Text

Proposed TS marked-up & Revised Pages,Incorporating Changes Associated w/951214 LAR
	ML20198B147
Person / Time
Site:	Prairie Island
Issue date:	12/29/1997
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20198B137	List: ML20198B133; ML20198B140; ML20198B147;
References
	NUDOCS 9801060209
	Download: ML20198B147 (58)
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ATTACHMENT 1 SUPPLEMENT 4 LICENSE AMENDMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specifications Appendix A, Technical Specification Pages Marked Up Pages as proposed in this Supplement Page Number Source for Markup Comments TS viil Original Submittal Added Containment Leakage Rate Program and revised affected pages TS-lx Supplement 3 Previously was page TS x

. TS x Supplement 3 Previously was page TS xl Table TS.4.1-Original submittal incorporated changes from Amendment 28, Page 1 of 2 129 TS.4.4 3 Original submittal Previously was page TS.4.4 4, page number revised due to Amendment 126, revised reference to 4.4.A.3 TS.6.0-9 Supplement (11/25/96)

Added requirement to perform testing every 31 days TS.6.0-10 Original submittal Added Containment i.eakage Rate Testing Program, added page header TS.6.011 Inserted new page Continuation of Containment Leakage Rate Testing Program TS.6.012 Supplement dated 11/25/96 Revised page number (was TS.6,011)

TS.6.013 Original submittal Revised page number (was TS.6.0-12)

TS.6.014 Original submittal Revised page number (was TS.S.013),

added page header TS.6.015 Supplement dated 11/25/96 Revised page number (was TS.6,0-14)

TS.6,0-16 Original submittal Revised page number (was TS.6,0-15),

added page header B.4.4 3 Orignal submittal Revised page number to incorporate Amendment 126 l

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l 9901060209 971229 pct ADOCK 05000282 l

1 P PDR-l

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TS vili TABLE OF CONTENTS (Contin d IS SECTION TITLE PAGE.

6.0 ADMINISTRATIVE CONTROLS TS.6.0 6.1 Responsibility TB.6,0-1 6.2 Organization TS.6.0-2 A. Onsite and Offsite Organizatic d TS.6.0-2 B. Plant Staff TS.6.0-2 6.3 Plant Staff Qualifications TS.6,0 4 6.4 Procedures TS.6.0-5 6.5 Programs and Manuals TS.6,0-6 A. Offsite Dose Calculation Ma 'al TS.6,0-6 B. Primary Coolant Bources Outt..de Containment TS.6.0-6 C. Post Accident Sampling TS.6,0-7 D. Radioactive Effluent Controls Pyrogram TS.6.0-7 E. Component Cyclic or Transient Limit TS.6,0-0 F.

(Reserv2d)

TS.6.0 0 G.

(Reserved)

TS.6 0-8 1*. (Reserved)

TS.6,0-8 1.

(Reserved)

TS.6,0-8 J. Explosive Gas and Storage Tank Radioactivity TS.6.0-9 Monitoring Program K. Diesel Fuel Oil Testing Program TS.6.0-9 L. Technical Specification Bkses Control Program TS.6.0-9

.NE pontalph @iM ga M RatelTesting;Propfts % ((.,!g MTS((l0l-1l0 6.6 Reporting Requirements TS.6,0-124 A. Occupational Exposure Report TS.6,0-12A B. Annual Radiological Environmental Monitoring Report TS.6.0-124 C. Radioactive Effluent Report TS.6,0-11.s D Monthly Operating Report TS.6,0-133 E. Core Operating Limits Report (COLR)

TS.6.0-153 6.7 High Radiation Area TS.6.0-154 3

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

6 i

f TS Ax r

l TEC10iICAL SPECIFICATIO!iS LIST OF TABLES t

TS TABLE TITLE 1-1 Operational Modes J3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points r

3.5+2A Reactor Trip System Instrumentation L S-2D Engineered r,afety Feature Actuation System Instrumentation 3.45-1 Event Me Atorir.1 Instrumentation i

4.1-1A Reactor 'rt lp Sy stem Instrumentation Surveillance Requirements 4.1-1B Engineered :..tety Feature Actuation System Instrumentation Surveillance Requirements 4.1-10 Miscellaneous Instrument Surveillance Requirements 4.1-2A Minimum Frequencies for Equipment Teste 4.1 2B Minimum Frequencies for Sampling Testa 4.2-1 Special Inservice Inspection Requirements 4.12 1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval e

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._. _ __.._, _._ _.., _.~ _ _. _.

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TS x6 I

APPENDIX A TEcini! CAL SPECIFICATIONS f

l LIST OF FIGURES

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l TS FIGURE TITLE 2.1*1 Reactor Core Safety Limits I.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 t

3.1 3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit-i Versus Percent vf RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I 131

[

-3.0-1 Spent Fuel Pool Unrestricted Region Burnup and Decay Time l

Requirements - OFA Fuel 3.0-2 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Requirements - STD Fuel 3.10-1 Required shutdown Margin Vs Reactor Boron Concentration 4.4-1 shield Building Design In-Leakage Rate c

5.6 1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Pool Checkerboard Interface Requirements 5.6-3 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, No GAD 5.6-4 Spent Fuel Pool checkerboard Region Burnup and Decay Time Requi%ements - STD Fuel, No GAD 5.6-5 Spent Fuel Pool Checkerboard Region Barnup and Decay Time Requirements - OFA Fuel, 4 GAD 5.6-6 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Puel, 4 GAD 5.6-7 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, O GAD 5.6-8 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 8 GAD 5.6-9 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 12 GAD 5.6-10 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 12 GAD 5.6-11 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 16 or More GAD 5.6-12 Spent Fuel Pool Checkerboard Region Burnup and Decay Time

' Requirements - STD Fuel, 16 or More GAD

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Table TS.4.1-28 (Page 1 of 2)-

TAB LE TS. 4.1 - 2D i

MINIMUM FPEOi,@lKIFS FOB l R LING TESTS i

i TEST FREOUENCY 1.

RCS Gross 5/ week Activity Determination

[

2.

RCS Isotopic Analysis for DOSE 1/14 days (when at power)

EQUIVALENT I-131 Concentration 3.

RCS Radiochemistry E determination 1/6 months (1) (when at power) 4.

RCD Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including 2-331, I-133, and I-135 the specific activity ex+

ceeds 1.0 uC1/ gram DOSE _

EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.

RCS Radiochemistry (2)

Monthly 6.

RCS Tritium Activity Weekly 7.

Deleted 8.

RCS T Jron Concentration * (3) 2/ Week (4) 9.

RWST Doron Concentration Weekly

10. Doric Acid Tanks Doron Concentration 2/ Week 11', caustic Standpipe NaOH Concentration Monthly I

12. Accumulator Boron Concentration Monthly j

13. Spent Fuel Pit Doron Concentration Mont4MWeekl;M L

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7S.4.4-34 b.

Cold DOP testing shall be performed after each complete or partial replacement of a HEPA filter bank or after any structaral maintenance on the system housing that could affec' the HEPA bank bypass leakage, c.

Halogenated hydrocarbon testing shall be performed atter each complete or partial replacement of a charcoal adsorber bank or af ter any structural n.alntenance on the system housing that could affect the charcoal adsorber bank bypass leakage.

t d.-

Each circuit shall be operated with the heaters on at least i

10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.

