COMPOSITION 5

NOTES i

a/ . Individual may fill the same position on Unit 2.

' 1 5/ One of the two required individuals may fill the same position on Unit 2.

- Shif t Supervisor (Shif t Engineer) with a Senior Reactor Operators l /) 3E License on Unit 1. \

f SF - Shift Foreman with a Senior Reactor Operators License on Unit 1.

R0 - Individual with a Reactor Operate s License on Unit 1. 4

- Auxiliary Operator.

j f) A0 SCRE - Station control Roos Engineer with a Senior Reactor Operators License.

j

! { Except for the Shift Supervisor, the Shift Crew composition may be one less

! than the minimum requirements of Figure 6.1-3 for a period of time not to k i

exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on duty shift i crew members provided immediate action is taken to restore the shift crew t i

' 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 crasan being late or absent.

I While the unit is in OPERATIONAL CONDITION 1, 2, or 3, an individual with a l j valid SRO license shall be designated to assume the Control Roos direction function. While the unit is in.0PERATIONAL CONDITION.4 or 5, an individual with a valid SRO or RD license shall be designated to assume the Control Roos  ;

e direction function.  !

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'TH.Ts fA G E INTEN 7TO A44LL Y i

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LA SALLE UNIT 1 6-14 Amendment No. 66 -

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ADMINISTRATIVE CONTROLS

' .6.1.1 HIGH RADIATION AREAS l

5. '1.1.1 Pursuant to Paragraph 20.203(c)(5) of 10 CFR 20, in lieu of the j " control device" or "alare signal" required by paragraph 20.203(c)(2) of i

10 CFR 20, each high radiation area in which the intensity of radiation is i greater than 100 area /hr" but less than 1000 area /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 j by such individuals, may be exempt from the RWP issuance requirement during the j

performance of their assigned duties in high radiation areas in which the intensity of radiation is greater than 100 area /hr* but less than 1000 area /hr*,

provided they are othemise 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 j acre of the following: ,

a. A ' radiation monitoring device which continuously indicates the j 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 i dose is received. Entry into such areas with this monitoring device l may be made after the dose rate level in the area has been established j and personnel have been made knowledgeable of them. .

c. A health physics qualified individual, i.e., qualified in radiation j

'7

' protection procedures, with a radiation dose rate monitoring device, -

1 who is responsible for providing positive control over the activities l within the area and shall perform periodic radiation surveillance at i the frequency specified by the Health Faysicist in the Radiation j .

Work Permit (RWP).

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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 l receive in one hour a dose greater than 1000 ares *, the computer shall be programmed to permit entry through locked doors for any individual requiring 1

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 shg11 be maintained under the Administra-j tion control of the Shift diERingon duty and/or the Health Physicist. K Der

! 6.1.1. 4 High-High Radiation areas, Its defined in 6.1.1.2 above, not equipped j with the computerized card readers shall be saintained in accordance with 10 CFR 20.203 c.2 (iii), locked except during periods when access to the area 4

I 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 7essin locked except during periods of access by personnel under an 1 approved RWP which shall specify the dose rate levels in the immediate work area j and the maximum allowable stay time for individuals in that area. For individual i

areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 ares

• that are located within i

[ i- " Measurement made at 18" free source of radioactivity.

j , LA SALLE UNIT 1 6-15 Amendment No. 66 i -

l .

i rNF0 ONLY - NO CH4nksg.5

^ -

i ADMINISTRATIVE CONTROLS .

(

HIGHRADIATIONAREAS(Continued) j individual areas accessible to personnel with radiation levels such that a

! major portion of the body could receive in one hour a dose in excess of 1000 4 ares" that are located within large areas, such as the containment, where no j enclosure exists for purposes of locking, and no enclosure can be reasonably j constructed around the individual areas, then that area shall be roped off, j conspicuously posted and a flashing light shall be activated as a warnin0 i device. In lieu of the stay time spec' fication of the RWP, direct or remote,

such as use of closed circuit TV cameras, continuous surveillance may be made i by personnel qualified in radiation protection procedures to provide positive l l

j exposure control over the activities within the area.

6.2 PLANT OPERATING PROCEDURES AND PROGRAMS

A. Written procedures shall be established, implemented, and maintained i i covering the activities referenced below

I

{ a. The applicable procedures recommended in Appendix A, of Regulatory j Guide 1.33, Revision 2, February 1978, t

i b. The emergency operating procedures required to implement the

/ requirements of NUREG-0737 and Supplement I to NUREG-0737 as stated j in Section 7.1 of Generic Letter No. C2-33, I c. Station Security Plan implementation,  ;

- i

! l j d. Generating Station Emergency Response Plan implementation,  !

e. PROCESS CONTROL PROGRAM implementation, l I

f. OFFSITE DOSE CALCULATION MANUAL implementation, and  !

g. Fire Protection Program imp 1,ementation. j

\

l l

4

" Measurement made at 18" from source of , radioactivity. l

\ .. . . l LA SALLE UNIT 1 6-16 Amendment No. M, 86

ADMINISTRATIVE CONTROLS PLANT OPERATING PROCEDURES AND PROGRAMS (Continued)

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 CFR1!0.~ -

C. TECHNICAL REVIEW AND CONTROL ProceduresrequiredbySpecification6.2.Aand6.2. Band]otherprocedures which affect nuclear safety, as determined by the annom nanager, and changes thereto, other than editorial or typographical changes, shall be reviewed as follows prior to implementation 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 appropriatediscipline(s).

2. Individuals performing these reviews shall meet the apalicable experience requirements of ANSI N18.1-1971, Sections . p t '. , 1 and be approved by the ager 4.3, y ,4 9,s.), ,,. ,g

3. Applicable Administrative Procedures recommended by Regul'atory Guide 1.33 Plant Emergency Operating Procedures, and changes thereto shall be submitted to the Onsite Review and Investigativp' 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,.0ffsite 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 l Specification 6.5. j

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

I LA SALLE - UNIT 1 6-17 Amendment No.107

i i

ADMINISTRATIVE CONTROLS

} D. Temporary changes to procedures 6.2. A and 6.2.B above may be made

} provided:

I

1. W intent of the original procedure is not altered.

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

4

3. h change is documented, reviend and approved'in accordance _with l Specification 6.2.C. within 14 asys of implementation.

i p, A,p E. Drills of the emergency procedures described in Specificationd.C. A.0 LJ shall be conducted at frequencies as specified in the Generating 5tations ,

! Emergency Plan (GSEP). These drills will be planned so that during the

! course of the year, communication links are tested and outside agencies j are contacted. .

! F. The following programs shall be established, implemented, and maintained: I

! 1. Primary Coolant Sources Outside Primary Containment

• A program to reduce leakage free those portions of systems outside I primary containment that could contain highly radioactive fluids i during a serious transient or accident to as low as practical levels.  !

The systems include LPCS*, HPCS, RNR/LPCI, RCIC, Iqydrogen recombiner, process sampling, containment monitoring, and staney gas treatment systems. h program shall include the following

a. Preventive maintenance and periodic visual inspection require-ments, and -

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

2. In-Plant Radiation Monitorino A program which will ensure the capability to accurately detamine the airborne 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. Post-accident Samolina A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous affluents, and containaient atmosphere samplas under accident conditions, h program shall include the following:

a. Training of personnel,

b. Pmeedures for sampling and analysis,

c. Provisions for maintenance of sampling and analysis equipment.

LA SALLE UNIT 1 6-18 Anordnent No. N, N, 86

i ALHINIS1RAllVE CONTROLS O ONL?(-Al C//AN'G"S7 PLANT OPERATING PROCEDURES AND PROGRAMS (Continued)

4. Radioactive Effluent Controls Proaraa

{

i A program shall be provided conforming with 10 CFR 50.36a for the i cor, trol of radioactive effluents and for maintaining the doses to l HEhBERS GF THE PUBLIC from radioactive effluents as low as reasonably acriezable. The program 1 i

shall be implemented by o(>er)ating procedures, andshall shall be contained in include

! remedial actions to be ta(en whenever the program im '(3)its are i

exceeded. The program shall include the following elements:

~

}

a.

Limitations on the operability of radioactive liquid and gaseous j monitoring instrumentation including surveillancs tests and set-4 point determination in accordance with the methodology in the 00CM, b.

i .

Limitations on the concentrations of radioactive material

released in liquid effluents to UNRESTRICTED AREAS conforming to

10 times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402, 1

c.

i j

MonitorinE.fluents gaseous e in accordance with 10 CFR 201302, samplin and with the methodology and parameters in the ODCM, l

d. Limitations on the annual and quarterly doses or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conform-ing to Appendix I to 10 CFR Part 50,

e. Determination of cumulative and )rojected dose contributions from radioactive effluents for tie current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days, f.

Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would  !

~ xceed e 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR Part 50,

g. Limitations on the dose rate resulting from radioactive materials i released in gaseous effluents from the site to areas at or beyond l the SITE BOUNDARY shall be limited to the following:

1. For noble gases: less than or equal to a dose rate of 500 mres/yr to the whole body and less than or equal to a dose rate of 3000 mres/yr to the skin, and

2. For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: less than or equal to a dose rate of 1500 arem/yr to any organ, I

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR Part 50,

\ -

?

LA SALLE - UNIT 1 6-19 Amendment No. 93

j _

l ADMINISTRATIVE CONTROLS Fo ONLY -- Nd CNAN'GESD t

PLANT OPERATING PROCEDURES AND PROGRAMS (Continued)

!. 1. Limitations on the annual and quarterly doses to a MEMBER OF THE

. PUBLIC from Iodine-131, Iodine-133 tritium and all radio-i  ; nuclides in particulate form with lialf-lives, greater than 8 days in gaseous effluents released from each unit to areas beyond the SITE BOUNDARY confoming to Appendix I to 10 CFR Part 50,  !

, , J. Limitations on venting and purging of the containment through the lt Primary Containment Vent and Purge Systes or Standby Gas Treatment System to maintain releases as low as reasonably

!l

achievable, ,

l k. Limitations on the annual dose or dose constitment to any MEMBER l

OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR Part 120.

