Text

Proposed Tech Specs Re Conformance of Administrative Controls Section 6 to Guidance of Std
	ML20249A764
Person / Time
Site:	Prairie Island
Issue date:	06/11/1998
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20249A757	List: ML20249A756; ML20249A760; ML20249A764;
References
	NUDOCS 9806180200
	Download: ML20249A764 (20)
	v • d • e

-_ - _.

L TS-x l

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Reactor Core Safety Limits 2.1-1 Unit 1 :nd Unit 2 R:::t:r C lant Synter M : tup Limitati n 2.1 2 Unit 1 :nd Unit 2 R ::ter C rl nt fyrt:r C0:1d09.: Limitation:

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 Require-ments - OFA Puel 3.8-2

. Spent Fuel Pool Unrestricted Region Burnup and Decay Time Require-ments - STD Fuel 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration L

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, No GAD 5.6-4 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fucl, 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 I

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

.15.6-12

' Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD' Fuel, 16 or More GAD 9806180200 980611 PDR ADOCK 05000282 P

PDR--

J sa

TS,3.1-10

" "1 91 10/27/S4

)

3.1.D. MAXIMUM COOLANT ACTIVITY

1. The specific activity of the primary coolant.(except as spec'ified l

in '3.1.D.2 and 3 below) shall be limited to:

a. Less than or equal to' 1.0 microcuries per gram DOSE 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*Fs

a. With the specific activity of the primary coolant greater than 1.0. microcurie per gram DOSE EQUIVALENT I-131 for more than 48 i

l;>

hours 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 specific' activity of the primary coolant greater than 100/E microcurie per gratn, 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.

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

TABLE TS.4.1-2B MINIMUM FREQUENCIES 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 i...a1 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 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 j

12. Accumulator Boron Concentration Monthly

13. Spent Fuel Pit Boron Concentration Weekly thquiredLatiallstirnes".!

TS.4.4-3 b.'

Cold DOP testing shall be performed after 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.

c.

Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any structural maintenance 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 ( 10%). The results of the test shall show the air distribution is uniform within 220%.

C.

Containment Vacuum Breakers The air-operated valve in each vent line shall be tested at quarterly intervals to demonstrate that a simulated containment 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 ccch refueling chutdcen in accordance with the requirements of Specification 4.4.A.3.

i 4

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i -

TS.6.0-8 DQRadiodcEiWEfflU5nE^ Contiolf Prodiam[(c6ntin6edj]

6. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of l

radioactivity when the projected doses in a period of one month l

from the liquid effluent releases would exceed 0.12 mrem to the total body or 0.4 mrem to any organ; or from the gaseous effluent releases would exceed 0.4 mrad for gamma air dose, 0.8 mrad for beta air dose, or 0.6 mrem' organ dose; l

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

20.601, Table II, Column 1; 8.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix I;

9. Limitations on the annual and quarterly doses to a member of the public from lodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix I; and

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

E.

Component Cyclic or Transient Limit This program provides controls to track the USAR, Section 4.1.4 cyclic and transient occurrences to ensure that components are maintained within the design limits.

F.

(Reserved)

G.

(Reserved)

H.

(Reserved)

I.

(Reserved)

TS.6,0-14 E.* Core Operatina Limits Report (COLR) (continued)

WCAP-10924-P-A, Volume 1, Addendum 4, "Westingheitse 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 II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: W-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOm Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993).

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 the safety analysis are met.

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

F;1Preess e and"Teheratbre Limit"Rers t F

7 17RCS@ressuieland[ temperature}limitifoEheatupV6ooldown,ilow temperature: operation,rcriticalityRand) hydrostatic {tentingjao belllas heatup.andiccoldown ratesishall be1 established and 2

' ocumentedlin[thel PTLRjfor the;folloking Technicolt Specificatiori d

sections e J3.1[A.11.c (4) g3.1;.'A.2. c (2) V 311. A. 2.c (3);,; 34.B_.1.a1 A + 3 A 4,, L3i3]A.5,;and(Table 14.1-1C.