)

5. Perform an air distribution test on the HEPA filter bank after any maintenance.or testing that could affect the air distribution within the systems.- The test shall be performed at rated flow i

rate (110%). The results of the test shall show the air distribution is uniform w:'hin 120%.

C. containment Vacuum Dreakers i

The air-operated valve in each vent line shall be tested at quarterly intervals to demonstrate that a simulated contain-mant vacuum of 0.5 psi will open the valve and a simulated accident signal will close the valve. The check valves as well as tna butterfly valves will be leak-tested during each

-refueling shutdown in accordance with the requirements of Speci-1 fication 4.4.A.;34.

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l TS.6,0-9 J.

Exolosive Oas and Storace Tank Radioactivity Monitorina Procram i

This program provides controls for potentially explosive pas mixtures l

contained-in the waste gas holdup system, the quantity of i

radioactivity contained in gas storage tanks, and the quantity of

?

radioactivity contained in unprotected outdoor liquid storage tanks.

l 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. Sach limite shall be appropriate to the system's

-design criteriar

2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 78,000 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 disselved or entrained noble gases:

l Condensate storage tanks Outside temporary tanks

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

K.

Diesel Fuel Oil Testina Proqram 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. Dieselyne1[oilf testingishalljbe[ performed (every(311dayoi L.

Technical Snecifications 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 made under appropriate administrative controls and reviews.

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i 79.6.0-10

!/f J yschid eaP Soggi fi cat ion' Bases Cont rop ProQram] (continded)]

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 Spec.ifications incorporated in the licences or

b. a change to the USAR or DaJes that involves an unreviewed safety question as defined in 10CFR50.59,

3. The Bases Control Program shall contain provisions U ansure that the Bases are maintained consistent with the USAR.

I

4. Proposed changes that meet the criteria of Specification 6.5.L.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

M.f contaf nment~Leakaoe" Rate"Testina ProuGUD 1

K program'shallibe satabli' shod t6~ implement]theileakage7 tate testing of the ' containment. au required' byi10 CFR.50.54 (o)L and 20 CFR 50i t

Appendix & Option bj. as modified by approved exemptionsi This program nhall be--in accordance.vith the guidelinen contained in Regulatory, Guide 1.163, ' Performance Based. Containment; Leak-Test Program E dated September l1ppsg The?peakccalculated~containmedt" internal 'presisure, f or ' the design' basis loss of coolant accidentLis less,than the centainment internal Y

o design.precaute, P.,g oi; 46;;saiR

':te " mas.imum7 allowable ~ primaryTecntaininent? 1eakage? rat e y;L.', Tat P.,-

phallibe 0,250 of primary, containment! air weight lper day. Por pipes connected,tolsystemel that;arelin the-auxiliary building special ventilation zone,cthettotal: leakage shall.belless;than 0;11 of i

primary - containmentIair3eight perf day at! pressure: P..( For pipes connected.to : systems 1that are exterior to;(bothL tfN shisid building t

1 and! the auxiliary ~ building opecial: ventilation zonei; the - total leakage past isolationtvalves shall be lent;than10f0 W of primary contat:: ment; airyeight; per day; at; presaure; P..;

Le akage ; Rate [accep'tiance ! cri t e r j a' a re {

aEPrimarp[contalnmentpleakagelratelacceptanceTeriterion isT1.0} Li.

^

prior to unittotartup,1fo11owingiteeting inEaccordance with the programb the1 combined leakage rate acceptance criteria areJs.0 60 L/ f orf allicomponents subhet to ; Type. B andJype C; testa and 3 p.75;Lf for Typej Af tentsj a

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4 4

TS.6.0-11 M.

gpnt aintnerikeakaue"Rsts TestinciProggg(coritinued);

2 h ?'Ai(16cki te e t inginceept a ricelerirje ri al a re i 1.')[pve0Q1;lai,rl_ lock]1eakagef'fratefjyl{0[0$J1dwhenjtested[atf ar46 peig 2){P6r~ ~each7doorIintergaskeiltestj[ leakage rate;(in,lsT OTOE1V whe.d pressurised to;a10;psig?

The ' provis f oris Tof ? 4"; 0*.' A;do~ notT appip" t 6 ' the i t est* f requenc14s ?

,specified in the Containment l Leakage Rate Testing Program',' The Containment ' Leakage Rate.jTest.i.ngfProgram pelpulates: accept able pytelhion;ot,ttst-jntervalej The px'ovisioniTof' 4!0W (ekeeptsthat'TtheVa11owedisbrVe111'Ance

' ntervalsl:areldefinedLby-the_ Containment Leakage Rate Testing i

~

Program): art applicab1le; toithel Containment;(. eakage RateLTesting L

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TS.6.0-124

6.6-Reporting Requirements The following reports shall be submitted in accordance with 10CFR50.4 A,

occunational Exposure Renort A tabulation on an annual basis of the number of station, utility and other personnel iincluding contractore) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according tc 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 accounted for. In the aggregate, at.1 east 80% of the total whole body dose received from extecnal sources.nould be assigned to specific major work functions. This report shall be submitted by April 30 of each year.

B.

bnnual Radiological Environmental Monitorina Report The Annual Radiological Environmental Monitoring Report covering the operation of the plart during the previous calendar year shall be submitted by May 15 of each yenr. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental menitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCMi, and in 10CFR50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

l The Annual Radiation Environmental Monitoring Reports shall include j

summarized and tabulated results in the format of Regulatory Guide 4.0, December 1975 of all radiological environmental camples 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.

Tha missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following s 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 Interlaborstory Comparison Program defined in the ODCM.

C.

D' ioactive Effluent Report The Radioactive Effluent Report covering the operation of the plant during he 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 objectiece outlined in tue ODCM and in conformance with ACCFR50.36a and 10CFRSC, Apper.lix I,Section IV.B.1.

TS.6.0-133 D.

Monthlv oneratina Peoog.s.

Routire reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

E.

Egre Ooert tina Limits Report (COLR)

1. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a ro.oad cycle, and shall be documented in the COLR for the followings arr a, Heat Flux Hot Channel Factor Limit (F

),

hoclear Enthalpy g

RTP Rise Hot Channel Factor Limit (Fan ),

PFDH, K(Z) and V(Z)

(Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band (Specificatsons 3.10 D.4 through 3.10.B.9)

n. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

d. Reactor Coolant System Flow Limit (Specification 3.10.J)

2. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD-8101-A, " Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102 A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version)

WCAP-9272-P-A, " Westinghouse Reloed Safety Evaluation Methodology", July, 1985 WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code", August, 1985 WCAP-10924-P-A, " Westinghouse Large-Break LOCA Best-Estimate Methodology", December, 1988 WCAP-10924-P-A, Volume 1, Addendum 4, "Wescinghouse Large Break LOCA Best Estimate Methodology", August, 1990 XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Disti.bution Control for Pressurized Water Reactors Phase II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: W-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLO p4 Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993).

I l

TS.6.0-134 Kylcsre ' Orie rat ina* Li nii t s 2 Reoort9 COLR) y c6nt inQed[;

NSPNAD-93003-A, " Transient Power Distributicn Methodology",

(latest epproved version) 4

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

4. The COLR,-including any midcycle revisions _or supplements, shall

[

be provided upon issusnee for each reload cycle to the NRC.