4

5. Radiological Environmental Monitoring Program

! A program shall be provided to monitor the radiation and radionuclides

! in the environs of the plant. The program shall provide (1) represen-l tative seasurements of radioactivity in the highest potential exposure

. pathways, and (2) verification of the accuracy of the effluent j monitoring progree and modeling of environmental exposure pathways.

be contained in the ODCM, (2 j

The guidanceprogram shall p)I to 10 CFR Part 50, and (3) of Appendix D conform include the to the

. following:

! a. , sam ling, analysis, and reporting of radiation and

Monitorin$desi!theenvironmentinaccordancewiththemethod-radionucl 4 ology and parameters in the 00CM, l i A Land Use Census to ensure that changes in the use of areas at j b.

4 and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by j the results of this census, and i

! c. Participation in a Interlaboratory Comparison Program to ensure i that independent checks on the precision and accuracy of the  ;

measurements of radioactive materials in environmental sample

matrices are perfomed as part of ttie quality assurance program for environmental monitoring.

l a

6. Inservice Inspection Program for Post Tensioning Tendons This program provides controls for monitorins any tendon degradation

] in pre-stressed concrete containments, inclucing effectiveness of its j; corrosion protection medium to ensure containment structural j integrity. The program shail include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3,1989, except that the unit 1 and 2 '

primary containments shall be treated as twin containments even though the Initial Structural Integrity Tests were not within 2 years of each other.

The Onsite Review and Investigative Function shall be responsible for reviewing and approving changes to the Inservice Inspection Program for Post Tensioning Tendons.

The provisions of 4.0.2 and 4.0.3 are applicable to the Tendon Surveillance Program inspection frequencies.

LA SALLE - UNIT 1 6-20 Amendment No. 107

k 6NFo mvlyAyo~~ cyAgus) j ADMINISTRATIVE CONTROLS j PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) 1 1

6.2.F.7 Primary Containment Leakaae Rate Testina' Progtg l A program shall be established to in)1ement the leakage rate testing of the f primary containment as required by 10 CFR 50.54foD and 10 CFR 50 Appendix J, i Option B, as modified by approved exemptions. 1s program shali be in 4

accordance with the guidelines entained in Regulatory Guide 1.163

" Performance-Based Containment Leak-Testing Program," dited September 1995.

l The peak calculated primary containment internal pressure for the design basis

loss of coolant accident, P., i's 39.6 psig.

The maximum allowable primary containment leakage rate, L., at P., is 0.635% of primary containment air weight per day.

Leakage rate acceptance criteria are:

a. Primary containment overall leakage rate acceptance criterion is $1.0 L

! During the first unit startup following testing in accordance with this,.

!" pr ram, the leakage rate acceptance criteria are s 0.60 L, tests. for the c ined Type B and Type C tests, and s 0.75 L, for Type A j b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is $0.05 L, when tested at 2 P,.

l 2) For each door, the seal leakage rate is s 5 scf per hour when the gap j between the door seals is pressurized to 2 10 psig.

t i The provisions of specification 4.0.2 do not apply to the test frequencies j specified in the Primary Containment Leakage Rate Testing Prograo.

t

! The provisions of specification 4.0.3 are applicable to the Primary Containment l Leakage Rate Testing Program.

6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVENT IN PLANT OPERATION

! The following actions shall be taken for REPORTABLE EVENTS:

! a. The Commission shall be notified and a Licensee Event Repo> submitted j pursuant to the requirements of Section 50.73 to 10 CFR Part 50, and j j

i

b. Each REPORTABl.E EVENT shall be reviewed b the Onsite Review and i Investigative Function. l i

1  ;

i I

l 4

6-20a Amendment No. 110 LA SALLE - UNIT 1

i, ADMINISTRATIVE CONTROLS i

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 i o>eration shall not be resumed until authorized by the NRC. The conditions of s tutdown shall be promptly reported to the Site Vice 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 (see Section 4 of these i specifications) are being met. All equipment failing to meet surveillance requirements and the corrective action taken shall be recorded;

5. Records of thanges to operating procedures;

6. Shif N.$EE")ogs; 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; 1

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

j

3. Records of new and spent fuel inventory and assembly histories; j I

4. Updated, corrected, and as-built drawings of the plant; LA SALLE - UNIT 1 6-21 Amendment No. 107

ADMINISTRATIVE CONTROLS

! PLANTOPERATINGRECORDS(Continued) @Fo out9-No2Ngd6Q i -. 5. Records of plant radiation and contamination surveys;

6. Records of offsite environmental monitoring surveys;

7. Records of radiation exposure i

contractors and visitors to theforplant, all plant personnel,ithw10 CFRincluding all in accordance

~

Part 20;

8. Records of radioactivity in liquid and gaseous wastes released to

the environment; -

4

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);

]

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

' 11. Inservice inspections of the reactor coolant systes;

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

13. Records of reactor tests and experiments; I 14. Records of Quality Assurance activities required by the QA Manual, except for those items specified in Section 6.5.A; l 15. Records of reviews perfomed'for changes made to procedures on equip-l, ment or reviews of tests and experiments pursuant to 10 CFR 50.59; i 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; j! 17. Records of analyses required by the radiological environmental

monitoring program; I 18. Records of reviews performed for changes made to the 0FFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PROGRAM; and i 19. Records of pre-stressed concrete containment ..

tendon surveillances.

6.6 REPORTING REQUIREMENTS 3

In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following identified reports shall be submitted j

}

i i

1 6-22 Amendment No.100 i -

LA SALLE UNIT 1

(_ - _- -- - . . - -

l .

GNFC ONL'(- No cygg55}

j ADMINISTRATIVE CONTt0L5 4

6.5 REPORTING REQUIREMENTS (Continued) to the director of the appropriate Regional Office of Inspection and Enforce-

, ment unless otherwise noted.

5 l A. Routine Reports i

1. Startup Report .

I 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 supplier, and (4) modifications that may have significantly altered the nuclear, themal, or tqydraulic perfore-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 values i with design predictions and specifications. Any corrective actions 1: that were required to obtain satisfactory operation shall also be described. Arty additional specific details required in license con-ditions based on other commitments shall be included in this report.

Startup reports shall be submitted within (1) 90 days following coe-l plation of the starte test program, (2) 90 d4ys following resumption or cossencement of commercial power operation, or (3) 9 months follow-jj j ing initial criticality, whichever is earliest. If the starte report I

lI does not cover all three events (i.e. , initial criticality, completion of starty test program, and resumption or commencement of commercial j power operation), supplementary reports shall be submitted at least j i overy 3 months until all three events have been completed.

i l 2. Annual Report -

A tabulation shall be submitted on an annual basis prior to March 1 I' of each year of the number of station, utility, and other personnel I

(including contractors) receiving exposures greater than 100 aree/yr and their associated aan rem exposure according to work and job functions (Note: this tabulation supplements the requireuents of

. Section 20.407 of 10 CFR 20), e.g., re, actor operations and surveil-

! 1ance, inservice inspection, routine maintenance, special maintenance

(describe maintenance), waste processing, and refueling. The dose

{ assignments to various duty functions may be estimated based on

! pocket dosimeter, TLD, or film badge measurements. Small exposures 1 totaling less than 20E of the individual total dose need~ not be 1 accounted for. In the aggregate, at least 80K of the total whole 1

bo@ dose received from external sources shall be assigned to specific j mejor work functions.

The results of specific activity analysis in which the primary coolant 4

exceeded the limits of Specification 3.4.5 shall be included in the

, Annual Report along with the f611owing infomation

(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 sagle in which the limit j was anceeded; (2) Results of the last isotopic analysis for radiciodine j LA SALLE UNIT 1 6-23 -

Amendment No. M, 86

ADMINISTRAT!YE CONTROLS l l

1 i performed prior to exceeding the limit, results of analysis while l

I limit was exceeded and results of one analysis after the radiciodine 1 j activity was reduced to less than limit. Each result should include ,

i date and time of sampling and the radiciodine concentrations; (3) l 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 /> prire to the first sample in which the limit was exceeded; (4) Grapt af the I-131 con-centration and one other radiciodine isotope coe,antration in micro- l curies per gram as a function of time for the Aration 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 Radio 1ocical Environmental Coerstine taport' l-The Annual Radic1ogical Environmental Operating Report covering the t operation of the unit during the previous calendar year shall be

!,- submitted before May 1 of each year. The report shall include .

summaries integretations, and analysis of trends'of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistant with the object' ves outlined in (1) the 00CM and (2) Sections IV.B.2. IV.B.3, and IV.C of 3

h 4.

Appendix I to 10 CFR Part 50.

-imana=17Radioactive Effluent Release Report ** SNI l_i s IThe SemiannMhive Effluentleleaseleport covering the

! L operation of the unit during the previous 6 months of operation shall j -  % ;;"=itted wi+h4m 80 daus af+= 3=an--a 1 and .1uly 1 of ameh u-- ,

1 The report shall include a summary of the quantities of.radioact.ive liquid and gaseous effluents and solid waste released.from the unit.

l The meterial provided shall be (1) consistent with the objectives i outlined in the ODCM and PCP and (2) in conformance with 10 CFR l  ; 50.36a and Section IV.B.1 of Appendix I to 10 CFR Part 50.

Monthly Operatino Report l

5. ca.

re.ssces sfeciItb~ ~/0CFR50.4 j Routine reports of operating statistics and shutdown experience, l i including documentation of all challenges to safety / relief valves, j shall be s_ubmitted on t montLiv hasis irector, Orrice o r -

learleactor RegulatFon, lin11ltation -137 US Nuclear Regulatory i

' Commission, Renional Office. Washington,ive ta arr rm inser uan me aw or eact monthDC 20555, fo' Towing the calendar month covered by the report.

l

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

• • A single siteittal may be made for a multi-unit station. The sL&mittal i should combine those sections that are common to all units at the station; i however, for units with separate radwaste stems, the sitsnittal shall j specify the releases of radioactive matori from each unit.

l l LA SALLE UNIT 1 6-24 Amendment No. 55, 86 l

i i

, . l l

gMINISTRATIVECONTROLS j Semiannual Radioactive Effluent Release Report (Continued) i s 7Anfh

<wi ch the thechanges Monthly Operating to the OFFSITE ReportVOSE withinCALCULATION 90 days in wh<MANUA i

dangelshwalmadtaffar+iva f= " +4aa a.Pecort~of any ma:_or 4

changes to the radioactive waste treatment sys'. ems snali ce sutettte

' with the Monthly Operating Report for the period in which the i

evaluation was reviewed and accepted by Onsite' Review and Investigative Function.