3.3kA 3 3

2(TheTanalvt'idallmethodsfuseditFdetermineEtheTRCS" pressure:Tand temperature; limits land Coldloverpressure' Mitigation System setpoints Tshall'belthose previously ; reviewed Land. approvsd by;the NRC,[specifiballyfthose1 described;in}the[followingdocumenti FCAPF14040{NPfA EReVisibiii2K?Hethodolo @ UseditojDdveloF Cold Overpressure. Mitigating l Systemisetpoints!and RCS :Heatup' and C6oldown! Limit]Cursesf(Includes $any,exemptionigrantedabplNRC[tB a

ASME;CodelCase N-52;41 1

3ETheIPTLR3 hall"bfpl6vided[tWtheiNRCiuponTissuanceTfor"each

"~

reactoravessel fluence lperiodfandsfor'anySrevisionior; supplement 1:heketoe: Change o Et o i thel curves j i setpoint s,1lor/ pa rame ter s fi n); the pTLRfresulting.from/newlorhaddisi'onal5 analysis offbeltline hateEial; propeities.'. wil1} bejeubmitted(tolthd (NNC' prior; to fissuarice ofjanjpdated;PTLRj f

TS.6,0-16 6.7

.High Radiation Area (continued) l 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 constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and-a flashing light shall be activated as a warning device.

B.3.1-8 EEV 10C f/21/93 3.1 REACTOR COOLANT SYSTEM Bases continued D.

Maximum Coolant Activity The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Prairie Island site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

Specification 3.1.D.2, permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than

'1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure TS.3.1-3, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure TS.3.1-3 should be minimized since the activity levels allowed by Figure TS.3.1-3 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing RCS temperature to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements in Table TS.4.1-2B provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

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l

ATTACHMENT 2 SUPPLEMENT 7 to 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 Revised Pages as proposed in this Supplement (Changes from current Technical Specifications sidelined)

TS-x TS.3.1-10 Table TS.4.1-28 TS.4.4-3 TS.6.0-8 TS.6.0-14 TS.6.0-16 B3.1-8 I

4

TS-x l

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Reactor Core Safety Limits 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 3urnup and Decay Time Require-ments - OFA Fuel 3.8-2 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Require-ments - 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.G-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 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 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 l

l

TS.3.1-10 i

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 DOSE 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 j

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 specific 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.

Next pages are Figure TS.3.1-3 and TS.3.1-12.

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

TABLE TS.4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS l

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 5 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-cecds 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 1

6.

_RCS Tritium Activity Weekly 7.

Deleted 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 Boron Concentration Weekly

Required at all times.

I

TS.4.4-3 b.-

Cold DOP testing shall be performed after 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, c.

Halogenated hydrocarbon testing shall be performed after p

each complete or partial replacement of a charcoal adsorber i

bank or after any structural maintenance 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 (r10%). 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 demonstrate that a simulated containment 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 wil'1 13 leak-tested in accordance with the requirements of Specification 4.4.A.3.

~

TS.6,0-8 D.* Radioactive Ef fluent Controls Program (continued)

6. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of one month from the liquid effluent releases would exceed 0.12 mrem to the total body or 0.4 mrem to any organ; er from the gaseous effluent releases would exceed 0.4 mrad for gamma air dose, 0.8 mrad for beta air dose, or 0.6 mrem organ dose;

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

20.601, Table II, Column 1; 8.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforning to 10CFR50, Appendix I;

9. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half-lives greater than eight days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix I; and

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

E.

Component Cyclic or Transient Limit This program provides controls to track the USAR, Section 4.1.4 cyclic and transient occurrences to ensure that components are maintained within the design limits.

F.

(Reserved)

G.

(Reserved)

H.

(Reserved)

I.

(Reserved)

1 s

/

+

.l.'.

TS.6,0-14

. E'..' Core. Ooeratina Limits Reoort (COLR) (continued)-

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 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,

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

NSPNAD-93003-A, " Transient Power Distribution Methodology",(latest approved version)

3. Tha core operating limits shall be determined such that all apelicable 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 issuance for each reload cycle to the NRC.

F. Pressure and Temocrature Limit Report

1. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following Technical Specification sections:

3.1. A. l. c (4 ), 3 1. A.2.c (2),

3.1. A. 2. c (3),

3.1.B.1.a, 3.3.A.3, 3.3.A.4, 3.3.A.5, and Table 4.1-1C.

2. The analytical methods used to determine the RCS pressure and temperature limits and Cold Overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC, specifically those described in the following document WCAP-14040-NP-A, Revision 2, " Methodology Used to Develop Cold overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Includes any exemption granted by NRC to ASME Code Case N-514)

.3. The PTLR shall be provided to the NRC upon issuance for each

-reactor. vessel fluence period and for any revision or supplement thereto. ' Changes to the' curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline

' material' properties will be submitted to the NRC prior to issuance -

of-an~ updated PTLR.

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TS.6.0-16 6.7 High Radiation Area (continued)

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 constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

1 m

B.3.1-8 3.1 REACTOR COOLANT SYSTEM Bases continued D.