O

TS.6,0-154

6. 7_

High Radiation Area A.' Puruuant to 10CFR20,'paragrakh 2 0.1601 (c), in lieu of the requirements of 10CFR20.1601, each high radiation area, as defined in l

10CFR20, in which the intensity of radiation de greater than'100-mrem /hr but less than or equal to '1000 mrem /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 (e.g., health physics technicians) or personnel continuously escorted by such individuals may be exenpt 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 his'2 radiation areas.

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

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

2.-A radiation monitoring device that continuously integcetes 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 dos" 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.

B.

In addition to the requirements of Specification 6.7.A above, areas with radiation levels greater than 1000 mrem /hr shall be provided

-with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be mainthined 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 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 qualifted in radiation protection procedures to provide positive-exposure control over the activities being performed within the area.

l

O TS 6,0-1(F

$17$M90"di4545$NSA5(5MhQnysd[

C.

For individual high radiation areas with radiation levels of greater than 1000 mrem /hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for puposes of locking. or that cannot be continuously guarded, and where no enclosure can be reasonably consructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

e

t B.4.4-33 4.4 CONTAINMENT SYSTEM TESTS Bases continued Several penetrations of the conw31nment vessel and the shield building could, in the event of leakage pait their isolation valves, result in leakage being conveyed across thc annulus by the penetrations themselves, thus bypassing the function of the Shield Building Ventilation System (Reference 5)

Such leakage is estimated not to exceed.325% per day.

A special zone of the auxiliary building has minimum-leakage construc-tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant trains of the Auxiliary Building Special Ventilation System.

This system, when activated, will supplant the normal ventilation and draw a vacuum throughout the zone such that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack.

The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure. Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day.

The resulting two-hour doses at the nearest SITE BOUNDARY and 30-day doses

~

at the low population zone radius of 1H miles are less than guidelines presented in 10CFR100.

Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage tnan the design basis. The staff has reevaluated doses for these higher inleakage rates and found that for a primary containment leak rate of 0.25% per day at peak accident pres-sure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment in-leakage (Reference 6).

ATTACHMENT 2 SUPPLEMENT 4 i

LICENSE AMENOMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6

- to the Guidance of Standard Igchnical Specifications Appendix A, Technical Specification Pages Revised Pages as proposed in this Supplement TS /ili TS-lx TS-x Table TS.4.1-2B (Page 1 of 2)

TS.4.4-3 TS.6.0-9 TS.6.0-10 TS.6.0-11 TS.6.0-12 TS.6.0-13 TS.6,0-14 TS.6.0-15.

TS.6,0-16 B.4.4-3

'd'.

TS-viii TAbt6 OF CONTENTS (Continued)

TS SECTION TITLE PAGE 6,0; ADMINISTRATIVE CONTROLS TS. 6 '. 0 - 1 6.1 Responsibility-

-TS.6;0-1 TS.6.0-2'-

6.2

_Organizstion

_TS.6.0-2 A. Onsite and Offsite Organizations B. Plant Staff TS.6.0-2 6,3 Plant Staff Qualifications

.TS.6.0-4 6.4 Procedures

.T4.6.0-5

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6.5 Programs and Manuals.

'46.6.0-6 A. Offsite Dose Calculation Manual TS.S.0 B. Primary Coolant Sources Outside Containment TS.6.0-6 C. Post Accident Sampling TS.6,0-7

. D. Padioactive Ef fluent Controls Prograni TS. 6, 0 -

.E. Component Cyclic or Transient Limit TS.6.0-8 F.

(Reserved)

TS.6,0-8 G.

(Reserved)

TS.6.0-8 H.

(Reserved)

TS.6,0-8 I.

(Reserved)

TS 6,0-8 J. Explosive Gas and Storage Tank Radioactivity TS 6.0-9 Monitoring Program K. Diesel Fuel Oil Testing Program TS.6.0-9 L. Technical Specification Bases Control Program IS.6,0-9 M. Containment Leakage Rate Testing Program TS 6.0-10 4

6.6 Reporcing Requirements TS.6,0-12 A. Occupational Expocure Report TS.6.0-12 B. Annual Radiological Environmental Monitoring Report-TS.6.0-12 C. Radioactive Effluent Report TS.6.0-12 D. Monthly Operating Report TS.6.0-13 E.. Core Operating Limits' Report (COLR)

TS.6,0-13 6.7.

High Radiation Area TS.6.J-15

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-TS-ix

-TECHNICAL SPECIFICATIONS LIST OF TABLES TS TbJ.LF TITLE 2

1-1 Operational Modes 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 Instrumentacion 3.15 Event Monitoring Instrumentation j

4.1-1A Reactor Trip System Inotrumentation Surveillance Requirenents 4.1-1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-1C Miscellaneour Instrument Surveillance Requiremente 4.1-2A Minimum Frequencies for Squipmert Teste 4.1-2B-Minimum Frequencies for Sampling Tests 4.2-1.

Special Inservice Inspection Requiremente 4.12-1 Steam Generator Tube Inspection C

4.13-1 Snubber Visual Inspection Interval

)

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Ec-l TS-x 7c-APPENDIX A TECHNICAL SPECIFICATIONS a.

LIST OF FIGURES TS FIGURE TITLE 2.1 Reactor Core Safety Limita 3.1 -1L Unit 1 and Unit 2-Reactor Coolant System HeaS p Limitationo f

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 3.8-1 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Requirements - OFA Fuel 3.8-2 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Requirements - STD Fuel 3.10-1 Required Shutdown-Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-1 Spent Fuel Pool Checkerboard Interface Requirements 5.6-3 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, No GAD 5.6-4 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, No GAD 5.6-5 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 4 GAD 5.6-6 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 4 GAD-5.6-7 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 8 GAD-5.G-8 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 8 GAD 5.6-9 Spent Fuel Pool Checkerboard _ Region Burnup and Decay Time Requirements - OPA Fuel, 12 GAD 5.6 Spent Fuel Pool Che.erboard Region Burnup and Decay Ti Requirements - STD Fuel, 12 GAD 5'.6-11 Spent Fuel P3ol Checkerboard Region Burnup and Decay Time Requiremento OFA Fuel, 16 or More GAD 5.6-12 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 16 or More GAD

~

c 6

l Table TS.4.1-2B (Page 1 of 2)

TABLE TS.4.1-2B MINIMUM FREOUENCIES FOR SAMPLING TESTS TEST FREOUENCY 1.

RCS Gross 5/ week Activity Determination 2.

RCS Isotopic Analysis for DOSE 1/14 days (when at power)

EQUIVALENT I-131 Concentration 3.

RCS Radiochemistry E determination 1/6 months (1) (when at power) 4.

RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram DOSE'_

EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.

RCS Radiochemistry (2)

Monthly

- 6.

RCS Tritium Activity Weekly 7.

Deleted N

1 8.

RCS Boron Concentration * (3) 2/ Week (4) 9.

RWST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week

11. Caustic Standpipe NaOH Concentration Monthly

12. Accumulator Boron Concentration Monthly

13. Spent Fuel Pit Doron Concentration Weekly

y p-TS.4.4-3 b.

cold DOP testing shall be performed af ter each complete or partial replacement of a HEPA filter bank or after any structural mainter.ance on the _ system housing that could affect the HEPA bank bypass leakage.