A

6. Core Doeratina Limits Recort i

a. Core o>erating limits shall be established and documented in the i

CORE 0'ERATlHG LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

j (1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

! (2) 1 The minimum Critical Power Ratio (MCPR)itsscram time, ta and power and flow de

3.1.3. pendent MCPR limits) for Technical $pecification

(3) The Linear Heat3.2.4.

Specification Generation Rate (LHGR) for Technical

( (4) The Rod Block Monitor Upscale Instrumentation Setpoints for '

4 Technical Specification Table 3.3.6-2.

b. The analytical > methods used to determine.the core operating '

l limits shall be those previously reviewed and approved by the i

i NRC in the latest approved revision or supplement of the topical reports describing the methodolo t

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

~

(1) j NEDE-24011-P-A, Reactor Fuel," ( latest approved revision)." General Ele 4

I (2) Consonwealth Edison Topical Report NFSR-0085, " Benchmark of 3 BWR Nuclear Design Methods,"..(latest approved revision).

(3) Connonwealth Edison Topical Report NFSR-0085, Supplement 1,

" Benchmark of BWR Nuclear Design Methods - Quad Cities j Gamma Scan Comparisons," (latest approved revision).

a (4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2,

" Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

a i

i i -

i LA SALLE - UNIT 1 6-25 Amendment No.103 li t

,i i .

! ADMINISTRATIVE CDNT110L5  !

1

3 c. The core' operating limits shall be detemined so that all applicable limits (e.g., fuel thermal-mechanical limits, core 1

J thermal-tydraulic limits ECCS limits, nuclear limits such as  !

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

the safety analysis are met.

- ~~

j ,

, d. ' The CORE OPERATING LIMITS REPORT, including ary std-cycle I revisions or supplements thereto, shall be provided upon j issuance, for each reload cycle, to the U.S. Nuclear Regulatory

Commission Document Control Desk with copies to the Reg < onal l Adeinistrator and Resident Inspector.

l .

I* *

, Pegbl Adm'njsMr oMe.

NRCBegkal OSSke.

C. Unique Reporting Requirements - - g

1. Special Reportsahall_be_ s_ubmitted to the 61 rec'ar of_the 0"f1 A poc". Ton anif Enforcement TRecTon IIGwithin P.he UssFper10d l spec' f en for eacn ymport.

6.7 PROCESS C00mt0L PROGRAM (PCP)*

l 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 perfomed shall be retained as required by Specification 6.5.B.18. This doc eentation shall j contain: l

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

l

and .

I

2) A determination that the change will maintain the overall con-

. fomance of the solidified waste product to existing requirements j of Federal, State, or other applicable regulations.

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

1 I

i 1

4 i

l _

1 *The Process Control Program (PCP) is common to La Salle Unit 1 and La Salle

Unit 2.

i i

LA SALLI UNIT 1 6-26 Amendeont No. 85, 86 f

i

l ADMINISTRATIVE CONTROLS 6.8 0FFSITE DOSE CALCULATION MANUAL (0DCM)*

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

6.8.2 Licensee initiated changes to the ODCM: I

a. Shall be documented and records of reviews performed shall be. retained as required by Specification 6.5.8.18. This documentation shall l!

contain: .

1) Sufficient information to support the change together with the  !

appropriate analyses or evaluations justifying the change (s), and A determination that the change will maintain the level of radio-2) active effluent control required by 10 CFR 20.106, 40 CFR i Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and i not adversely impact the accurac

{

dose, or setpoint calculations. y or reliability of affluent,

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

c. Shall be submitted to the Commission in the form of a cosplete, leg-i ,ible cooy of the entire ODCM as a part of or concurrent with the non' annualRadioactive Effluent Release Report for the period of the report iri which any change to the 00CM was made effective. Each change shall be identified by marking's in the margin of the affected -

Ahhusl clearly indicating the area e the pages, shall indicate the date (e.g., month / year) pay was the change thatimplemented.

was changed, and 6.9 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS 6.9.1 Licensee initiated major changes to the radioactive waste treatment

! systems (liquid, gaseous and solid):

a. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the Onsite 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;

2. Sufficient detailed infomation 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; "The DFFSITE DOSE CALCULATION MANUAL (0DCM) is common to La Salle Unit 1 and La Salle Unit 2.

LA SALLE UNIT 1 6-27 Amendment No. 85 t

- ~ =-

l '

CINFO ONLY-No CHANG @

j ADMINISTRATIVE CONTROLS I

MAJORCHANGESTORADI0ACTIVEWASTETREATMENTSYSTEMS(Continued) 1

4. An evaluation of the chanin which shows the predicted releases j of radioactive materials ' n liquid a.W gaseous effluents and/or quantity of solid waste that Mffer from those previously

! predicted in the license application and amendments thereto; 1

i 5. An evaluation of the change which shows the expected maximus exposures to individual in the unrestricted 4tte and to the general population that differ from those previously estimated

} in the license application and amen 3nents thereto; l

l 6. Ae arison of the predicted releases of radioactive asterials, .

in li uid and gaseous effluents and in-solid waste, to the

' actua releases for the period to when the changes are to k uk; -

i

! 7. An estimate of the exposure to plant operating personnel as a j result of the change; and

8. Documentation of the fact that the change was reviewed and i

found acceptable by the Onsite P.eview and Investigative Function.

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

i

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a i

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I LA SALLE UNIT 1 6-28 Amendment No. 85

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

1 l LINEAR HEAT GENERATION RATE

.l 1.22 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit i length of fuel rod. It is the integral of the heat flux over the heat 1 transfer area associated with the unit length. LNGR is monitored by the j ratio of LNGR to its fuel specific limit, as specified in the CORE OPERATING LIMITS REPORT.

1 LQfd.C 1Y1IHLEUNCTIONAL TEST 1.23 A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, j 1.e, all relays and contacts, all trip units, solid state logic elements,

! etc. of e logic circuit, from sensor through and including the actuated

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

l 1.24 Deleted 1 MEMBERS (S) 0F THE PUBLIC i 1.25 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally j associated with the plant. This category does not include employees of the '

j licensee, its contractors, or vendors. Also excluded from this category are l 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.

l MINIMUM CRITICAL POWER RATIO

! 1.26 The MINIMUN CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which

exists in the core.

1

! 0FFSITE DOSE CALCULATION MANUAL 1.27 The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology 3

and parameters used in the calculation of off. site doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous j and liquid effluent monitoring Alarm / Trip Setpoints, and in the conduct i of the Environmental Radiological Monitoring Program. The 00CM shall 1 also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Technical Specification Section 6.2.F.4 and (21 descriptions of the information that sho 4 included in the Annual Radiological Environmental Operating and emi-

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

l

]

LA SALLE - UNIT 2 1-4 Amendment No. 101

5.0 DESIGN FEATURES Wk~ @ CM kk 4

5.1 SITE

+

EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

j LOW POPULATION ZONE 1

5.1.2 The low population zone shall be as shown in Figure 5.1.2-1. j j SITE BOUNDARY FOR GASE0US EFFLUENTS l 4

l 1

5.1. 3 The site boundary for gaseous effluents shall be as shown in Figure 5.1.1-l'.

{

I SITE BOUNDARY FOR LIQUID EFFLUENTS

+

5.1. 4 The site boundary for liquid affluents shall be as shown in Figure 5.1.1-1.

f 5.2 CONTAINMENT CONFIGURATION i 5.2.1 The primary containment is a steel lined post-tensioned concrete i

structure consisting of a drywell and suppression chamber. The drywell is a i

steel-lined post-stressed concrete vessel in the shape of a truncated cone closed 4

I by a steel dome. The drywell is above a cylindrical steel-lined post-stressed concrete suppression chamber and is attached to the suppression chamber through i a series of downcomer vents. The drywell has a minimum free air volume of i 229,538 cubic feet. The suppression chamber has an air region of 164,800 to j 168,100 cubic feet and a water region of 128,800 to 131,900 cubic feet.

DESIGN TEMPERATURE AND PRESSURE l 5.2.2 The primary containment is designed and shall be maintained for:

j a. Maximum internal pressure: 45 psig.

b. Maximum internal temperature

drywell 340*F.

4 suppression chamber 275'F.

c. Maximum external pressure: 5 psig. -

1

d. Maximum floor differential pressure: 25 psid, downward.

5 psid, upward.

SECONDARY CCNTAINMENT i

5.2.3 The secondary containment consists of the Reactor Building, the equipment access structure and a portion of the main steam tunnel and has a minimum free volume of 2,875,000 cubic feet.

4

LA SALLE - UNIT 2 5-1 l

~

j ONLY-NO CHANGC Illinois River

^

~IXCLUSION AREA AND

=

h, SITE BOUNDARY FOR GASEOU LIQUID Liquid Effluent Discharge Point 7

1  :

20 O.

n 00 ;

1 Mile i l

Scale in Feet '

l N Site Boundary)

(Property Line

~

Waste Stabilization .

Pond I

/ l

- i i

i Exclusion Area

• I

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LaSalle' Lake

' i i

IStatio Venti dJ

(,

I Stack  :

1 I L_____J Figure 5.1.1-1 LA SALLE - UNIT 2 5-2 Amendment No. 69

DESIGN FEATURES 5.3 REACTOR CORE l 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 '

i l natural or slightly enriched uranium dioxide The bundles may contain water rods or water boxes.(UO ) as fuel material.I.initedsub

Zircalloy or IIRLO or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel l assemblies shall be limited to those fuel designs that have been analyzed with l 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 l 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 (8 C) and/or hafnium metal. Thecontrolrodassemblyshallhaveanominalaxiaiabsorber l 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 maintaines

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.

l 2. 1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valv,e.

l 3. 1500 psig from the discharge shutoff valve to the jet pumps.

c. For a temperature of 575'F.

