ML20135A509

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Proposed Tech Specs Implementing Mods Proposed by Suppl to License Amend Request & Revising TS Pp in Final Form to Reflect Changes Proposed by Suppl
ML20135A509
Person / Time
Site: Prairie Island  Xcel Energy icon.png
Issue date: 11/25/1996
From:
NORTHERN STATES POWER CO.
To:
Shared Package
ML20135A502 List:
References
NUDOCS 9612030269
Download: ML20135A509 (31)
v • d • e


Text

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! ATTACHMENT 1 l

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l SUPPLEMENT -

to LICENSE AMENDMENT REQUEST DATED December 14,1995 i Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Soecifications Appendix A, Technical Specification Pages  !

from LAR Dated December 14,1995, Exhibit C Marked Up Pages as proposed in this Supplement i

TS-v i TS-ix TS-x l TS-xi l TS-xii TS.6.0-2 TS.6.0-3 TS.6.0-4 T.S.6.0-7 T.S.6.0-8 TS.6.0-9 TS.6.0-11 TS.6.0-14 .

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9612030269 961125 PDR ADOCK 05000282 p PDR l

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, TS-v TABLE OF CONTENTS (Continued)

T9 SECTION TITLE PAGE l 4.0 SURVEILLANCE REQUIREMENTS TS.4.0-1 l 4.1 Operational Safety Review TS.4.1-1  !

4.2 Inservice Inspection and Testing of Pumps and Valves Requirements TS.4.2-1 A. Inspection Requirements TS.4.2-1 l B. Corrective Measures TS.4.2-2 C. Records TS.4.2-3 4.3 Primary Coolant System Pressure Isolation Valves TS.4.3-1 4.4 Containment System Tests TS.4.4 1 A. Containment Leakage Tests TS.4.4-1 B. Emergency Charcoal Filter Systems TS.4.4 3 ,

C. Containment Vacuum Breakers TS.4.4-4 D. Deleted E. Containment Isolation Valves TS.4.4-5 >

F. Post Accident Containment Ventilation System TS.4.4-5  !

G. Containment and Shield Building Air Temperature TS.4.4-5 H. Containment Shell Temperature TS.4.4-5 I. Electric Hydrogen Recombiners TS.4.4-5 l 4.5 Engineered Safety Features TS.4.5-1 l A. System Tests TS.4.5-1 "

l 1. Safety Injection System TS.4.5-1

2. Containment Spray System TS.4.5-1  ;

i 3. Containment Fan Coolers TS.4.5-2 l l

4. Component Cooling Water System TS.4.5-2 l S. Cooling Water System TS.4.5-2 l B. Component Tests TS.4.5-3  !
1. Pumps TS.4.5-3 1
2. Containment Fan Motors TS.4.5-3 l
3. Valves TS.4.5-3 I 4.6 Periodic Testing of Emergency Power System TS.4.6-1 A. Diesel Cenerators TS.4.6-1 B. Station Batteries TS.4.6-3 C. Pressurizer Heater Emergency Power Supply TS.4.6-3 i 4.7 Main Steam Isolation valves TS.4.7-1 l

4.8 Steam and Power Conversion Systems TS.4.8-1 A. Auxiliary Feedwater System TS.4.8-1 B. Steam Cenerator Power Operated Relief Valves TS.4.8-2 C. Steam Exclusion System TS.4.8-2 l 4.9 Reactivity Anomalies TS.4.9-1 l 4.10 DilsEsdP A'-**-- "- *------* ' "- '*--*- "------ 'c ^'n' E'~53231: Cell:: tier 2nd t.n:Iyric TS.^.10-1 S. Land U Cencue TS.^.10-2 C. Ir.tcrl t-Cr2!Ory Corp 2r'20 . "r ;r- TS . ^ .10 -- 2 ,

4.11 DyljjjpRedicenti: Scure: Lech:g: Tt TS.^.11-1 i

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4 4 , TS-ix 1

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TABLE OF CONTENTS (continued)

I TS BASES SECTION TITLE PAGE '

2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit . "creter Cere B.2.1-1 l KW Rss6tbMC(o(F{#EstfriiiiitFE$n,$*,!MiBESgMfF#

MR]eactj{[_Co6A%]ilysdiff{5N d8,)jjjj[ikphj[gd$$dyfjd l 2.2 Safety Limitdiggg. "creter Cerlant Sycte: Preccur B.2.2-1 2.3 Limiting Safety System Settings, Protective B.7. 3-1 l Instrumentation 3.0 BASES FOR' LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 l 3.1 Reactor Coolant System B.3.1-1 -

A. Operational Components B.3.1-1 l B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6

)