Halogenated hydrocarbon testing shall be performed after c.

each complete or partial replacement of a charcoal adsorber bank or after any structural maintet.ance.on the system housing that could affect the charcoal adsorber bank bypass leakage.

d.

Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.

5. Perform an air distribution test on the HEPA filter bank after any maintenance or testing that could affect the air distribution within the systems. The test shall be performed at rated flow rate (110%). The results of the test shall show the air distribution is uniform within 120%.

C. Containment Vacuum Breakers The air-operated valve in each vent line shall be tested at quarterly intervals to demonotrate that a simulated contain-mant vacuum of 0.5 psi will open the valve and a simu3ated accident signal will close the valve.

The check valves as well as the butterfly valves will be leak-tested during each refueling shutdown in accordance with the requirements of Speci-ficaticn 4.4.A.3.

L

TS.6.0-9.

J.

Exnlosive Gas and Storace Tank Radioactivity Monitorino Procram This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system, the quantity of radioactivity contained in cas storage tanks, and the quantity of radioactivity con

.ned in unprotected outdoor liquid storage tanks.

The program shall include:

1. The limits'for rencentration 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; t

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

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 1.0 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

K.

Diesel Fuel Oil Testino Procram 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 accordarce with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. Diesel fuel oil testir_g shall be performed every 31 days.

L.

Technical Specifications Bases Control Procram This program provides a means for processing changes to the Bases of these Technical Specifications.

P

1. Changes to the Bsses or the Technical Specifications shall be made

.under appropriate administrative controls and reviews.

TS.6'.t-10 L.

Technical Specificat$on' Bases control Procram (continued) i

2. Licensees may make changes-to Bases without prior NRC approval _

provided the changes do not involve either of the followings

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 es defined-in 10CFR50.59.

3. The Baces Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

4. Proposed changes that meet the criteria of Specification 6.5.L.2 above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

M.

Containment Leakace Rate Testino Procram A program shall be established to implement the leakage rate testing of the containment as _ required by 10 CFR 50.54 (o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure, P.,

of 46 psig.

The maximum allowable primary containment leakage rate, L.,

at P.,

shall be 0.25% of p imary containment air weight per day. For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.1% of primary containment air weight per day at pressure P..

For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.01% of primary containment air weight per day at pressure P.

Leakage Rate _ acceptance criteria are

a. Primary containment leakage rate acceptance criterion is s 1.0 L,.

Prior.to unit startup, following testing in accordance with the program, the combined leakage rate acceptance criteria are s 0.60 L for all ' components subject to Type B and Type C tests and s 0.75 L, for Type.A tests;

1 TS.6.0-11 M.

gontainment Leakace Rate Testina Program (continued) b.' Air lock' testing' acceptance criteria are:

1) ~ Overall air lock leakage rate is s 0.05 L when. tested at =46 psig

2) ' For each door intergasket test, leakage rate is s 0.01 L, when pressurized to =10 psig.

-The provisions et 4.0.A do not apply to the test frequencies.

specified in the Containment Leakage Rate Testing Program. The Containment Leakage Rate Testing Program stipulates acceptable extension of test intervals.

The provisions of 4.'J.B (except that the-allowed surveillance intervals are defined by the Containment Leakage Rate Testing Program) are applicable to the Containment Leakage Rate Testing Program.

.=

p TS.6.0-12

'6.6 Reporting Requirements The following reports shall be submitted in accordance with 10CFR50.4 A.

Occupational Exposure Reoort A tabulation on an annual basis of the number of station, utility at.d 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 requirementr 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 Radioloolcal Environmental Monitorino 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 lof trends of the results of the radiological ervironmental 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, Lections 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 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.

The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the followings 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 recults 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 1,Section IV.B.1.

j l

,e TS.6,0-13 D.

Monthly Doeratino Reporte Routine reports of operating statistics.and shutdown experience shall' be submitted on a monthly basis no later-than the 15th of each month following the calendar month covered by the report.

-E.

Core Operatino Limits-Report (COLR)

1. Core operating limits shall be established prior to each reload I

cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the followings arr

a. Heat Flux Hot Channel Factor Limit (F

),

Nuclear Enthalpy g

RTP Rise Hot Channel Factor Limit (Fox ), PFDH, K(Z) and V(Z)

(Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10.B.9)

c. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

d. Reactor Coolant System Flow Limit (Specification 3.10.J)

2. The analytical methods used to determine the core operating _

limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD-8101-A, " Qualification of Reactor Physics Methods for

-Application to PI Unita" (latest approved version)

NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version)

WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology", July, 1985 WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code", August, 1985 WCAP-10924-P-A, " Westinghouse Large-Break LOCA Best-Estimate Methodology", December, 1988 WCAP-10924-P-A, Volume 1, Addendum 4,

" Westinghouse Large Break LOCA Besc Estimate Methodology", August, 1990 XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: W-COBRA / TRAC 2-Locp Upper Plenum Injection Model Update to Support ZIRLO n 7

Cladding Opcions", April 1993 (approved by NRC SE dated November 26, 1993).

o

,e TS.6.0 E.

Core Operat(na Limite Report (COLR) (continued)

NSPNAD-93003-A, " Transient Power Distribution Methodology",

(latest approved version)

3. The core operating limits shall be determined such that all applicable limits (e.g.,

fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of t.34 safety analysis are met.

4. The COLRi including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

l l

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- - -, - ~ - -. - ~.

- ~

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+

TS.6.0-15

?

.6.7,High Radiation, Area-

A.

-Pursuant to 10CFR20,. paragraph 20 1601(c), in lieu,of the l

requirements ofL10CFR20.1601,' each_high radiation. area, as definedfin 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.

l nand conspicuously posted as_a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP)'. :Individuale qualified in. radiation protection

-procedures (e.g., health physics technicians) or personnel continuously escorted by such individuals may be exempt from the RWP i

issuance requirement during the performance of their' assigned: duties in high radiation;ateas with. exposure rates less than or equal to_

1000 mrem /hr, provided they are otherwise.following: plant radiation j

protection procedures for entry into such high radiation areas..

l Any' individual or group of individuals permitted _to enter such areas

- i

'shall be provided with or accompanied by'one or_ nore of the-followings.

1. A radiation monitoring dev3ce_that continuously _ indicates the 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 leveln=in the 4

area have been established and personnel are aware of them.

3. Jul individual-qualified in reliation protection procedures with a radiation dose rate monitoring device, who is responsible for providina positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the radiation protection manager.

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 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 stay times:for individuals in those areas. In lieu-oflthe stay time

.opecification of the RWP, direct or remote (such as closed circuit TV or _ transaltting radiation monitoring device) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide-positive exposure control over the activities being performedivithin~the area.

c

+,

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~

~.-.

n

TS.6.0-16 6.7 liigh Radiatio 11 Area (continuted)

C.

For individual high radiation areas with radiation levels of greater than 1000 mrem /hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for puposes of locking, or that carmot be continuously guarded, and where no enclosure can be reasonably consructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

l t

4 B.4.4-3 4.4 CONTAIUMENT SYSTEM TESTS Bases continued Several penetratic a of the containment vessel and the,hield building could, in the event of leakage past the.r isolation valves, result in leakage being conveyed across the annulus by the penetrations themselves, thus bypassing the function of the Shielt Building Ventilation System (Reference 5).