VOLUME 5.4.2 The system is -total water 21,000 andfeet cubic steam at avolume nominalofTthe ofreactor vessel and recirculation 533*F.

m_

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5.5 EF Kr 9-^^

c n : ~CEDELET@

l GiJi.T The sieteoMial tower shall be located as shown on Figure 5.L LA SALLE - UNIT 2 5-4 Amendment No. 101

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 descriptions of departmental responsibilities and relationships, and l Job descriptions for key personnel positions, or in equivalent forms '

of documentation. These requirements shall be documented in the 1 Quality Assurance Manual.

hSEU "A"i m -.+~ '

2. Kha station ManaanB 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 l 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, ejsf w p g l B. The Shift mme~1en* shall b/ responsible for directing and commanding i the overall operation of thefracJlity on his shift. The primary management

} responsibility of the Shiftmmarvisorisha11 be for safe operation A l

of_the nuclear facility on h_is shift unde _r _all condition 1/A management)

/triracuve ugnea ey tne ute Vice Prestcent empnasizing this primary I

management responsibility and that clearly establishes the command duties

! of the Shift Supervisor shall be reissued to all station personnel on a nnual basis./ -

j _

C. The shift manning for the station ___sha_11 be a_s shown Jn_ noure s.1-2_

Tndividual filling tne posmon or systems Eng' neering Supervisor shall j

l

[

meet the minimum acceptable level for " Technical Manager" as described in

Section 4.2.4 of ANSI N18.1-1971. The individual filling the position

of Operations Manager shall meet the minimum acceptable level- for " Plant j (Manager" as dateHhed in Section 4.2._1 of ANSI N18.1-1971.

s' i

l LA SALLE - UNIT '2 6-1 Amendment No. 93

~

• N Mawa.3er; 0%e. Ohhhtr vi,5or j Ne kk lednice NL/So

> r ,

I bcc-:"ZTi"?f" DMwiSTRATIVECONTROLS 7 At least one licensed Reactor Operator shall be in the control room I

1.

4 when fuel is in the reactor. In addition, while the reactor is in

! OPERATIONAL CONDITION 1, 2 or 3, at least one licensed innior A i Reactor Operator who has been designated by the Shift Kunervtolito ,

l assume the control room direction responsibility shall be in Te (

j Control Room. g i j 2. A radiation protection technician

• shall be on site when fuel is un-the reactor. .

l l 3. All CORE ALTERATIONS shall be observed and directly supervised by

< either a licensed Senior Reactor Operator or Senior Reactor Operator l l Limited to Fuel Handling who has no other concurrent responsipilities. l j during this operation.

f .

4. A site Fire Brigade of at least 5 members shall be maintained onsite j , at all times *. The Fire Brigade- shall not includeVEe"'IKTYte*. .

j & ?arvisor. the Station coni.rui Au- r ----n and the 2 other members l

Iof the minimum sh' ft crew necessary for safe shutdown of the unit and I i any personnel required for other essential functions during a fire l l

emergency. l l S. The Independent Safety Engineering Group (ISEG) shall function to l 4 examine unit operating characteristics, NRC issuances, industry

advisories, Licensee Event Reports and other sources of plant design l i and operating experience information, including plants of similar i design, which may indicate areas for improving unit safety. The ISEG l l shall be composed of at least three, dedicated, full-time engineers of j multi-disciplines located on site and shall be augmented on a part-  !

j time basis by personnel from other parts of the Cosmonwealth Edison

Company organization to provide expertise not represented in the

group. The ISEG shall be responsible for maintaining surveillance l l 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 l

} for revised procedures, equipment modifications, maintenance activi-ties, ooerationn activities or other means of improving unit safety j CIMSERT "Eh to the m te Qua' ity Vertricanon virector una sne usuon nanagel"- f A l

, 6. he Station Control Roca Engineer pt.Mt) may serve as the Shift

! [ Technical Advisor (STA) during abnormal operating and accident c

hSERT T- - -

tions. During these conditions, the SCRE or other on duty STA shall l

provide thermal hydraulics, technical (to the safe operation of the unit.

support reactor to and engineering theplant Shift Supervisor analysis with re in the a 4 .

i *The radiation protection technician and Fire Brigade composition may be less

! than the minimum requirements for a period of time not to exceed two hours in 1 order to acconnodate unexpected absence provided immediate action is taken to j fill the required positions.

i l( #Not responsible for sign-off feature.

1 5

IA sal M - UNIT 2 6-2 Amendment No. o3

_ _ _ . _ . . _ __ __. ..___ __ . . _ . . . _ _ _ _ _ ___ _ . . _ _ _ . ~ . _ _ _ _ _ _ _ . .

4

~

4 ATTACHMENT B I PROPOSED AMENDMENTS TO THE '

j LICENSEITECHNICAL SPECIFICATIONS I

j Insert B Safety Assessment / Site Quality Verification Manager and the Plant Manager.

Insert C 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.

B-6

t ATTACHMENT B PROPOSED AMENDMENTS TO THE LICENSE / TECHNICAL SPECIFICATIONS Insert A l The individual filling the ANSI N18.11971 Section 4.2.1 position of Plant Manager (" Plant Manager"),

l 1

l l

l l

l i l

j B-5 t

I

i

ADMINISTRATION CONTROLS i

To assure capability for performance of all STA functions:

! a. The shift foreman (SR0) shall participate in the SCRE shift relief i

turnover.

b. During the shift, the shift engineer and the shift foreman (SRO)

! shall be made aware of any significant changes.in plant status in a timely manner by the SCRE.

l c. During the shift, the shift engineer and the shift foreman (SRO)

shall remain abreast of the current plant status. The shift fore-i man ($90) shall return to the control roos two or three times per j shift, where practicable, to confer with the $CRE remrding plant

status. Where not practical to return to the contro' roce, the f shift foreman (SRO) shall periodically check with the SCRE for a plant status updata. The shift foreman (SRO) shall not abandon duties critical to reactor operation, unless specifically ordered by the shift engineer.

}

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i LA SALLE - UNIT 2 6-2a Amendment No. 62

dhM W\ STRATI M

_._.... ... ,=diimF0LS

7. The amount of overtime 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 Sepe'rvisor shall hold a l Senior Reactor Operator License. /

D. Qualifications of the station management and o[erating staff shall meet minimum acceptable levels as described in ANSI N18.1, " Selection and Training of Nuclear Power Plant Personnel,' dated Marr.h 8,1971. The Health Physics Supervisor shall meet the r9quirements of radiation pro-taction manager of Regulatory Guide 1.8, feeptember 1975. The ANSI N18.1-1971 qualification requirements fo9 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 worktng experience in the special+v, 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, "Supervis6r 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.

(

LA $AllE - UNIT 2 6-3 Amendment No. 93

l ADMINISTRATIVE CONTROLS i

. G. DELETED (The Review and Investigative Function and the Audit Function are l described in the Quality Assurance Manual Topical Report CE-1-A).

l rN Fo oNLY, I No CHANCES l

1 1

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l l

l 1

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(

Ut SALLE - UNIT 2 6-4 Amendment No. 93 (Next page is 6-13)

j ..gmerry,w .

l - \

i 1 I

J v.

) Floure 6.1-3 - -

3 MINIMUM SHIFT CREW CONPOSITION- j 4  !

WITM LAIIT 1 IN GONDITION 1. Z, OR 3 l i

P051TIDM IRABER OF INDIVIDUAL 5 REQUIRED TO FILL POSITION C02ITIONS 1, 2 and 3 l i CONDITIONS 4 and 5

• M l l' l' l sr l' None  !

l M N '

1 l M N 1 l .sCm 1* .- -

None .

.i 1 .

or, whenever a SCRE (SH0/STA) is not included in the shift crew j composition,.the minis m shift crew composition shall be as j follows:

l i I

WITH UNIT 3. IN GONDITION 1, 2, OR 3 i PD5U10K j

g MLMIER OF INDIVIDUAL 5 REQUIRED TO FILL P051TIDM l COMITIONS 1, 2 and 3 COW ITIONS 4 and 5  !

SE l' a l

8 SF 1 None

.M N 1 M E 1

8 l f FA -

1 hw  !

'4 i wtIn. untI 1 In wnuluoM 4 OR 5 UR DEFUELED P0alIl0m. ..=3 = 9 0F I m Iv1 DUAL 5 REQUINtp TO FILL PO5ITION J J: t' COMITI0it51, 2 and 3 COWITIONS 4-and-5_ 1 h SE .

1" 1"

, $r .1 None R0 . 1 1 5

. AD 1-t f,4 STA 1 '

None LA SALLE - UNIT 2 -

6-13 Amendment No. 47

. _ - . . . -. ~

1 l

l 1

l ATTACHMENT B l PROPOSED AMENDMENTS TO THE l LICENSEITECHNICAL SPECIFICATIONS l

Insert D FIGURE 6.13 l MINIMUM SHIFT CREW COMPOSITION *)

POSITION *) MINIMUM CREW NUMBER l

t EACH UNIT IN ONE UNIT IN CONDITION 1, EACH UNIT IN CONDITION 4 CONDITION 1,2, OR 3 2, OR 3, AND ONE UNIT IN OR 5 OR DEFUELED CONDITION 4 OR 5 OR DEFUELED l SM 1 1 1 l 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 l of Figure 8.1-3 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, i

, (b) Table Notation:

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

i 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 l hold a license for both units. Otherwise, for each unit, provide a relief operator who holds a license for the unit l asshned.

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 i control room command function. With both Units in CONDITION 4 or 5, an individual with a valid SRO or RO license I shall be designated to assume the control room command function.

(d) The STA position may be filled by any individual who meets the Commission Policy Statement on Engineering Expertise on Shift.

B-7 i

l l

l

{ .

i 3 , ..

] Figure 6.1-3 (Continued) '

! MINIMLM SHIFT CREW COMPOSITION .

j -

)

NOTES I -

E ndividual I say fill the same position on Unit 1.

E 0ne of the two required individuals may fill the same position on Unit 1.

j SE - Shift Supervisor (Shift Engineer) with a Senior Reactor operators License on Unit 2.

i l SF - Shift Foreman with a Senior Reactor Operators' License on Unit 2.