D. Maximum Coolant Activity B.3.1-7 E. Deleted F. Isothermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 1 3.4 Steam and Power conversion Systems B.3.4-1 ]

3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 L 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Dsistid"rdieretive Effluent: 3.3.9 1 E Y$3Eid Effluent: ' E.3.9-1 E. Crrreur Effluent: E.3.9 2 C. Solid Pedirective " :t: 3.3.9 ^

D. D::: Fre: All "renium Fuel Cycle Seur :: E.3.9 5 E. E F Effluent Monitering Inctru er.t tier " 3.9 5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits- B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 )

G. Control Rod Operability Limitations B.3.10-9 i H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10

l. 3.11 Core Surveillance Instrumentation B.3.11-1 l 3.12 Snubbers B.3.12-1 l 3.13 Control Room Air Treatment System B.3.13-1 l 3.14 Deleted
3.15 Event Monitoring Instrumentation B.3.15-1 1

I t

TS-x >

TABLE OF CONTENTS (continued)

TS BASES SECTION TITLE PAGE 4.0 BASES FOR SURVEILLANCE REQUIREMENTS i i

4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 r and Valves Requirements 4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves ,

4.4 Containment System Tests B.4.4-1 4.5 Engineered Safety Features B.4.5-1 4.6 Periodic Testing of Emergency Power Systems B.4.6-1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Dsl~ifidRa4&atten Envirement:1 .Menitering Pre-rcr E.^.10-1 I"~5IEhle C:llectier D T ,. n A Tf e m Fmm-..n nd ^.n:17:1 E.^.10 1 D L 1 A _1 C. Inter 1:baratery Cerperiter Pregrz E.^.10 1 j 4,11 D4}spidRedic tive Seur : Lecheg Test E.^.11-1 l 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 DsisiiBFire Detectier end Pret:: tier Systerr B.^.16-1 4.17 d1M5Redir::tiveEffluent: Surveillene 3.^.17 1 4.18 Reactor Coolant Vent System Paths B.4.18-1 1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1 l l

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. . TS-xi i l

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]

i TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE i

l 1-1 Operational Modes .

l 3.5-1 Engineered Safety Features Initiation Instrument Limiting _

Set Points 3.5-2A Reactor Trip System Instrumentation 3.5-2B Engineered Safety Feature Actuation System Instrumentation '

3.9-1 ": directive Liquid Effin nt ":niterin; Instrurentatien 3.9-2 " dicertiv: 02 ceu Effluent "enit: ring instrurent tier +

3.15-1 Event Monitoring instrumentation Precer: E Centninrrnt

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4.1-1A Reactor Trip System Instrumentation Surveillance Requirements 4.1-1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-1C Hiscellaneous Instrumentation Surveillance Requirements 4.1 2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements .

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^ .10-3 RF"? "ep : tin; Level: fer "edi ::tivity Concentretien: ir r_...s.__..,. . - . . . - . . . . e-__,.

4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval

^ l'-1 Radicactiv: Liquid Effluent "enitering Instrunentatier Surveillen:: "cquirrrente

^ .17 2 Redieretive C :::ur Effluent ":nitoring in trunent tier Surveillene: Requiren:nte

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.17-3 Redienctive Liquid " :t Senpling and ^.n 1yri Pregren

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Liquid Effluent: Fren Prcirie Irl nd "ucle:r Cener: ting j n1.. .... I n.g .m .. M....f

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TS-xii APPENDIX A TECHNICAL SPECIFICATIONS l LIST OF FIGURES i

TS FIGURE TITLE ,

2.1-1 [:LisEt~ppjC6Fi*JsNty7LIRit~fS fety Linite, P.::cter Cere, Sern:1  !

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3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations  !

3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations '

3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit  ;

Versus Percent of RATED THERMAL POWER with the Primary Coolant l Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131  !

3.8 1 Spent Fuel Pool Unrestricted Region Minimum Burnup Requirements 3.9-1 Prairi I: lend Mu:1cer Cencreting Plant Site Ecundary fer Liquid  !

Effluent:

3.9-2 Prairi: I:1:nd Nucle r Centreting P1:nt Site Ecundary fer C :ccur Effluent 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration I 4.4 1 Shield Building Design In-Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Pool Checkerboard Region Minimum Burnup Requirements ,

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l c , TS.6,0-2 6.2 Organization A. Onsite and Offsite Organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

1. Lines of authority, responsibility and communication shall be defined and established throughout highest management levels, intermediate levels, and all operating organization positions.

These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Updated Safety Analysis Report.