Such leakage is estimated act to exceed.025% per day.

A special zone of the auxiliary building has mindeum-leakage construc-tion and controlled access, and is designated as a special ventilation tone where such leakage would be collected by either of two redundant trains of the Auxiliary Building Special Ventilation System. This system, when activated, will supplant the normal ventilation and draw a vacuum throughout the zone cuch that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack.

The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure.

Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day.

The resulting two-hour doses at the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of IM miles are less than guidelines presented in 10CFR100.

Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage than the design basis. The staff has reevaluated doses for these higher inleakage rates and found that for a j

primary containment lesk rate of 0.25% per day at peak accident pres-sure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lover secondary containment in-leakage (Reference 6).

w-

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' ATTACHMENT 3 SUPPLEMENT 4 LICENSE AMENDMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specifications Final License Amendment Pages Note: All of the current Te.hnical Specification Chapter 6 pages were deleted, including Table 6.1-1, by the original submittal and replaced with pages TS.6.0-1 through TS.6.0-15.

Page Number Source Document Page Number Source Document TSil Original submittal TS.6.0 2 Supplement (11/25/96)

TS-v Supplement (11'25/96)

TS.6.0 3 Supplement 3 TS-vili Supplement 4 TS.6.0 Supplement (11/25/96)

TS-lx Supplement 4 TS.6,0-5 Supplement 2 TS-x Supplement 4 TS.6.0-6 Original submittal TS-xl deleied Supplement 3 TS.6.0-7 Supplement (11/25/96)

TS xii deleted - Supplement 3 TS.6.0-8 Supplement (11/25/96)

TS-xiii deleted - Original submittal TS.6.0-9 Supplement 4 T'

-10 Original submittal TS.6.0-10 Supplement 4 TS.3.1-11 deleted - Original submittal TS.6,0-11 Supplement 4 Table Supplement 4 TS.6.0-12 Supplement 4 TS.4.1-26 (Page 1 of 2)

TS.6.0-13 Supplement 4 TS.4.4-3 Supplement 4 TS.6.0-14 Supplement 4 TS.4.6-1

. Original submittal TS.6.0-15 Supplement 4 TS.S.1-1 Original submittal TS.6.0-16 Supplement 4

- TS.S.1-2 Originc' submittal B.3.1 9 deleted - Original submittal TS.6.0-1 Original submittal B.4.4-3 Supplement 4

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'TS-11

. TABLE OF CONTENTS (Continued)=

NfS SECTION--

TITLE PAGE

-3.;' LIMITING' CONDITIONS FOR OPERATION

-3.0 Applicability.

TS.3.0-1 3.1 Reactor' Coolant System

-TS.3.1-1.

.A. Operational Components-TS.3.1-1 1.- Reactor-Coolant-Loops and-Coolant Circulation TS.3.1-1 2.; Reactor Coolant System Pressure Control TS.3.1 a.! Pressurizer:

TS,3.1-3:

b. Pressurizer Safety Valves

_TS.3.1 c. Pressurizer Power Operated Relief hj Valves 1 TS.3.1-4

3. Reactor: Coolant Vent. System TSi3.1-5 B. Pressure / Temperature Limits TS.3.1-6 1._ Reactor, Coolant; System TS.3.1 2. Pressurizer TS.3.1-6

3. Steam Generator-TS.3.1-7

-C. Reactor Coolant System Leakage TS.3.1-8

1. Leakage Detection

.TS.3.1-8

2. Leakage Limitations TS.3.1-8

3. Pressure Isolation Valve Leakage TS.3.1-9 D. Maximum Coolant Activity TS.3.1-10

.i E. Deleted

^

F. Isothermal Temperature Coefficient.(ITC)

TS.3.1 ;

3.2 Chemical and Volume Control System TS.3.2-1 p

3.3 Engineered Safety Features TS.3.3-l' A.-SC0ety Injection and Residual Heat R9moval Systems' TS23.3-1 B. Containment Cooling Systems-TS.3.3-4

- j C. Component Cooling Water System TS.3.3-5 D. Cooling Water System TS.3.3-7 J3.4 Steam and Power Conversion System TS.3.4-1 A. Steam Generator Safety and-Power Operated

' Relief Valves-TS.3.4-1 B. Auxiliary Feedwater System TS.3.4-1 C. Steam Exclusion System--

-TS.3.4-3 D. Radiochemistry TS.3.4-3 3.5, 1 Instrumentation System TS.3.5-1 x

=";

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

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?

- e TS-4 TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE' 4.0-SURVEILLANCE 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 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. Containraent 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 4.5 Engitsecred Safety Features TS.4.5-1 A. System Tests TS.4.5-1

1. Safety Injection System TS.4.5-1

2. Containment Spray System TS.4.5-1

3. Containment Fan Coolers TS.4.5-2

4. Corponent Cooling Water System TS.4.5-2

5. Cooling Wat7r System TS.4.5-2 B. Component Tests TS.4.5-3

1. Pumps TS.4.5-3

2. Containment Fan Motors TP.4.5-3

3. Valves.

TS.4.5-3 4.6 Periodic Testing of Emergency Power System TS.4.6-1 A. Diesel Generators TS.4.6-1 B. Station Batteries TS.4.6-3 C. Pressurizer Heater-Emergency Power Supply TS.4,6-3 4.7 Main Steam Isolation Valves TS.4.7-1 4.8 Steam and Power Conversion Systems TS.4.8-1 A. Auxiliary Feedwater System TS.4.8-1 B. Steam Generator Power Operated Relief Val-TS.4.8-2 C. Steam Exclusion Syatem TS.4.8-2

-4.9 Reactivity Anomalies TS.4.9-1 4.10 Deleted 4.11 Deleted

' s' TS-viii TABLE OF' CONTENTS (Continued)

TS SECTION TITLE PAGE

.6.0

ADMINISTRATIVE CONTROLS TS.6.0-1.

6.1 Responsibility.

TS.6,0-1 6.2 Organization -

TS 6.0-2 A onsite and Offsite Organizations TS.6.0-2 B. Plant Staff TS 6.0-2 6.3 Plant Staff Qualifications.

TS.6,0-4 6.4 Procedures TS.6,0-5 6.5 Programs and Manuals TS.6.0-6 1

A. Offsite Dose Calculation Manual'

TS.6.0-6 B. Primary Coolant Sources outside Containment TS.6,0-6 C. Post Accident Sampling TS.6,0-7 D. Radioactive Effluent Controls Program TS.6,0-7 E. Component Cyclic or Transient Limit TS.6,0-6 F.

(Reserved)

TS.6.0-8 G.

(Reserved)

TS.6,0-8

+

H.

(Reserved)

TS.6,0-8 I.