4 A0 - Individual with a Reactor Operators License on Unit 2.

! AO - Auxiliary Operator.

i SCRE - Station Control Room Ertgineer with a Senior Reactor Oper'ators Licents. .

! Except for tne Shift Supervisor, the Shift Crew Composition'aey'be one less

' than the minimum requirements of Figure 6.1-3 for a period of time not to

{ exceed 2hoursinordertoaccommodateuneapectedabsenceofondutyshift i crew members provided immediate action is taken to restore the shift crew i

composition to within the sinima requirements of Figure 6.1-3. This '

3 provision does not permit any shift crew position to be unmanned upon shift J

change due ta an oncoming shift creuman being late or* absent. .

While the unit is in OPERATIONAL COMITI0ft 1, 2, or 3, an individual with a j valid SRQ license shall be designated to assume the Control Roos direction i function. While the unit is in OPERATIONAL CDN0lTION 4 or 5, an individual i with a valid SR0 or R0 license shall be designated to assume the Control Roos frection function. *

! 1 1 * -

i i .

.17fp $446' TVretdTroNALL l -. -f.- .

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9. .! -LA SALLE - UNIT 2 p14 - Amendment No. 47

j i l

I . WESTMT

, "";%: T'".TIO CONTROLS 6.1.1 HIGH RADIATION AREAS 6.1.1.1 Pursuant to Paragraph 20.203(c)(5) of 10 CFR 20, in lieu of the " con- -

trol device" or " alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, 4

each high radiation area in which the intensity of radiation is greater than 100 area /hr* but less than 1000 aren/hr* shall be barricaded and conspicuously j

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 t%'s l individuals, may be exempt from the RWP issuance requirement during the per-l formance of their assigned duties in hi h radiation areas in which the intensity of radiation is greater than 100 aren/h but less than 1000 aren/hr*, provided

! they are otherwise following plant radiation protection procedures for entry 1 ,

into such high radiation areas. Any individual or groep of individuals pemitted I to enter such areas shall be provided with or accompanied by one or more of the l following:

a. A radiation monitoring device which continuously indicates the l radiation dose in the area.

b. A radiation monitoring device which continuously integrates the i

radiation dose rate in the area and alams when a preset integrated i

dose is received. Entry into such areas with this monitoring device i any be made after the dose rate level in the area has been established

! and personnel have been made knowledgeable of them.

i j c. A health physics qualified individual, i.e., qualified in radiation

{ protection procedures, with a radiation dose rate monitoring device, 4

who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at I the frequency specified by the Health Physicist in the Radiation l Work Pemit (RWP).

• 6.1.1. 2 In addition to tim requirements of 6.1.1.1, above, for areas accessible to personnel with radiation levels such that a na or portion of the body could receive in one hour a dose greater than 1000 m , the computer shall be prograaned to permit entry through locked doors for any individual requiring l access to arty such High-High Rad'ation Areas for the time that access is required.

i 6.1.1.3 Keys to manually open computer controlled High Radiation Area doors and High-High Radiation Area _ doors sh611 be maintained under the Administra -

tion control of the Shift QervisofE85 duty and/di the Health Pitysicist.

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 <ndividual 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, i direct surveillance to prevent unauthorized entry may be substitutad. Doors shall remain lockea except during periods of access by personnel under an I approved RWP which shall specify the dose rate levels in the immediate work j area and the maximum allowable stay time for individuals in that area. For i

I

}

(

l i LA SALLE - UNIT 2 6-15 Amendment No. U , 70 i

INFO CML'1 -

N o cHA M Es ADMINISTRATIVE CONTROLS HIGH RADIATION AREAS (Continued) individual areas accanible to personnel with radiation levels such that a major portion of the bo @ could receive in one hour a dose in excess of 1000 arem* that are located within large areas, such as the containment, where no enclosure exists for purposes of locking, and no enclosure can be reasonably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote, such as use of closed circuit TV cameras, continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

6.2 PLANT OPERATING PROCEDURES AND PR0GRANS ,

A. Written procedures shall be established, implemented, and maintained covering the, activities referenced below:

a. The applicable procedures recommended in Appendix A, of Regulatory Guide 1.33, Revision 2, February 1978,

b. The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement I to B) REG-0737 as stated in Section 7.1 of Generic Letter No. 82-33,

c. Station Security Plan implementation,

d. Generating Station Emergency Response Plan iglementation,

e. PROCESS CONTROL PROGRAN implementation,

f. OFFSITE D0SE CALCULATION MANUAL implementation, and

, g. Fire Protection Program implementation.

/

• Measurement made at 18" from source of radioactivity.

N .

1 LA SALLE - UNIT 2 6-16 Amendment No. #7, 70

j' i

ADMINISTRATIVE CONTROLS i

< PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) 1 B. Radiation control procedures shall be maintained, made available to all

station personnel, and adhered to. These procedures shall show permissible j radiation exposure and shall be consistent wtth.the requirements of ,

i 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.B and her procedures i which affect nuclear safety, as determined by the ~amnanager, and changes thereto, other than editorial or typographical changes, shall be i reviewed as follows prior to implementation except as noted in g

Specification 6.2.D:

l 1. Each procedure or procedure change shall be independently reviewed by i a qualified individual knowledgeable in the area affected other than j 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 i i reviews shall be performed by the qualified review personnel of the '

i appropriatediscipline(s).

2. Individuals performing these reviews shall meet the acclicable -

experience requirements of ANSI N18.1-1971, Secti s .2 , and j

be approved by the Ga"IT5h Manager.g .

4. pd ' . 'j

_3 M.S.{or4.6 3

3. Applicable Administrative Procedures recommended by Regulatory Guide i

1.33, Plant Emergency Operating Procedures, and changes thereto shall l

be submitted to the Onsite Review and Investigative Function for j review and approval prior to implementation.

l 4. Review of the procedure or procedure change will include a j i

determination of whether or not an unreviewed safety question is j involved. This determination will be based on the review of a written i 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 j questions shall be obtained prior to Station approval for

implementation. -

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

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

1

7. Editorial and Typographical changes shall be made in accordance with

~

j station procedures.

C l

1 1 LA SALLE - UNIT 2 6-17 Amendment No. 93 i

?

! ADMINISTRATIVE CONTROLS i

I D. Temporary changes to procedures 6.2.A and 6.2.8 above may be made l

provided: ,

l 1. The intent of the original procedure is not altered.

! 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. -

i 3. The change is documented, reviewed and approved in accordance with_

E.

Specification 6.2.C. within 14 days of implementation.

Drills of the emergency procedures described in Specification '.L?d U y,q

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, consnunication links are tested and outside agencies i are contacted.

l F. The following programs shall be established, implemented, and maintained:

1. PrimaN Coolant Sources Outside Primary Containment '

l i A program to reduce leakage from those portions of rystems outside j primary containment that could contain highly radioactive fluids

curing a serious transient or accident to as low as practical levels.

l The systems include LPCS, HPCS, RHR/LPCI, RCIC, hydrogen recombiner,

process sampling, containment monitoring, and standby gas treatment systems. The program shall include the following l a. Preventive maintenance and periodic visual inspection require-l ments, and _

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

j 2. In-Plant Radiation Monitoring A program which will ensure the' capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the followings

a. Training of personnel, ~

b. Procedures for monitoring, and

c. Provisions for maintenance of sas[ ling and analysis equipment.

3. Post-accident Sampline A program which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere saspies 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.

(

LA SALLE - UNIT 2 6-18 Amendment No. Sf, 70

w. ,. - - -,

l ADMINISTRATIVE CONTROLS U~ ANN lj i

PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) 4

4. Radioactive Effluent Controls Proaram i

A program shall be provided conforming with 10 CFR 50.36a for the i control of radioactive effluents and for maintaining the doses to i MEMBERS OF THE PUBLIC from radioactive affluents as low as reasonably 4 achievable. The program 1 21 shall be implemented by op(er)ating proceduresshall and (3 shall be contained in th

'. remedial actions to be taken whenever the pro, gram 11m)its are exceeded. The program shall include the following elements:

I a. Limitations on the operability of radioactive liquid and gaseous i

l monitoring instrumentation includino surveillance tests and set-soint determination. in accordance with the methodology in the i

M,  ; - ,

9 j b. Limitations on the concentretions of radioactive material i released in liquid effluents to UNP.ESTRICTED AREAS conforming to j

10 times the concentration value in Appendix:8, Table 2 Column 2 to 10 CFR 20.1001-20.2402, j

4

c. sampling j -

Monitoring,luents gaseous eff in,accordance with 10 CFA40.4302and and with the analysis methodology and parameters in the 00CM, l-h d. Limitations on the annual and quarterly doses or dose commitment j

i to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released from each unit to UNRESTRICTED AREAS conform-j ing to Appendix 1 to 10 CFR Part 50, i e. Determination of cumulative and >rojected dose contributions i from radioactive effluents for t te current calendar quarter and current calendar year in accordance with the methodology and j parameters in the ODCM at least every 31 days, f.