2. The plant manager shall report to the corporate vice president specified in 6.2.A.3, shall be responsible for overall safe operation of the plant, and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.
3. A corporate vice president shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety.
4. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

B. Plant Staff The plant staff organization shall include the following:

1. An operator to perform non-licensed duties shall be assigned to each reactor containing fuel and one additional operator to l perform non-licensed duties shall be assigned when either or l both reactors are operating in MODES 1, 2, 3, or 4. ^1:0, if l cn unit ic ir MODE 1, 2, 3, er ^ cnd the ether unit ic ir MODE 5 er 5, cc : minitur the cr-cite ct ffing chnll include tre cenier raccter eperaterc(SRO) :nd tre licenced reccter cper terc(RO).
2. At least one licensed operator shall be present in the control room for each reactor containing fuel. In addition, while l either unit is in MODE 1, 2, 3, or 4, at least one licensed senior reactor operator shall be present in the control room.

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TS.6.0-3

3. Shift crew composition may be less than the minimum requirement of 10CFR50.54(m)(2)(1) and 6.2.B.1 and 6.2.B.7 for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the l minimum requirements. j
           . 4. An individual qualified in radiation protection procedures                                                                                                                        I shall be on site when fuel is in a reactor. The position may be                                                                                                                   i vacant for not more than 2 hours, in order to provide for                                                                                                                          l unexpected absence, provided immediate action is taken to fill the required position.
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6. The operations manager or assistant operations manager shall '

hold an SRO license. P i 7. The shift technical advisor (STA) shall-provide advisory l technical support to the shift supervisor in the areas of l thermal hydraulics, reactor engineering, and plant analysis ! with regard to the safe operation of the unit. Personnel l performing the function of the STA shall be assigned to the l shift crew when a unit is in MODE 1, 2, 3, or 4. l l i I I e

a ' TS.6.0-4 6.3 Plant Staff Qualifications Each member of the plant' staff shall meet or exceed the minimum-qualifications of Regulatory Guide 1.8, Revision 1, September 1975 except for (1) personnel who perform the function of shift technical-advisor shall gjg8203E6ay]Ejijiid have a bachelors degree or - equivalent in a scientific or engineering discipline.with specific training in plant design, and response and analysis of the plant for transients and accidents, and (2) the operations manager who shall meet the requirements of ANSI _N18.1-1971,'except that NRC license requirements are as specified in Specification 6.2.B 6. i l L l l l 1 i I i 1 l 1 e -

s TS.6,0 7 C. Post Accident Samplinz This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

1. Training of personnel;
2. Procedures for sampling and analysis; and
3. Provisions for maintenance of sampling and analysis equipment.

D. Radioactive Effluent Controls Program This program conforms to 10CFR50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. This program shall allocate releases equally to each unit. The liquid radwaste treatment system, waste gas treatment system, conta nment purge release vent, and spent fuel pool vent are shared by both units. Experience has also shown that contributions from both units are released from each auxiliary building vent. Therefore, all releases will be allocated equally in determining conformance to the design objectives of 10CFR50, Appendix I. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM; )
2. Limitation on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to Appendix B to 10CFR20.1 - 20.601, Table II, Column 2;
3. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10CFR20.1302 and with the l methodology and parameters in the ODCM; I l
4. 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, conforming to 10CFR50, Appendix I; i
5. Determination of cumulative and prcj ected dose contributions '

from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCMKDit'sidisihiti5sI(6fIpf6fiEtidEddss 66htP'.intfises7f6HYsdissetiseViffluintsti ! ace 6fdsridsnitMthe Nethyy y s f ij M h M M E M E M [ ~ ~ ~ ~ ~ ~ 

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  • TS.6,0-8
6. Limitations on the functional capability and use of the liquid and gaseous. effluent treatment systems to ensure that appropriateportions of these systems are used to reduce releases of radioactivity when the projected doses in a period of one month from the liquid effluent releases would exceed 0.12 arem to the total body or 0.4 mrem to any organ; or from the gaseous effluent releases would exceed 0.4 mrad for gamma air dose, 0.8 mrad for beta air dose, or 0.6 arem organ dose;
                .7. Limitations on the dose rate resulting from radioactive material released in gaseous affluents to areas beyond the site boundary conforming to the dose associated with Appendix B to 10CFR20.1 - 20.601, Table II, Column 1;
8. 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 10CFR50, Appendix I;
9. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix

! I; and

10. Limitation on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40CFR190.

E. Component Cvelic or Transient Limit This program provides controls to track the USAR, Section 4.1.4 cyclic and transient occurrences to ensure that components are maintained within the design limits. F. (Reserved) C. (Reserved) i ! H. (Reserved) l-i- I. (Reserved) 4 4 4 4

, l 9 TS.6.0-9  : J. Explosive Gas and Storare Tank Radioactivity Monitorine Prorram This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system, the quantity of  ! radioactivity contained in gas storage tanks, and the quantity of  ; radioactivity contained in unprotected outdoor liquid storage tanks. l The program shall include: c i

1. The limits for concentration of oxygen in the waste gas holdup i system and a surveillance program to ensure the limits are  ;

maintained. Such limits shall be appropriate to the system's design criteria;  !