(Reserved)

TS.6.0-8 J. Explosive Gas and Storage Tank Radioactivity TS.6,0-9 Monitoring Program K. Diesel Fuel Oil Testing Program TS.6.0-9 L. Technical Specification Bases Control Program TS.6.0-9 M. Containment Leakage Rate Testing Program TS.6,0-10 6.6 Reporting Requirements TS.6.0-12 A. Occupational Exposure Report TS.6,0-12 B. Annual Radiological Environmental Monitoring Report TS.6.0-12 C. Radioactive Effluent Report TS 6,0-12 D. Monthly Operating Report TS.6.0-13 E. Core Operating Limits Report (COLR)

TS.6,0-13 6.7 High Radiation Area TS.6,0-15 L

TS-ix TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 1-1 Operational Modes 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2A Reactor Trip System Instrumentation 3.5-29 Engineered Safety Feature Actuation System Instrumentation 3.15-1 Event Monitoring Instrumentation 4.1-1A Reactor " iip System Instrumentation Surveillance Requirements 4.1-1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1 1C Miscellaneous Instrument 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 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval

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TS-x-l APPENDIX A= TECHNICAL SPEC 7FICATIONS

-Lift OF FIGURES 5

TS FIGURE TITLE-2.1-1 Reactor Core Safety Limits

~

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 D*?S 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=

3.8-1 Spent Fuel Pool Unrestricted Region Burnup and Decay. Time Requirements - OFA Fuel

-3.8-2 Spent' Fuel Pool Unrestricted Region Burnup and Decay Time Requirements - STD Fuel 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration-

-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 Interface Requirements 5.6-3 Spent Fuel Pool checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, Fo GAD 5.6-4 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, No GAD 5.6-5 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 4 GAD 5.6-6 Spent Fuel, Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 4 GAD 5.6-7 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 8 GAD 5.6-8 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 8 GAD 5.6-9 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 12 GAD 5.6-10 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 12 GAD 5.6-11 Spent Fuel Pool. Checkerboard Region Burnup and Decay Time Requirements - OFA Fuel, 16 or More GAD 5.6-12 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 16 cr More GAD

TS.3.1-10 3.1.D. MAXIMUM COOLANT ACTIVITY-1.

The specific activity of the primary coolant (except as specified in 3.1.D.2. and 3 below) shall be limited to a.

Less than or equal to 1.0 microcuries per gram DOST EQUIVALENT-I-131,-and b.

Less than or equal to 100/E microcuries per gram of gross radioactivity.

2.

If a reactor is critical or the. reactor coolant system average temperature is greater than or equal to 500*F:

a.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure TS.3.1-3, the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With the gpecific activity of the primary coolant greater than 100/E microcurie per gram, the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.

If a reactor is at or above COLD SHUTDOWN, with the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.1-2B until the specific activity of the primary coolant is restored to within its limits.

1 4

i Table TS.4.1-2B (Page 1 of 2) i TABLE TS.4.1-2B MINIFUM FREQUENCIES FOR SAMPLING TESTS 1TCT FRtOUENCY 1.

RCS Gross 5/ week Activity Determination 2.

RCS Isotopic Analysis for DOSE 1/14 days (when at power)

ROUIVALENT I-131 Col. centration 3.

RCS Radiochemistry E detennination 2 /6 months (1) (when at power) 4.

nCD Isotopic Analycle for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including I-131, I-133, and I 135 the specific activity ex-coeds 1.0 uCi/ gram DOSE _

EQUIVA* ENT 1-131 or 100/E uCi/ gram (at or above cold shutdeyn), and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following ther al POWER change exceed.ng 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5.

R,3 Radiochemistry 8,2)

Monthly 6.

RCS Tritium Activity Weekly 7,

Deleted 8.

RCS Boron Concentration * (3) 2/ Week f (4) 9.

RWST Boron Concentration Weekly

10. Boric Acid

nke Boron Concentration 2/ Week

'11. Caustic Standpice NaOH Concentration Monthly

12. Accumulator Boron Concentration Monthly

13. Spent Fuel Pit Boron Concentration Weekly

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TS.4.4-3 b.

Cold DOP testing shall be performed af ter each complete or partial replacement of a HEPA filter bank or after any structural maintenance on the system housing that could affect the HEPA bank bypass leakage.

i c.

Halogenated hydrocarbon testing shall be performed after i

cach enmplete or partial replacement of a charcoal adsorber bank or after any structural maintenance on the system housing that could affect the charcoal admorber bank bypass-l 1eakage.

I d.

Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.

5. Perform an air distribution test on the HEPA filter bank after any maintenance or testing that could affect the air distribution within the systems.

The test shall be performed as rated flow rate (1 0%). The resulta 9f the test shall show the air 1

distribution is uniform within 120%.

C. containment vacuum Breakers The air operated valve in each vent line shall be tested at quarterly intervals to demonstrate that a simniated contain-mant vacuum of 0.5 psi will open the valve and a simulated accident signal will close the valve. The check valves as well as the butterfly valves will be leak-tested during each refueling shutdown in accordance with the requirements of Speci-fication 4.4. A.3.

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s TS.4.6-1 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM Applicability Applies to periodic testing and surveillance requirements of the emergency power uystem.

Obiective To verify that the emergency power sources and equipment are OPEPABLE.

Freecifi cation The following tests and surveillance shall be performed:

A.

Diesel Generators 1.

At least once each month, for each diesel generator Verify the fael level in the day tank, a.

b.

Verify the fuel level in the fuel storage tank, c.

Deleted Veiify the fuel transfer pump can be started and transfers fuel d.

from the storage systam to the day tank.

Verify the diesel generator can start and gradually accelerate. Verify e.

the generator voltage and frequency can be adjusted to 4160 g 420 volts and 60 1 1.2 Hz. Subsequently, manually sychronize the generator,

_a gradually load to at least 1650 kW (Unit 2:

5100 kW to 5300 KW), and operate for at least 60 minutes. This test should be conducted in consideration of the r.anufacturer's recommendations regarding engine prelube, warm up, loading and shutdown procedures where possible.

l

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TS.5.1-1 5.0 DESIGN FEATURES j

5.1' SITE

+

The Prairie Island Naclear Generating Plant is locatud on property owned by Northern States Power (HSP) Company at a site on the west bank of the Mississippi River, epproximately 6 miles northwest of the city of Red Wing, Minnesota. The minimum distance from the center line of either reactor to the site exclusion boundary is 115 meters, and the low population aone distance is 1 1/2 milea. The nearest population center of 25,000 or more people is South Saint Paul. These site characteristics comply with definitions in 10CFR100 (Reference 1).

The U.S. Army Corp of Engineers controls the land within the exclusion area that la not owned by NSP. The Corps has made an agreement with NSP to prevent residential construction on this land for the life of the plant (Reference 2).

These specifications use atmospheric diffusion factors based on the NRC staff evaluations. Its evaluation of accidental airborne releaser is based on a relative concentration of 9.8 x 10** seconde per cubi: meter at the site boundary. Its evaluation of routine releases in based on a relative concentration of 1.5 x it. seconds per cubic meter (Reference 3).

I The flood of re ord in 1965 preduced a water surface elevation cf +688 feet MSL at the site. The calculated probable maximum flood (pMF) level is +703.6 seet mean sea level (PSL), and the estimated wave runup could

. See Section 2.4.2 of this report.)

Plant grade reach +706.7 feet MS?..

(

level is +695 feet MSL.

Flood protection structures have been provided. The two turbine support facilities, the common auxiliary building, and the two shield buildings have been physically connected by a concrete flood wall, most of the length of which constitt, a the concrete foundation walls for the various buildings. The top of this wall-supports the metal siding for the

-buildings at about elevation +705 feet MSL. Fourteen doors through the flood wall, or into the various buildings (including the separate screen house), are provided with receivers for the erection.of flood protection panels to prevent flood water from reaching safet,' related facilities.