Limitations on the operability and use of the liquid and gaseous

' effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a 31-day period would exceed 2 percent of the guidelines for the annual dose or dose  !

j commitment conforming to Appendix I to 10 CFR Part 50, i g. Limitations on the dose rate resulting from radioactive materials j released in gaseous effluents from the site to areas at or beyond i

the SITE B0UNDARY shall be limited to the following:  ;

l 1. For noble gases: less than or equal to a dose rate of 500  ;

ares /yr to the whole body and less than or equal to a dose j rate of 3000 ares /yr to the skin, and ,

2. For Iodine-131 Iodine-133, tritium and for all j radionuclides In particulate form with half-lives greater than 8 days: less than or equal to a dose rate of 1500 arem/yr to any organ, l

) h. Limitations on the annual:and quarterly air doses resulting from 1

noble gases released in gaseous affluents from each unit to areas beyohd the SITE BOUNDARY conforsing to Appendix I to

} 10 CFR Part 50, 4'

d LA SALLE - UNIT 2 6-1g Amendment No. 77

l j ADMINISTRATIVE CONTROLS QFO ONLY-NQ CHAWG5S} ^ --

j PLANT OPERATING PROCEDURES AND PROGRAMS (Continued) i 1. Limitations on the annual and quarterly doses to a MEMBER OF THE PUBLIC from Iodine-131, Iodine-133, tritium, and all l

radionuclides in particulate form with half-lives greater than 8

, days in gaseous effluents released from each unit to areas

, beyond the SITE BOUNDARY conforming to Appendix I to 10 CFR i Part 50, i

! l l J. Limitations on venting and p wging of the containment through the Primary Containment Vent and Purge System or Standby Gas j Treatment System to maintain releases as low as reasonably

achievable,

k. Limitations on the annual dose or dose commitment to any MEMBER ,

i 0F THE PUBLIC due to releases of radioactivity and to radiation  !

l from uranium fuel cycle sources conforming tn 40 CFR Part 190. I j 5. Radiological Environmental Monitoring Program I

l l A program shall be provided to monitor the radiation and radionuclides j in the environs of the plant. The program ahall provide (1) represen- i i tative measurements of radioactivity in the highest potential exposure l j pathways, and (2) verification of the accuracy of the effluent l 3

monitoring program and modeling of environmental exposure pathways. l i The program shall (1) be contained in the ODCM, (2) conform to the l

) guidance of Appendix I to 10 CFR Fart 50, and (3) include the j j fo11cwing:

i a. Monitoring, sampling, analysis, and reporting of radiation and

! radionuclides in the environment in accordance with the method-

! ology and parameters in the ODCM,

b. A Land Use Census to ensure that changes in the use of areas at j and beyond the. SITE BOUNDARY are identified and that i modifications to the monitoring program are made if required by j .the results of this census, and

c. Participation in a Interlaboratory Comparison Program to ensure ~

. that independent checks on the precision and accuracy of the  ;

8 measurements of radioactive materials in environmental sample i matrices are performed as part of the quality assurance program i for environmental ronitoring.

6. Inservice. Inspection Program for Post Tensioning Tendons This program provides controls for monitoring any tendon degradation .

in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity. The program shall include baseline measurements prior to initial operations. The Tendon Surveillance Program, inspection .

frequencies, and acceptance criteria shall be in accordance with i Regulatory Guide 1.35, Revision 3, 1989, except that the unit I and 2 primary containments shall be treated as twin containments even though LA SALLE - UNIT 2 6-20 Amendment No. 93

l ADMINISTRATIVE CONTROL 5 QFO ONLY- AlO CHANGES) i l PLANT DPERATINS PROCEPURES AND PROGRAMS (Continued)

{

l the Initial Structural Integrity Tests were not within 2 years of each other.

l i The Onsite Review and Investigative Function shall be responsible for

i. reviewing and approving changes to the Inservice Inspection Program i for Post Tensioning Tendons.

The provisions of 4.0.2 and 4.0.3 are applicable'to the Tendon i

. Surveillance Program insepetion frequencies. i i

j 6.2.F.7 primary Containment Leakana Rate Testina Procram i i

! A program shall be established to implement the leakage rate testing of the

primary containment as required by 10 CFR 50.54(o) and 10 CFR 50 Appendix J,  ;

! Option 8, as modified by approved exemptions. This program shall be in i

accordance with the guidelines contained in~ Regulatory Guide 1.163, "Perfonnance-Based Containment Leak-Testing Program," dated September 1995.  ;

i ,

j The peak calculated primary containment internal pressure for the design basis  ;

loss of coolant accident, P , is 3g.6 psig.

The maximum allowable primary containment leakage rate, L,, at P,, is 0.635% of l

[ primary containment air weight per day.

4 l- Leakage rate acceptance criteria are:

l

< a. Primary containment overall leakage rate acceptance criterion is $1.0 L i During the first unit startup following testing in accordance with this,.

pr for the c ran, ined the Typeleakage 8 and Type rate Cacceptance tests, and scriteria 0.75 L, are s 0.60 for Type A L, tests.

b. Air lock testing acceptance criteria are:

i j 1) Overall air lock leakage rate is 50.05 L, when tested at 2 P,.

1 1 2) For each door, the seal leakage rate is s 5 scf per hour when the gap

! between the door seals is pressurized to 2 10 psig.

I

' The provisions of specification 4.0.2 do not apply to the test frequencies specified in the Primary Containment Leakage Rate-Testing Program.

, The provisions of specification 4.0.3 are applicable to the Primary Containment

Leakage Rate Testing Program.

i 6.3 ACTION TO BE TAKEN IN THE EVENT OF A REPORTABLE EVD!T IM PLANT OPERATION

! I

! The following actions shall be taken for REPORTABLE EVENTS:  !

a. The Comeission shall be notified and a Licensee Event Report i submitted pursuant to the requirements of Section 50.73 to j 10 CFR Part 50, and

! b. Each REPORTABLE EVENT shall be reviewed pursuant to i Specification 6.1.G.2.c(1).

1 l

] LA SALLE - UNIT 2 6-20a Amendment No. 95

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 innediately pursuant to Specification 2.1.1, 2.1.2 and 2.1.3 and critical reactor o>eration shall not be resumed until authorized by the NRC. The conditions of s sutdown shall be promptly reported to the Site Vice 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 DPERATING 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 of operation at each power level;, including power levels and periods

2. Records of principal maintenance and activities, including inspection i and repair regarding principal items of equipment pertaining to  ;

nuclearsafety;

3. Records and reports of reportable events;

4. Records and periodic checks, inspection and/or calibrations performed see Section 4 of these to verify that the specifications) aresurveillance being met. Allrequirements equipment (failing to meet i

i surveillance requirements and the corrective action taken shall be recorded;

5. Records of chances to operating procedures; 6.

P%n3 Shift (u..M i....ef O ogs; and

7. Byproduct material inventory records and source leak test results.

8. 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; 7.

Records of radiation contractors exposure and visitors to theforplant, all plant personnel,ithw10 CFRincluding all in accordance Part 20;

'. 8. Records of radioactivity in liquid and gaseous wastes released to the environment; LA SALLE - UNIT 2 6-21 Amendment No. 93

/

N_M..

m nots

) ;

PLANT OPERATING RECORDS (Continued)

- 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); ,

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

11. Inservice inspections of the reactor coolant system;

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

13. Recortis of reactor tests and experiments;

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

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 i required by Specification 3.7.9 including the date at which the l

service life commences and associated installation and maintenance .

records; i

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

} Records of reviews performed for changes ma'de to the 0FFSITE DOSE ,

! 18. I l CALCULATION MANUAL and the PROCESS CONTROL PROGRAN; and

!. I I

19. Records of pre-stressed concrete containment tendon surveillances. i 6.6 REPORTING REQUIREMENTS Code of '

l In addition to the applicable reporting reavirements of Title 10,bmitted

Federal Regulations the following identiffad reports shall be su l

tothedirectorofthea ment unless otherwise ed. nogpropriate Regional Office of Inspection A. Routine Reports j

1. Startup Report ,

s l A summary report of plant startup arid power escalation testing I 1

j shall be submitted following (1) receipt of an operating Itcense, i

(2) amendment to the license involving a planned increase in 4

power installation of level, (3)factured.by a fueldifferent that has fuel asupplier, differentand desi(4) gn

!' or has been manu J modifications that may have significantly altered the nuclear $all thermal, or hydraulic performance of the plant. The report s f in general include a description of the measured values of the 1

operating conditions or characteristics obtained during the test i

program and a e rison of these values with design predictions and l

specifications. corrective actions that were required to obtain satisfactory operat on shall also be described. Any additional

' specific details required in license conditions based on other 1

commitments shall be included in this report. '

1 sa catir tstT 2 6-22 Amendment No. 84

} 8WTNDTih

. ~

.. -,o. - '"

CONTROLS l

{ 6.6 REPORTING REQUIREMENTS (Continued) i Startup reports shall be submitted within (1) 90 days following j

completion of the startup test program, (2) 90 days following i resumption or commencement of commercial power o>eration, or (3) 9 months following initial criticality, whiciever is earliest.

If the startup report does not cover all three events (i.e.,

s

{

initial criticality, resumption or commencement completion of startup of commercial testoperation power program,)and

, supple-mentary reports shall be submitted at least every 3 months until all l three events have been completed.

l 2. Annual Report j A tabulation shall be subaitted on an annual basis prior to March 1 of each year of the nimber of station, utility, and other personnel (including contractors) receiving exposures greater than 100 aren/yr

.l andtheirassociatedaanreaexposureaccordin!1toworkandjob j functions (Note: this tabulation su i Section 20.407 of 10 CFR 20), e.g., pplements ".he requirements ofreactor l

lance,ibemaintenance)

(descr inservice inspection, routineThe wasteprocessing,andrefueling. maintenance dose special m assignments to various, duty functions may be estimated based on pocket dosimeter, TLD or film badge sensurements. Small exposures totaling less than 20k of the individual total dose need not be i accounted for. In the aggregate, at least 80% of the total whole j body dose received from external sources shall be assigned to specific 3 major work functions.

I The results of specific activity analysis in which the primary j coolant exceeded the limits of Specification 3.4.5 shall be included i in the Annual Report along with the following information: (1)Reac-I tor 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 tn

' whics the limit was exceeded; (2) Results of the last isotopic analy-sis for radiciodine performed prior to exceeding the limit results of analysis while limit was exceeded and results of one ana, lysis i after the radiciodine activity was reduced to less than limit. Each l 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 />

! priortothefIrs(tsampleinwhichthelimitwasexceeded;(4) Graph

! of the I-131 concentration and one other radiciodine isotope concen-l tration in microcuries per ran as a function of time for the dura-

tion of the specific activi (5).The time duration when above the activity specific steady-state of thelevel} and pr sary coolant exceeded the radiciodine limit.

i i

i i s..

! LA SALLE - UNIT 2 6-23 Amendmeni No. 69 1

l i

l

i- /$hM/M.57M71

{ fr2:- a-i m-CONTROLS 1

3. Annual Radiological Environmental Operatino Report"

] The Annual Radiological Environmental Operating Report covering the 4

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 outlined in (1) the 00CM and (2) Sections IV.B.2', IV.B.3, and IV.C of I

l d Appendix I to 10 CFR Part 50.