2. A surveillance program to ensure that the quantity of  :

radioactivity contained in each gas storage tank is less than or equal to 28]p00!?S,000 curies of noble gases (considered as dose equivalent Xe-133); and

3. A surveillance program to ensure that the quantity of
  • radioactivity contained in each of the following tanks shall be limited to 10 curies, excluding tritium and dissolved or entrained noble gases: ,

Condensate storage tanks Outside temporary tanks

4. The provisions of TS 4.0 are applicable to the Explosivo Gas and Storage Tank Radioactivity Monitoring Program surveillance i frequencies.

K. Diesel Puel Oil Testine Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. L. Technical Specifications Bases Control Procram This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases or the Technical Specifications shall be l

made under appropriate administrative controls and reviews. f

2. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
a. a change in the Technical Specifications incorporated in the license; or
b. a change to the USAR or Bases that involves an unreviewed safety question as defined in 10CFR50.59.

i 

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i o o TS.6.0-11 6.6 Reporting Requirements The following reports shall be submitted in accordance with 10CFR50.4 A. Occupational Exposure Report A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g. , reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. This tabulation supplements the requirements of 10CFR20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be t accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. This report shall be submitted by April 30 of each year. B. Annual Radiological Environmental Monitorine Report The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses 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 the Offsite Dose Calculation Manual (ODCM), and in 10CFR50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Cuide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. l The missing data shall ' e submitted as soon as possible in a supplementary report. The reports shall also include the following; a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from phesne reactor ((iyy; and the results of licensees participation ! in the Interlaboratory Comparison Program defined in the ODCM. l C. Radioactive Effluent Report The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR50.36a and 10CFR50, Appendix I, Section IV.B.1. 

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6.7 High Radiation Area- ( i A. Pursuant to 10CFR20, paragraph 20.1601(c), in lieu of the requirements of 10CFR20.1601, each high radiation area, as defined in 10CFR20, in which the intensity of radiation is greater than 100'.  ! arem/hr but less' than [${gifgg]1000 mres/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  ; Fermit-(RWP). Individuals qualified in radiation protection j procedures (e.g., health physics technicians) or personnel i continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates less than or equal to j 

                                             .1000 mrea/hr, provided they are otherwise following plant radiation                   j protection procedures for entry into'such high radiation areas.                       !

Any individual or group of individuals permitted to enter such areas ] shall be provided with or accompanied by one or more of the i following: l i

1. A radiation monitoring device that continuously. indicates the j radiation dose rate in the area. ,

1

2. A radiation monitoring device that continuously integrates the l radiation dose rate in the area and alarms when a preset l l integrated dose is received. Entry into such areas with this j l monitoring device may be made after the dose rate levels.in the i area have been established and personnel are aware of them. l l
3. An individual qualified'in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the radiation protection manager. ,

j B. In addition to the requirements of Specification 6.7.A above, areas t with radiation levels greater than -- --" ' -- 1000 mrem /hr shall be i provided with locked or continuously guarded doors to prevent . unauthorized entry and the keys shall be maintained t;nder the  ! administrative control of the Shift Supervisar on duty or health i physics supervision. Doors shall remain locked except during periods l; of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV or transmitting radiation monitoring device) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. j t 4 e - _ __ , , _ . - - - ,- r - - ,-  %-. - - - -

i ATTACHMENT 2 , l SUPPLEMENT l to l l LICENSE AMENDMENT REQUEST DATED December 14,1996 , Conformance of Administrative Controls Section 6 l to the Guidance of Standard Technical Specifications l l Appendix A, Technical Specification Pages Revised Pages as proposed in this Supplement l l ! TS-v l TS-ix l TS-x TS-xi TS-xii ' TS.6,0-2 i TS.6.0-3 > TS.6.0-4 T.S.S.0-7 T.S.6.0-8 TS.6.0-9 TS.6.0-11 TS.6.0-14 l l l 4 i I i

__..,__,. . _ . _ _ _ . _ _ . _ . _ _ . - _-_.-___._-___.m._ _ . . . 4. 

  *               ,                                                                                                     TS-v I&&}.E OF CONTENTS (Continued)

TS'SECTION TITLE PACE , 4,0 ' SURVEII.IANCE REQUIREMENTS TS.4.0-1 4.1 -Operational Safety Review TS.4.1-1 . 