The cooling cater pumps in the screenhouse are designed to operate up to i JJood level of +695 feet MSL without flood protection measures, and up to a level of +707 feet MSL with the erection of flood protection panels.

The main transformer foundation is at +695 feet MSL. The transformer will function to a flood level of+698 feet MSL.

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TO.5.1-2 l

The plant is desigt.ed for a design basis earthquake having a horizontal l

ground acceleration of 0.129 and an operational basis earthqua?.e having a horizontal ground acceleration of 0.06g.

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Be f erer.ces 1.

USAR, Section 2.2.1 2.

USAR, Section 3.4.5 3.

SER, Sections 2.3.4 and 2.3.5

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6.0 ADMINISTRATIVE CONTROLS 6.1 Responsibility l

A.

The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during i

his absence.

The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

B.

The shif t supervisor (SS) shall b9 responsible for the control room i

-command function. During any absence of the SS from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active senior reactor operator (SRO) license shall be designated to assume the control room command function. During any absence of the SS f rota the control room while the unit is-in MODE 5 or 6, an individual with an active SRO licende or reactor operator license shall be desigt.ated to assume the control room command function.

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6.2 Organization A. Onsite and Offsite Oraanirations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The oncite and offsite organizations chall include the posicions 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 relationchips, and job descriptions for key personnel-positions,-or in equivalent forms of documantation. These requirements, including the plant specific titles of those personnel fulfilling the responsibilities of the positiens delineated in these Technical Specifications, anall be documented in the Updated Safety Analysis Repc'it.

2. The plant manager shall report to the corporate vice president i

specified in 6.2.A.3, shall be responsible for overall safe operation of the plant, and shall have control over those onsite (etivities neceJeary 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 providh.g technical support to the plant to ensure nuclear safety.

4. The individuals who train the operating statt, carry ouc 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 cperating in MODES 1, 2, 3, or 4.

2. At lear 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 senior reactor operator -

shall be present in the control room.

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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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpectsd absence of on-duty shift crew members provided immediate action is taken to restore the shif t crew composition to within the minimum requirements.

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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

5. Administrative procedures shall be developed and implemented to limit i

the working hours of personnel who perform safety related functions (e.g., licensed SROs, licensed Ros, health physicists, auxiliary.

operators, and key maintenance personnel).

The procedures shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.

Any deviations from the working hour guidelines shall be authorized in advance by the Plant Manager or designee in accordance with approved administrative procedures and with documentation of the basis for granting the deviation.

controle shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or designee, to ensure that excessive hours have not been assigned. Routine deviation from the working hour guidelines shall not be authorized.

6. The operations manager or assistant operations manager shall hold an SRO license.

7. The shift technical advisor (STA) shall provide advisory technical support to the shif t supervisor in the areas of thermal hydraulics, 1

reactor engt.neering, and plant analysis with regard to the safe operation of-the unit. Personnel performing the function of the STA shall be assigned to the shift crew when a unit is in MODE 1, 2,

3, or 4.

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TS.6,0-4 l

6.3 Plaat Staff Qualifications Each member of_the plant staff shall meet or exceed the minimum 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 have a bachelors degree or equivalent in a scientific or engineering discipline with specific i

training in plant design, and response and analysis of the plant for transients and accidents, and (2) the operations manager who shall mest i

the requirements of ANSI N18.1+1971, except that NRC license requirements are as specified in Specification 6.2.B.6.

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i 6.4 Procedures Written procedures shall be established, implemented, and maintained covering the following activities:

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I A.

The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 19781 l

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B.

The emergency operating procedures required to implement the requiremento of INREG-0737 and to IMREG 0737, Supplement 1, as stated in Ceneric Letter 02+331 I

c. - Quality control for ef fluent 'and enviror. mental monitoring; f

D.

Fire protection program implementations and E.

All programs specified in specification 6.5.

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6.5 programs and Manuals The following programs shall be established, implemented and maintained.

[

A.

Offrite Dose Calculation Manual (ODCM)

The ODCM shall contain the methodology and parameters used in the 1

calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program. The ODCM shall also contain the radioactive effluent controle and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Monitoring, and Radioactive Effluent Reports required by Specification 6.6.B.and Specification 6.6.C.

i Changes to the'ODCM

1. Shall-be documented and records of reviews performed shall be retained. This documentation shall contains

a. suf ficient information to support the change (s) together with the appropriate ar.alyses or evaluations justifying the change (s),

b. a determination that the changets) maintain the levels of radioactive effluent control required by 10CFR20.1302, 40CFR190, 10CFR50.36a, and 10CFR50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose or setpoint calculations; t

2. Shall become effective after approval by a member of plant management designated by the-Plant Manager.

3. Shall be submitted to the NRC in the form of a complete legible a

copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed. The date (i.e.,

month and year) the change was implemented shall be indicated.

B.

Primary Coolant sources outside containment This program provides controls to minimize leakage from those portions of systems outside-containment that could contain_ highly radioactive fluida during a serious transient or accident to levels

.as. low as practical. The systems include portions of Residual Heat Removal,. Safety Injection, and Containment Spray Systems. The program shall include the following:

1. Preventive' maintenance and periodic visual inspection requirements, and-2.' Integrated leak test requirements for each syctem at refueling cycle intervals or-less.

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TS.6.0-7 l

C.

Post Accident Samolina i

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 i

accident conditions. The program shall include the following:

1. Training of personnels l

l 1

2. procedures for sampling and analysis; and j

3. Provisions for maintenance of sampling and analysis equipment.

D. -Radioactive Effluent Controls Procram This program conforms to 10CFR50.36a for the control of radioactive effluents and for maintaining the doses to members of tha public frem 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 released in liquid effluents to unrestricted areas, conforming to Appendix I

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 methodology and parameters in the ODCM;

4. Limitations on the annual and qusrterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents teleased from each unit to unrestricted areas, conforming to 10CFR50, Appendix I;

5. Determination of cumulative dose contributions from radioactive affluents for.the current calendar quarter and current calendar i

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;

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

i TS.6,0 8 l

l

6. Limitations on the functional capability and use of the liquid and f

gaseous effluent treatment systema 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 effl at rei.eanes would exceed 0.12 mrem to the total body or to any organs or from the gascous effluent

~-

releases would om va 0.4 mrad for gamma air dose, 0.8 mrad for beta air doce, or o.6 mrem organ doses

7. Limitations on the dose rste resulting from radioactive material released Jn gaseous effluents to areas beyond the site boundary conforming to the doso associated with Appendix B to 10CFR20.1 -

i 20.601, Table II, Column li

8. Limitations on the annual and quarterly air dc.es 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 quart?rly 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 gaseoua effluents released from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix It 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 Li3[L This program provides controls to tr&ck the USAR, Section 4.1.4 cyclic and transient occurrences to ensure that components are maintained within the des.igo limits.

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(Reserved)

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Explosive Gas and Storace Tank Radioactivity Monitorina Procram This. program provides controls tor potentially explosive gas mixtures I

contained'in the waste gas holdup system, the quantity of l

radioactivity centained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall includes

1. The limits for concentration of oxygen in the waste gas holdup system and a surveillance program to ensure the limits are maintained. Buch 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 76,800 curies of noble gases (considered as dose j

equivslent Xe 133); and.