=

==i===n=F Radioactive Effluent Release Report ** & W % D l 4.

he Semiannual Radioactive Effluent Release Report covering the operation of the unit during the previous 6 months of operation shall n ow + tad wi + M= sn dava af+ -- Jann--o 1 =ad in1" ' af ---h u---

The report shall include ' asummary of th'e quantities of radioacitive liquid and gaseous affluents 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. Month 1y Operating- Report gs spec /kc4 /vGFgcL[

Routine reports of operating statistics and shutdown experience, including documentation of all challenges to safety / relief valves, shall be submitted on a monthly basf s to thefDirector, ofnce vi fNu' Clear Reactor RieguiaT,lon, Mall Station PI-137, US Nuclear Regulatory Commission, Washington, DC 20555, with a copy of_the appropriate LReatona10ffice. to Arriva/no later than the 13th oT each month follos ng the calendar month covered by the report.

A AnychangestotheOFFSITEDOSECALCULATIONMANUALshallbesubmitted])

with the Monthly Operating Report within 90__ days in which the chana (s as_magg, f 4 edd.Rion mFeeDorT z OF any maior Chance 5 t>

tfFe rac io,,e_fjq11==

active waste treatment systions shall pe submitted with t w Monthly Operating Report for the period in which the evaluation was A reviewed and accepted by Onsite Review and Investigative Function.

6. Core Operating Limits Report ..

a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

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

• • A single submittel 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 2 6-24 Amendment No. 69

i f

ATTACHMENT B PROPOSED AMENDMENTS TO THE l LICENSE / TECHNICAL SPECIFICATIONS 4

j Insert E i

j 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.

r 1

i a

e I

1 B-8

ADMINISTRATIVE CONTROLS -

~

b Core Doeratina Limits Report (Continued) j (1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1.

i (2) The minimum Critical Power Ratio l dependent MCPR limits, and power a(MCPR) scram time nd flow dependent MCPR  !

limits for Technical Specification 3.2.3. Effects of analyzed equipment out of service are included.

(3)

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

]

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

1

b. The analytical methods used to determine the core operating i

limits shall be those previously reviewed and approved by the NRC.

For LaSalle County Station Unit 2, the topical reports are: j (1) ANFB Critical Power Correlation, ANF-ll25 ) A) and  !

Supplements 1 and 2, Advanced Nuclear Fue orporation, j April 1990.

4 1

(2) Letter, Ashok forC. Referencin Thadani (NRC to R.A. Cgpeland (SPC),

" Acceptance  !

9x9-IX/X BWR FuelJuly Design,"g 28, 1993. of) ULTRAFLOW Spacer on 1

(3) Advanced Nuclear Fuels Corporation Critical Power Methodology for Boiling Water Reactors / Advanced Nuclear i 1 Fuels Corporation Critical Power Methodology for Boiling i j

Water Reactors: Methodology for Analysis of Assembly '

Channel Bowing Effects /NRC Correspondence, XN-NF-524(P)(A)

Revision 2 and Supplement 1 Revision 2, Supplement 2, 1

Advanced Nuclear Fuels Corporation November 1990.

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

1 (5) HUXY: A Generalized Multirod Heatup Code with 10 CFR 50, Appendix K Heatup Option, A), Supplement 1 Fuels ANF-CC-33(P)(Nuclear Revision 1; and Supplement 2, Advanced i

Corporation, August 1986 and January 1991, respectively.

i (6) Advanced Nuclear Fuel Methodology for Boiling Water Reactors XN-NF-80-19(P Supplemen,t 3 Appendix F),(A),' Volume 1,4.Supplement and Supplement Advanced Nuclear 3, i

Fuels Corporation, November 1990.

)

(7) Exxon Nuclear Methodology for Boiling Water Reactors:

i Application of the ENC Methodology to BWR Reloads, j Company, June)1986.XN-NF-80-19(P (A), Exxon Nuclear Volume 4, Revision 1 4

j (8)

Exxon Nuclear Methodology for Boiling Water Reactors THERMEX: Thermal Limits Methodology Summary Description, 3 XN-NF-80-19 P

Company, Jan(ua)r(y)1987.A , Volume 3, Revision 2, Exxon Nuclear i

1' LA SALLE UNIT 2 6-25 Amendment No.101 t

ADMINISTRATIVE CONTROLS Core Oneratino Limits Renort (Ccntinu2d) '

i j' (9) Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel, XN-NF-85-67 Company, September 1986.(P)(A) Revision 1, Exxon Nuclear 4

(10) Advanced Nuclear Fuels Corporation Generic Mechanical 1 Design for Advanced Nuclear Fuels Cor 9x9-9X BWR Reload Fuel, ANF-89-014 (A),1991.(P)poration Revision 1 and

! Supplements 1 and 2, October 9x9-J i (11) Volume 1 - STAIF - A Computer Pro Analysis in the Frequency Domain, gram for BWR Stability i Volume 2 - STAIF - A i Computer Program for BWR Stability Analysis in the Fre uenc

] INFO Ot4Q. 074fP)(Af, Domain,CodeQualificationReport SiemensPowerCorporation, July 1994.

EMF-CC-i j NO CHNMES (12) RODEX2 Fuel Rod Theriaal-Mechanical Response Evaluation j Model, XN-NF-81-58(P)(A), Revision 2 Supplements 1 and 2, 4

Exxon Nuclear Company, March 1984.

4 i

(13) XCOBRA-T: A Computer Code for BWR Trans'ient Thermal-HydraulicCoreAnalysis,XN-NF-84-105(P),(A)ienent4 Volume 1 and i Volume 1 Supplements 1 and 2 Volume I supp Advanced Nuclear Fuels Corpor;ation, February 1987 and June 1988, respectively.

i (14) Advanced Nuclear Fuels Corporation Methodology for Boiling )

Water Reactors EXEM BWR Evaluation Model Advanced Nuclear Fuels Corporation, Janua,ry 1993.ANF-91-048(P)(A), )

t (15) l Exxon Nuclear Methodology for Boiling Water Reactors -

- Neutronic Methods for Design and Analysis, XN-NF-80-19 P Nuclear Comp (an)y(A)Richland, WA 99352, March 1983. Volum (16) Exxon Nuclear Plant Transient Methodology for Boiling Water Reactors, XN-NF-79-71(P)(A), Revision 2 Supplements 1, 2, and 3, Exxon Nuclear Company, March 1986.

(17) Generic Mechanical Design Criteria for BWR Fuel Designs, ANF-89-98(P)fA), Revision 1andRevision1Su Advanced Nuc< ear Fuels Corporation, May 1995.pplement 1,

(

j

(18) NEDE-24011-P-A, " General Electric Standard Application for

! Reactor Fuel," (latest approved revision).

i (19) Commonwealth Edison To ical N 4 BWRNuclearDesignMet$ods,"($ortNFSR-0085,"Benchmarkof atest approved revision).

(20) Commonwealth Edison Topical Report NFSR-0085, Supplement 1, i

" Benchmark of BWR Nuclear Design Methods - Quad Cities i Gamma Scan Comparisons," (latest approved revision).

i i (21) Commonwealth Edison Topical Report NFSR-0085, Supplement 2, j " Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision).

(22) Cosmonwealth Edison To ical Re i CASM0

! Supple ments /MICR08 1 and 2 URN BWR Nu!1 Revision ign Methods," ear De!0,ort NFSR-0

1992, respectively;,SER letter dated MarchDecember 22, 1993. 1991, March 19 5

LA SALLE UNIT 2 6-25 a Amendment No.101 i

ADMINISTRATIVE CONTROLS C' ore Oneratina Limits Renort (Continued) -

c. The core operating limits shall be determined so that all a plicable limits  !

t ermal-hydraulic :C ., fuel thermal-mechanical limits, core  !

ts, ECCS limits, nuclear limits such as shutdown margin, and transient and accident analysis limits) of j the safety analysis are met.

1

d. The CORE OPERATING LINITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear R ulatory l Commission Document Control Desk with copies to the Reg onal Administrator and Resident Inspector. .

\

B. Deleted b M Mw4ishkrcf C. Unique Reporting Requirements s @ C T & G W be -

p I

1. Snecial Reports shall be ph=i++=d to the inecuor of_the_ Office o Enspection amt EnfarcenFeQegion IID wi(th1rFuhe, time perred -

specirieu ror each report. ,

6.7 PROCESS CONTROL PROGRAM (PCP)*

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

6.7.2 Licensee initiated changes to the PCP:

a. Shall be documented and records of reviews performed shall be retained l as required by Specification 6.5.8.18. This documentation shall contain:  !

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

2) A determination that the change will maintain the overall con-formance of the solidified waste product to existin  :

I of Federal, State, or other applicable regulations.g requirements  !

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 I and La Salle Unit 2.

l l

I l l

l l

l l ,

1

)

2

{ LA SALLE UNIT 2 6-26 Amendment No. 101

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6.8 0FFSITE DOSE CALCULATION MANUAL (00CM)*

6.8.1 The ODCM shall be approved by the Commission pr'or to implementation.

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 radi-oactive 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 4 not adversely impact the accuracy or reliability of effluent, I dose, or setpoint calculations.

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

c. Shall be submitted to the Commission in the form of a complete, leg-ible copy of the entire ODCM as a part of or concurrent with the

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hmiannup Radioactive Effluent Release Report for the period of the report Tn which any change to the 00CM was made effective. Each My/ change pages, clearly shallindicating be identified the area of the page bythatmarkings in the mar was changed, and shall indicate the date (e.g., month / year) the change was implemented.

6. 9 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6.9.1 Licensee initiated major changes to the radioactive waste treatment systems (liquid, gaseous, and solid):

a. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the Onsite Review and Investigative Function. Thedhcussionofeachchange shall contain:

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

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• The OFFSITE DOSE CALCULATION MANUAL (00CM) is common to La Salle Unit 1 and La Salle Unit 2.