                                             .4.2    Inservice Inspection and Testing of Pumps and Valves Requirements                                                   TS.4.2-1                ,

A. Inspection Requirements TS.4.2-1 i B. Corrective Measures TS.4.2-2 l C.-Records TS.4.2-3 l 4.3 Primary Coolant System Pressure Isolation Valves TS.4.3 1 4.4 Containment System Tests TS.4.4-1 A. Containment Leakage Tests 'TS.4.4-1 i B. Emergency Charcoal Filter Systems TS.4.4-3  : C. Containment Vacuum Breakers TS.4.4-4 i D. Deleted l l E. Containment Isolation Valves TS.4.4-5l  ! 

                                                    'F. Post Accident. Containment Ventilation System                      TS.4.4-5               l G. Containment and Shield Building Air Temperature                    TS.4.4-5               i H. Containment Shell Temperature-                                     TS.4.4-5 l                                                     I. Electric Hydrogen Recombiners                                      TS.4.4-5 4.5   Engineered Safety Features                                            TS.4.5-1 A. System Tests                                                       TS.4.5-1                {

l 1. Safety Injection System TS.4.5-1  ; l 2. Containment Spray System TS.4.5-1 j ! 3. Containment Fan Coolers TS.4.5-2 .i L 4. Component Cooling Water System TS.4.5-2 l

5. Cooling Water System TS.4.5-2  :

L B. Component Tests TS.4.5-3 l

1. Pumps TS.4.5-3  !
2. Containment Fan Motors TS.4.5-3 i 3. Valves TS.4.5-3 l 4.6 . Periodic Testing of Emergency Power System TS.4.6  ;

A. Diesel Generators TS.4.6-1 l B. Station Batteries TS.4.6-3 i C. Pressurizer Heater Emergency Power Supply TS.4.6-3  ; 4.7 Main Steam Isolation Valves TS.4.7-1 l 4.8 Steam and Power Conversion Systems TS.4.8-1 -! A. Auxiliary Feedwater System TS.4.8-1 l B. Steam Cenerator Power Operated Relief Valves TS.4.8-2 C. Steam Exclusion System TS.4.8-2 4.9 Reactivity Anomalies TS.4.9-1 4.10 Deleted  ! 4.11 Deleted l-l l i ' i 4 3 i

TS-ix TABLE OF CONTENTS (continued) TS BASES SECTION TITLE PAGE 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limits 3.2.1-1 A. Reactor Core Safety Limits B.2.1-1 B. Reactor Coolant System Pressure Safety Limits B.2.1-5 2.2 Safety Limit Violations, Reactor Coolant System Pressure B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Deleted F. Isothermal Temperature Coefficient (ITC) B.3.1-9 i 3.2 Chemical and Volume control System B.3.2-1 j 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4-1 3.5 Instrumentation System B.3.5-1 , 3.6 Containment System B.3.6-1 l 3.7 Auxiliary Electrical System B.3.7-1 1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Deleted 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 C. Contro~ Rod Operability Limitations B.3.10-9 l H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 l 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1 

 +

TS-x TABLE OF CONTENTS (continued) TS BASES SECTION TITLE PAGE i 4.0 BASES FOR SURVEILLANCE REQUIREMENTS  ; 4.1 Operational Safety Review B.4.1-1  ; 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 ' and Valves Requirements  ; 4.3 Primary Coolant System Pressure Isolation B.4.3-1 l Valves ) 4.4 Containment System Tests B.4.4-1 4.5 Engineered Safety Features B.4.5-1 1 4.6 Periodic Testing of Emergency Power Systems B.4.6 1 1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Deleted 4.11 Deleted 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 Deleted 4.17 Deleted 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1 l

, 1 

     +                                                                                      .

TS-xi  ! l 

                                                                                          .l TECHNICAL SPECIFICATIONS i

F LIST OF TABLES  : l TS TABLE TITLE j i 1-1 Operational Modes , i 3.5-1 ! Engineered Safety Features Initiation Instrument Limiting Set Points , 3.5-2A Reactor Trip System Instrumentation l 3.5-2B Engineered Safety Feature Actuation System Instrumentation P 3.15-1 Event Monitoring instrumentation  ! i 4.1-1A Reactor Trip System Instrumentation Surveillance Requirements - 4.1-1B Engineered Safety Feature Actuation System Instrumentation , Surveillance Requirements  ! 4.1-1C Miscellaneous Instrumentation Surveillance Requirements i 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests I 4.2 1 Special Inservice Inspection Requirements , 4.12-1 Steam Generator Tube Inspection v 4.13-1 Snubber Visual Inspection Interval-I t i i i i l I I 

  .     . _ . _ _ _ _ . . _ . _ - ~ _ . _ _ _ . . . . _ _ . _ . . _ . _ . . . _ _ . _ _ . . _ . _ _ . . _ _ _ . . . . . . _ . . _ . .