Maximum Coolant Activity The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 limits fellowing a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Prairie Island site, such as SITE BOUNDARY location and meteorological conditions. were not considered in this evaluation.

Specification 3.1.D.2, permitting POWER OPERATION to continue for limited time periods with the primary coolar.t's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure TS.3.1-3, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure TS.3.1-3 should be minimized since the activity levels allowed by Figure TS.3.1-3 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.

Reducing RCS temperature to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements in Table TS.4.1-2B provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective actien.

Next page is B.3.1-10.

i

ATTACHMENT 3 SUPPLEMENT 6 LICENSE AMENDMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specifications l

l Final License Amendment Pages Note: All of the current Technical 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 TS-il Original submittal TS.6.0-3 Supplement 3 TS-v Supplement (11/25/96)

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

TS-viii Supplement 4 TS.6.0-5 Supplement 2 TS-ix Supplement 4 TS.6.0-6 Original submittal TS-x Supplement 7 TS.6.0-7 Supplement (11/25/96)

TS-xi deleted - Supplement 3 TS.6.0-8 Supplement 7 TS-xii deleted - Supplement 3 TS.6.0-9 Supplement 5 TS-xiii deleted - Original submittal TS.6.0-10 Supplement 4 TS.3.1-10 Supplement 7 TS 6,0-11 Supplement 4 TS.3.1-11 deleted - Original submittal TS.6.0-12 Supplement 4 Table Supplement 7 TS.6.0-13 Supplement 6 l

TS.4.1-2B l

(Page 1 of 2)

TS.6.0-14 Supplement 7 l

TS.4.4 3 Supplement 7 TS.6.0-15 Supplement 4 TS.4.6-1 Original submittal TS.6.0-16 Supplement 7 TS.5.1-1 Original submittal B.3.1-8 Supplement 7 TS.S.1-2 Original submittal B.3.1-9 deleted - Original submittal TS.6.0-1 Original submittal B.4.4-3 Supplement 4 TS.6,0-2 Supplement (11/25/96) l i

l Transmittal Manifest

)

Northern States Power Company j

Prairie Island Nuclear Generating Plant l

Prairie Island Nuclear Generating Plant, Units 1 and 2 Closeout of TAC NOS. M99269 and M99270 "AFW Pump ATWS Event Operability" Correspondence Date: May 26,1998 Safety Audit Committee Westinghouse Wayne Shamla RSO-8 1

N S Kury 1

A B Cutter 1

K C Midock 1

Gerard Goering 1

J A Usem 1

W C Rothert 1

C Schrock 1

Mark McKeosun RSO-8 1

J P Sorensen 1

M D Wadley RSQ-8 1

L P Matis RSQ-10 1

Extra Distribution:

Mike Johnson 1

T C Silverberg 1

J O Hill 1

K J Albrecht 1

T E Amundson 1

Rachel Kaitala (SAC) 1 NL Administrative:(copies.22+1)

Gene Eckholt 1

Jeff Kivi 1

Jack Leveille 1

Dale Vincent 1

NL File (original) 1 PI Records Mgmt 1

USAR File No Commitment No Design Change No LA Related No

Non-NSP Comments: None Manifest Date: June 12,1998 RMS Doc Type:

Prep complete:

NRC-NSP. DOC l

l

I Transmittal Manifest Northern States Power Company Prairie Island Nuclear Generating Plant P! ant Performance Review (PPR) - Prairie Island Correspondence Date: May 27,1998 Safety Audit Committee Westinghouse Wayne Shamla RSO-8 1

N S Kury 1

A B Cutter 1

K C Midock 1

Gerard Goering 1

J A Usem 1

W C Rothert 1

C Schrock 1

Mark McKeown RSO-8 1

J P Sorensen 1

M D Wadley RSQ-8 1

L P Matis RSO-10 1

Extra Distribution:

1 T C Silverberg 1

J O Hill 1

K J Albrecht 1

T E Amundson 1

Rachel Kaitala (SAC) 1 NL Administrative:ccopies-22+7)

Gene Ec'< holt 1

Jeff Kivi 1

Jack Leveille 1

Dale Vincent 1

NL File (original) 1 Pl Records Mgmt 1

USAR File No Commitment No I

l Design Change No LA Related No

Non-NSP Comments: None i

Manifest Date: June 12,1998 l

RMS Doc Type:

Prep complete:

NRC-NSP. Doc

_ _ - _ _ - _ _ _ - _ - _ _ _ _ -

ML20249A764

Text

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