LA SALLE - UNIT 2 6-27 Amendment No. 69 l.____--_. -

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? C :s;3G;~;^/ CONTROLS MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS (Continued)

4. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously I

predicted in the license application and amendments thereto; l

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; I l

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

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i l LA SALLE - UNIT 2 6-28 Amendment No. 69 l

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l ATTACHMENT C j SIGNIFICANT HAZARDS COSIDERATION l

} Summary of the Proposed License and Technical Specification Chanaes:

The requested amendment proposes to revise LaSalle County Nuclear Station (LaSalle) Unit 1 License Condition 2.C.(30)(a)., and Unit 1 and 2 Technical 3

Specifications, Section 5, Design Features, and Section 6, Administrative Controls. The proposed change includes (1) Incorporat:ig organizational titles, (2) changing the submittal frequency of the Radiological Effluent Release Report, and (3) other administrative changes. In summary, these changes j include: '

1. Revised Organizational Titles LaSalle implemented revised station organizational titles.

The requirement that the Shift Technical Advisor (STA) function be fulfilled by the Station Control Room Engineer is being revised. The STA function of an on-  !

shift technical advisor to the Shift Manager will be fulfilled by quallfled l Individuals consistent with the requirements of NUREG-0660, Action item I.A.1.1,  !

as clarified by NUREG-0737. I

2. Change submittal frequency of the Radiological Effluent Release Report The Radiological Effluent Release Report is currently submitted semiannually per Definition 1.27, and Specifications 6.6.A.4 and 6.8.2.c. This reflects the requirements of 10CFR50.36a which were in effect prior to August 31,1992.

10CFR50.36a was amended in August 1992 and allows for annual report submittal. This change achieves consistency between the Technical Specifications and the amended requirement of 10CFR50.36a.

3. Other Administrative Changes Other changes are proposed to clarify other organizational title changes, expand the qualification for technical reviewers and revise submittal addresses for certain reports. With respect to technical reviewers, the revision will permit individuals qualified to ANSI N18.1-1971 Sections 4.3 (Supervisors),4.5.1 (Operators), or 4.6 (Technical Support Personnel) to perform technical reviews of procedures identified in Specification 6.2.C. This provides for other appropriate expertise to be utilized in the technical review process.

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ATTACHMENT C SIGNIFICANT HAZARDS COSIDERATION Comed has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10CFR50.92(c), a proposed amendment to an operating license involves no significant hazards if operation of the facility in accordance with the proposed amendment would not:

A. Yhe proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes do not affect any accident Initiators or precursors and do not change or alter the design assumptions for systems or

components used to mitigate the consequences of an accident. The l proposed changes do not affect the design or operation of any system, structure, or component in the plant. There are no changes to parameters governing plant operation, and, no new or different type of equipment will l be Installed.

l The proposed changes provide clarification, consistency with station l procedures, programs, the Code of Federal Regulations (10CFR), other l Technical Specifications, and Improved Technical Specifications. These changes do not impact any accident previously evaluated in the UFSAR.

There is no relaxation of applicable administrative controls. Those administrative requirements which have no effect on safe operation of the l plant are eliminated.

B. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not affect the design or operation of any plant system, structure, or component. There are no changes to parameters governing plant operation, and, no new or different type of equipment will l

be Installed. The organizational and administrative changes proposed have no effect on the design or operation of any system, structure, or component

in the plant. There are no changes to parameters governing plant operation; no new or different type of equipment will be installed.

C. The proposed changes do not involve a significant reduction in a margin of safety.

The proposed changes do not affect the margin of safety for any Technical Specification. The initial conditions and methodologies used in the

! accident analyses remain unchanged; therefore, accident analyses results I

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I ATTACHMENT C SIGNIFICANT HAZARDS COSIDERATION are not impacted. Plant safety parameters or setpoints are not affected. All responsibilities described in the Technical Specifications for administrative controls will continue to be performed by Individuals possessing the I requisite qualifications. Clarifications, relecations, and nomenclature changes neither result in a reduction of personnel responsibilities, nor do they cause a relaxation of programmatic controls. There are no resulting i effects on plant safety parameters or setpoints.

i Guidance has been provided in " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the I application of standards to license change requests for determination of the  ;

existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve i significant hazards considerations. These proposed amendments most closely fit the example of a purely administrative change to the Technical specifications to achieve consistency throughout the Technical specifications, correction of an error, or a change in nomenclature.

The proposed amendment does not involve a significant relaxation of the criteria used to establish safety limits, a significant relaxation of the bases for the limiting safety system settings, or a significant relaxation of the bases for the limiting conditions for operations. The proposed change does not reduce the margin of safety as defined in the basis for any Technical Specification.

Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), Comed has concluded that the proposed change does not involve significant hazards considerations.

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ATTACHMENT D ENVIRONMENTAL ASSESSMENT STATEMENT APPLICABILITY REVIEW Comed has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR Part 51.21. It has been determined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFR Part 51.22(c)(9). This conclusion has been determined because the changes requested do not pose significant hazards considerations or do not involve a significant increase in the amounts, and no significant changes in the I types of any effluents that may be released off-site. Additionally, this request does not involve a significant increase in individual or cumulative occupational radiation exposure.

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ATTACHMENT C SIGNIFICANT HAZARDS COSIDERATION Summary of the Proposed License and Technical Soecification Channes:

The requested amendment proposes to revise LaSalle County Nuclear Station i

(LaSalle) Unit 1 License Cor.dition 2.C.(30)(a)., and Unit 1 and 2 Technical l Specifications, Section 5, Design Features, and Section 6, Administrative I

Controls. The proposed change includes (1) incorporating organizational titles, (2) changing the submittal frequency of the Radiological Effluent Release l Report, and (3) other administrative changes. In summary, these changes include:

1. Revised Organizational Titles LaSalle implemented revised station organizational titles.

The requirement that the Shift Technical Advisor (STA) function be fulfilled by the Station Control Room Engineer is being revised. The STA function of an on-l shift technical advisor to the Shift Manager will be fulfilled by quallfled

! Individuals consistent with the requirements of NUREG-0660, Action item I.A.1.1, ,

i as clarified by NUREG-0737.

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l 2. Change submittal frequency of the Radiological Effluent Release Report

! ae Radiological Effluent Release Report is currently submitted semiannually

! per Definition 1.27, and Specifications 6.6.A.4 and 6.8.2.c. This reflects the l requirements of 10CFR50.36a which were in effect prior to August 31,1992.

l 10CFR50.36a was amended in August 1992 and allows for annual report l submittal. This change achieves consistency between the Technical l Specifications and tlic amended requirement of 10CFR50.36a.

3. Other Administrative Changes Other changes are proposed to clarify other organizational title changes, expand j the qualification for technical reviewers and revise submittal addresses for certain reports. With respect to technical reviewers, the revision will permit individuals qualified to ANSI N18.1-1971 Sections 4.3 (Supervisors),4.5.1 (Operators), or 4.6 (Technical Support Personnel) to perform technical reviews of procedures identified in Specification 6.2.C. This provides for other appropriate expertise to be utilized in the technical review process.

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ATTACHMENT C SIGNIFICANT HAZARDS COSIDERATION l Comed has evaluated this proposed amendment and determined that it involves no significant hazards considerations. According to 10CFR50.92(c), a proposed amendment to an operating license involves no significant hazards if operation I

of the facility in accordance with the proposed amendment would not:

A. The proposed changes do not !nvolve a significant increase in the l probability or consequences of an accident previously evaluated.

l The proposed changes do not affect any accident initiators or precursors i and do not change or alter the design assumptions for systems or l components used to mitigate the consequences of an accident. The proposed changes do not affect the design or operation of any system, structure, or component in the plant. There are no changes to parameters governing plant operation, and, no new or different type of equipment will be installed. i The proposed changes provide clarification, consistency with station l procedures, programs, the Code of Federal Regulations (10CFR), other  !

Technical Specifications, and improved Technical Specifications. These changes do not impact any accident previously evaluated in the UFSAR.

There is no relaxation of applicable administrative controls. Those administrative requirements which have no effect on safe operation of the plant are eliminated.

B. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes do not affect the design or operation of any plant system, structure, or component. There are no changes to parameters governing plant operation, and, no new or different type of equipment will be installed. The organizational and administrative changes proposed have no effect on the design or operation of any system, structure, or component in the plant. There are no changes to parameters governing plant operation; no new or different type of equipment will be installed.

C. The proposed changes do not involve a significant reduction in a margin of safety.

The proposed changes do not affect the margin of safety for any Technical Specification. The initial conditions and methodologies used in the accident analyses remain unchanged; therefore, accident analyses results C-2

ATTACHMENT C SIGNIFICANT HAZARDS COSIDERATION are not impacted. Plant safety parameters or setpoints are not affected. All responsibilities described in the Technical Specifications for administrative controls will continue to be performed by Individuals possessing the requisite qualifications. Clarifications, relocations, and nomenclature changes neither result in a reduction of personnel responsibilities, nor do they cause a relaxation of programmatic controls. There are no resulting effects on plant safety parameters or setpoints.

Guidance has been provided ir: " Final Procedures and Standards on No Significant Hazards Considerations," Final Rule,51 FR 7744, for the application of standards to license change requests for determination of the existence of significant hazards considerations. This document provides examples of amendments which are and are not considered likely to involve significant hazards considerations. These proposed amendments most closely fit the example of a purely administrative change to the Technical specifications to achieve consistency throughout the Technical specifications, correction of an error, or a change in nomenclature.

The proposed amendment does not involve a significant relaxation of the criteria used to establish safety ilmits, a significant relaxation of the bases for the limiting safety system settings, or a significant relaxation of the bases for the limiting conditions for operations. The proposed change does not reduce the margin of safety as defined in the basis for any Technical Specificatlon.

Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), Comed has concluded that the proposed change does not involve significant hazards considerations.

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i ATTACHMENT D j ENVIRONMENTAL ASSESSMENT STATEMENT j APPLICABILITY REVIEW i

$ Comed has evaluated the proposed amendment against the criteria for identification of licensing and regulatory actions requiring environmental

assessment in accordance with 10 CFR Part 51.21. It has been determined that

, the proposed changes meet the criteria for categorical exclusion as provided for i under 10 CFR Part 51.22(c)(9). This conclusion has been determined because the changes requested do not pose significant hazards considerations or do not 4 involve a significant increase in the amounts, and no significant changes in the i types of any effluents that may be released off-site. Additionally, this request l does not involve a significant increase in individual or cumulative occupational radiation exposure.

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