I 6 , TS-xii j 1 APPENDIX A TECHNICAL SPECIFICATIONS f LIST OF FICURES  ! i I TS FIGURE ' TITLE l 2.1-1 Reactor Core Safety Limits l [ I ! 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations ' 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT'I-131 Primary Coolant Specific Activity Limit  ; Versus Percent of RATED THERMAL POWER with the Primary Coolant  ; Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 l l 3.8-1 Spent Fuel Pool Unrestricted Region Minimum Burnup Requirements l 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration l l 4.4-1 Shield Building Design In-Leakage Rate

l. 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout i
l. 5.6-2 Spent Fuel Pool Checkerboard Region Minimum Burnup Requirements  !

B.2.1-1 Origin of Safety Limit Curves at 2235 psig with delta-T and Locus of Reactor Conditions at which SG Safety Valves Open f i 

                                                                                                                                                     )

l l l t I i l l l 

               .  -.             -     -      -.   --     --- -               - ~ _ _    .- - . _ .
  *       ,                                                                 TS.6,0-2 l

6.2 Organization A. Onsite and Offsite Organizations l- Onsite and offsite organizations shall be established for plant l operation and corporate management, respectively. The onsite and ' offsite organizations shall include the positions for activities l affecting safety of the nuclear power plant. 1 l l 1. Lines of authority, responsibility and communication shall be j defined and established throughout highest management levels, intermediate levels, and all operating organization positions. ' These relationships shall be documented and updated, as l appropriate, in organization charts, functional descriptions of i departmental responsibilities and relationships, and job j descriptions for key personnel positions, or in equivalent j forms of documentation. These requirements, including the plant i specific titles of those personnel fulfilling the . i responsibilities of the positions delineated in these Technical . Specifications, shall be documented in the Updated Safety l Analysis Report. l l

2. The plant manager shall report to the corporate vice president specified in 6.2.A.3, shall be responsible for overall safe operation of the plant, and shall have control over those  ;

onsite activities necessary for safe operation.and maintenance ' of the plant.

3. A corporate vice president 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, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

B. Plant Staff The plant staff organization shall include the following:

1. An operator to perform non-licensed duties shall be assigned to each reactor containing fuel and one additional operator to perform non-licensed duties shall be assigned when either or both reactors are operating in MODES 1, 2, 3, or 4.
2. At least one licensed operator shall be present in the control room for each reactor containing fuel. In addition, while
either unit is in MODE 1, 2, 3, or 4, at least one licensed l senior reactor operator shall be present in the control room.

_. .. - . . ~ . _ _ .- . -._ .. . - - . . - - - . - - -. l

  • l t

b TS.6,0-3

  • l 3. Shift crew composition may be less than the minimum requirement  !

of 10CFR50.54(m)(2)(1) and 6.2.B.1 and 6.2.B.7 for a period of time not to exceed 2 hours in order to accommodate unexpected  : absence of on-duty shift crew members provided immediate action j is taken to restore the shift crew composition to within the  ; minimum requirements.

  • i l 4. An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.  ;

l 5. Administrative procedures shall be developed and implemented to [ l limit the working hours of plant staff who perform safety > l related function (e.g., licensed SR0s, licensed Ros, health physicists, auxiliary operators, and key maintenance personnel). j controls shall be included in the procedures such that i l individual overtime shall be reviewed monthly ensure that , excessive hours have not been assigned. ' L I l 6. The operations manager or assistant operations manager shall hold an SRO license. l 7. The shift technical advisor (STA) shall provide advisory l l technical support to the shift supervisor in the areas of l thermal hydraulics, reactor engineering, and plant analysis  ! with regard to the safe operation of the unit. Personnel l' , performing the function of the STA shall be assigned to the I shift crew when a unit is in MODE 1, 2, 3, or 4. l 1 I

l . b 

        ~

TS.6.0-4 6.3 Plant Staff Qualifications 1 l Each member of the plant staff si:all meet or exceed the minimum l qualifications of Regulatory Guide 1.8, Revision 1, September 1975 l except for (1) personnel who perform the function of shift technical advisor shall hold an SRO license and havc a bachelors degree or l equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (2) the operations manager who shall meet the requirements of ANSI N18.1-1971, except that NRC license requirements are as specified in Specification 6.2.B.6. l I I 

   .                                                                                l TS.6.0-7 C. Post Accident Samplinz                                                  ,

This program provides controls that ensure the capability to obtain , and analyze reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following-

1. Training of pcrsonnel;
2. Procedures for sampling and analysis; and i
3. Provisions for maintenance of sampling and analysis equipment.