3. A sur/eillanca program to ensure that the quantity of l

radioactivity contained in each of the following tanks shall be limited to 10 curies, excludi*.g tritium and dissolved or entrained noble gaues:

4 condensate storage tanks outside temporary tanks

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

K.

Diesel Fuel Oil Testjng,,fyp,gnm A diesci fuul' 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 accepta.;ce criteria, all in accordance with the limAto specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. Diesel fuel oil-testing shall be performed every 31 days.

L.. Ig,chnica grecifications Bases Control Program This program provides a means for processing changes to the Bases of these Technical specifications.

1.-Changes to the Basco or the Technical Specifications shall be made under' appropriate' administrative controls and reviews.

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78.6.0-10 l

L.

Technical Snecification Bases Control Procram (continued)

2. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the followings

a. a change in the Technical Specifications incorporated in the licenset or j

b. a change to the USAR or Bases that involves an unreviewed

{

safety question as defined in 10CFR50.59.

3. The Bases control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

4. Proposed changes that mnet the criteria of Specification 6.5.L.2 above shall be reviewed and approved by the NRC pr$or to l

implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

Containment Leak oe Rate Testina,pr**rgm M.

S A program shall be established to implement the leakage rate tecting of the containment as required by 10 CFR 50.54(o) a..d 10 CFR 50, Appendix d, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Performance-Based Containment Leak-Test Regulatory Guide 1.163, Program," dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure, P.,

of 46 psig.

The maximum allowable primary containment leakage rate, L,,

at P.,

shall be 0.25% of primary containment air weight per day. For pipes i

connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.1% of primary containment air weight per day at pressure P., For pipes cont.ected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone, the total leakage past isolation valves shall be less than 0.01% of primary containment air weight per day at pressure P.,

i Leakage Rate acceptance criteria are t

a.-' Primary containment leakage rate acceptance criterion is s 1.0 L.,

Prior to unit startup, following testing in accordance with the

. program, the combined leakage rate acceptance criteria are s 0.60 L for all components' aubject to Type B and. Type C tests and s 0.75 L, for-Type A tests;

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Containment Leakage Rate Testino Procram (continued) i i

b. Air lock testing acceptance criteria aret

,p Overall air lock leakage rate is s 0.05 L, when tested at a46 psig

2) For nach door intergasket test, leakage rate is s 0.01 L _ when pressurised to =10 peig.

The provisions of 4.0.A do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. The Containment Leakage Rate Testing Program stipulates acceptable extension of test intervile.

The provisions of 4.0.D (except that the allowed surveillance intervals are defined by the Containment Leakage Rate Testing Program) are applicable to the Containment Leakage Rate Testing Program.

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l 6.6 Reporting Requirements l

i The following reports shall be stomitted in accordance with 10CFR50.4

[

A.

Occuoational Exposure Report l

A tabulation on an annual basis of the nurther of station, utility and

'I othcr personnel (2ncluding contractors) receiving exposures greater than 100 mrem /"r and their associated man rem exposure according to j

work and job functions, e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste _ processing, and refueling. This i

tabulation supplements the requirements of 10CFR20.2206. The dose assignments to various duty functions may_be estimated based on pocket dosimeter, TLD, or film br.dge 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.

D.

Annual Radiolocical Environmental Monitorino 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 i

radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Of f site Dose Calculation Manual (ODCli), and in b

10CFR50, Appendir. I, Sections IV.B.2, IV.B.3, and IV.C.

The Annual Radiation Envirsnmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Guide 4.0, 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.

T'.e missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include 'he followings a cummary 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 Efflugnt 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 r

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 10CFRS% Appendix I,_Section-IV.B.1.

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TS.6,0 13 l

l D.

B2nthly Operatina Reports l

Ecutine reports of operating statistics and shutdown experience shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

E.

Core Doeratina Limits Reoort (COLR) l i

1. Core operating limits shall be established prior to each reload j

cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the followings j

nn

a. Heat Fitx Hot Channel Factor Limit (F

), Nuclear Enthalpy

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(Fa

),

PFDH, K(2) - and V(2)

Rise Hot Channel Factor Limit n

(Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10.B.9)

c. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

d. Reactor Coolant System Flow Limit (6pecification 3.10.J)

2. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documento:

USPNAD-8101-A, " Qualification of Reactor Physics Methods for Application to PI Units" '(latest approved version)

NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Unite" (latent approved version)

WCAP-9272-P A,

" Westinghouse Reload Safety Evalustion Methodology *, July, 1985 WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code", August, 1985 WCAP-10924-P A, " Westinghouse Large-Break LOCA Best-Estimate Methodology", December, 1988 WCAP-10924-P A, Volume 1, Addendum 4, " Westinghouse Large Break LOCA Best Estimate Methodology", August, 1990 XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase 7I", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Peport H-COBRA / TRAC 2-Locp Upper Plenum Injection Model Update to Support ZIRLO,

3 Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993).

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E.

Core Ongratina Limits Renort (COLRi (continued)

NSPNAD-93003-A,

Transient Power Distribution Methodology",

(latest approved version)

3. The core operating limita shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core l

thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

4. Tit COhR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

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6.7 Ifigh Radiation Area

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A.

Pursuant to 10CFR20, paragraph 20.1601(c), in lieu of the requirements of 10CFR20.1601, each high radiation area, as defined in 20CFR29, in which the intensity of radiation is greater than 100 i

mrem /hr but less than or equal to 1000 mrem /hr, shall be barricaded l

and conspicuously posted asin high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuale qualified in radiation protection procedures (e.g., health physics technicians) or personnel continuously escorted by such individuals may be exempt from the RWP

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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 th y are otherwise following plant radiation protection procedures for entry into such high radiation areas.

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

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

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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 mai be made after the dose rate levels it. 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 f requency specified by the radiation protection manager.

B.

In addition to the requirements of Specification 6.7.A above, areas j

with radiation levels greater than 1000 mrem /hr shall be provided 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 duting periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work creas 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 te made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

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6.7 High Radiation Area (continued)

C.

For individual high rad 1ation areas with radiation levels of greater than 1000 mrem /hr. accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for puposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably consructed around the individual area, that incividual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

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e B.4.4-3 4.4 COfTTAINMENT SYSTEM TESTS Elian continued Several penetrations of the containment vessel and the shield building could, in the event of leakage past their isolation valves, result in leakage being conveyed across the annulus by the penetrations themselves, thus bypassing the function of the Shield Building Ventilation System (Reference 5).

Such leakage is estimated not to exceed.025% per day.

A special zone of the auxiliary building has minimum-leakage construc-tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant trains of the Auxiliary Building Special Ventilation System.

This system, when activated, will supplant the normal ventilation and draw a r.cuum throughout the zone such that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack.

The design basis loss-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure. Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environo is 0.01% per day.

The resulting two-hour doses a*. the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of 1H miles are less than guidelines presented in 10CFR100.

Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage than the design basis. The staff has resvaluated doses for these higher inleakage rates and found that for a primary containment leak rate of 0.25% per day at peak accident pres-aure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment in-leakage (Reference 6).

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ML20198B147

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