D. Radioactive Effluent Controls Program This program conforms to 10CFR50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. This program shall allocate releases equally to each unit. The liquid radwaste treatment system, waste gas treatment system, containment purge release vent, and spent fuel pool vent are shared by both units. Experience has also shown that contributions from both units are released from each auxiliary building vent. Therefore, all releases will be allocated equally in determining conformance to the design objectives of 10CFR50, Appendix I. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
2. Limitation on the concentrations of radioactive material I

released in liquid effluents to unrestricted areas, conforming to Appendix B to 10CFR20.1 - 20.601, Table II, Column 2; l

3. Monitoring, sampling, and analysis of radioactive liquid and i i gaseous effluents in accordance with 10CFR20.1302 and with the methodology and parameters in the ODCM; ,

i

4. Limitations on the annual and quarterly doses or dose i commitment to a member of the public from radioactive materials  !

in liquid effluents released from each unit to unrestricted , j areas, conforming to 10CFR50, Appendix I; ]

5. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM. Determination of projected dose contributions for radioactive effluents in accordance with the methodology in the ODCM at least monthly;

w TS.6,0-8

6. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriateportions of these systems are used to reduce releases of radioactivity when the projected doses in a period of one month from the liquid effluent releases would exceed 0.12 mrem to the total body or 0.4 mrem to any organ; or from the gaseous effluent releases would exceed 0.4 mrad for gamma air dose, 0.8 mrad for beta air dose, or 0.6 mrem organ dose;
7. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with Appendix B to 10CFR20.1 - 20.601, Table II, Column 1;
8. 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 10CFR50, ,

Appendix I; j

9. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix I; and
10. Limitation on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40CFR190.

E. Component Cyclic or Transient Limit This program provides controls to track the USAR, Section 4.1.4 cyclic and transient occurrences to ensure that components are maintained within the design limits. F. (Reserved) l G. (Reserved) I H. (Reserved) l l I. (Reserved) j i l l l l

4 e TS.6,0-9 l J. Exolosive Gas and Storare Tank Radioactivity Monitorine Program This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The program shall include:

1. The limits for concentration of oxygen in the waste gas holdup system and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria;
2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 78,800 curies of noble gases (considered as dose equivalent Xe-133); and l
3. A surveillance program to ensure that the quantity of l radioactivity contained in each of the following tanks shall be limited to 10 curies, excludin5 tritium and dissolved or i entrained noble gases ,

i' Condensate storage tanks Outside temporary tanks l

4. The provisions of TS 4.0 are applicable to the Explosive Gas 1 and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

K. Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of j both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and l sediment. L. Technical Snecifications Bases Control Program i This program provides a means for processing changes to the Bases of

these Technical Specifications.

l

1. Changes to the Bases or the Technical Specifications shall be made under appropriate administrative controls and reviews.
2. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
a. a change in the Technical Specifications incorporated in the license; or
b. a change to the USAR or Bases that involves an unreviewed safety question as defined in 10CFR50.59.
s*

o' , TS.6,0-11 t l l 6.6 Reporting Requirements l l The following reports shall be submitted in accordance with 10CFR50.4 l A. Occupational Exposure Report A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g., reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, and refueling. This tabulation supplements the requirements of 10CFR20.2206. The dose - assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. This report shall be submitted by April 30 of each year. B. Annual Radiological Environmental Monitoring Report The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses 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 the Offsite Dose Calculation Manual (ODCM), and in  ! 10CFR50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C. The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken l during the report period. In the event that some results are not I available for inclusion with the report, the report shall be I submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of sampling locations keyed to a table giving distances and directions from the reactor site; and the results of licensees participation in the Interlaboratory Comparison Program defined in the ODCM. C. Radioactive Effluent Report The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR50.36a and 10CFR50, Appendix I, Section IV.B.1.

4 e' , TS.6.0-14 6.7 High Radiation Area l A. Pursuant to 10CFR20, paragraph 20.1601(c), in lieu of the requirements of 10CFR20.1601, each high radiation area, as defined in 10CFR20, in which the intensity of radiation is greater than 100

mrem /hr but less than or equal to 1000 mrem /hr, shall be barricaded I and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RUP). Individuals qualified in radiation protection procedures (e.g., health physics technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates less than or equal to 1000 mrem /hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

1. A radiation monitoring device that continuously indicates the l radiation dose rate in the area.
2. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them. ,
3. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the radiation protection manager.

l B. In addition to the requirements of Specification 6.7.A above, areas I with radiation levels greater than 1000 mrem /hr shall be provided I with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable l stay times for individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV or transmitting radiation monitoring device) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area, i l

s
  • e' ,

a ATTACHMENT 3 i l SUPPLEMENT  ! to  ; LICENSE AMENDMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6 i to the Guidance of Standard Technical Soecifications , d ,. Response to NRC letter dated September 25,1996, to Roger O. Anderson from Beth A. l Wetzel,' Subject, " Review of Northern States Power Company Proposed Quality. . Assurance Program Description Changes and Prairie Island License Amendment  ! Related to Technical Specifications Administrative Controls (TAC Nos. M95130,  ! M95131, and M95709)." 

                                                                                                                                .l I
                             ' Answer to Paragraph 2 (a)' deleted text under proposed Section 6.4, " Procedures" (currently Section 6.5 " Plant Operating Procedures")'
                              . Each current Technical Specification requirement addressed in the December 14,1995              'j License Amendment Request Exhibit A Pages 8 - 10,' Item 14 is addressed as follows-l 6.5.A.1, This requirement is now a commitment under Regulatory Guide 1.33,                  I Revision 2, Appendix A, Section 3, Procedures for Startup, Operation, and                   ;

Shutdown of Safety-Related PWR Systems. Thus, this LAR proposes to remove [ this specific requirement from the Technical Specifications and the procedure l requirement has not been relocated to the Operational Quality Assurance Plan.'  ! b 6.5.A.2, This requirement is now a commitment under Regulatory Guide 1.33, t t Revision 2, Appendix A, Section 2, General Plant Operating Procedures, items

k. and I., and Section 6, Procedures for Combating Emergencies and Other Significant Events, item x. Thus, this LAR proposes to remove this specific  ;

requirement from the Technical Specifications and the procedure requirement has not been relocated to the Operational Quality Assurance Plan. l 6.5.A.3, This requirement is now a commitment under Regulatory Guide 1.33, , Revision 2, Appendix A, Section 2, General Plant Operating Procedures, item c., ! Section 5, Procedures for Abnormal, Offnormal, or Alarm Conditions, and i Section 6, Procedures for Combating Emergencies and Other Significant Events. { Thus, this LAR' proposes to remove this specific requirement from the Technical i

s* o' , Attachment 3  : USNRC Page 2 of 3 Specifications and the procedure requirement has not been relocated to the Operational Quality. Assurance Plan. 6.5.A.4, This requirement is now a commitment under Regulatory Guide 1.33, Revision 2, Appendix A, Section 8, Procedures for Control of Measuring and Test Equipment and for Surveillance Tests, Procedures , and Calibrations, Item

b. Thus, this LAR proposes to remove this specific requirement from the Technical Specifications and the procedure requirement has not been relocated to the Operational Quality Assurance Plan.  !

6.5.A.5, The current Technical Specification requirements for the Facility Emergency Plan will be relocated to the plant Emergency Plan. As discussed in the subject License Amendment Request, these procedures will continue to be j under regulatory control through 10CFR50.54(q), 50.54(t) and 10CFR50, i Appendix, Section V. Thus this License Amendment Request proposes to l remove these specific requirements in.m the current Technical Specifications. 6.5.A.6, This requirement is now a commitment under Regulatory Guide 1.33, ! Revision 2, Appendix A, Section 6, Procedures for Combating Emergencies and Other Significant Events, item w. The requirement to shut the plant down for site flood levels higher than 692 feet above MSL will be relocated to the USAR. Thus, this LAR proposes to remove this specific requirement from the Technical l Specifications and the procedure requirement has not been relocated to the Operational Quality Assurance Plan. 6.5.A.8, This requirement was revised in accordance with License Amendments , 122/115. l 6.5.B lead-in paragraph, These requirements are regulatory requirements of ! 10CFR20 and also commitments under Regulatory Guide 1.33, Revision 2, Appendix A, Section 7, Procedures for Control of Radioactivity (For limiting materials released to environment and limiting personnel exposure), item 7. Thus, this LAR proposes to remove these specific requirements from the l Technical Specifications and these procedure requirements have not been

relocated to the Operational Q';ality Assurance Plan.

h l

yf oj Attachment 3 USNRC l Page 3 of 3 6.5.B.3, These requirements have been relocated to the Emergency Plan. Thus, this LAR proposes to remove these specific requirements from the Technical Specifications and the procedure requirements have not been relocated to the Operational Quality Assurance Plan. 6.5.C, " Maintenance and Test", These requirements are generally contained in Appendix A to Regulatory Guide 1.33. However, for clarity these requirements have been relocated intact to the Operational Quality Assurance Plan in an amendment to be submitted to the NRC under separate letter. 6.5.D, These specific requirements were removed from the Technical Specifications in accordance with License Amendment 122/115. I J}}

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