ML20039B963

Amends 52 & 46 to Licenses DPR-42 & DPR-60,respectively, Limiting Conditions for Operation & Establishing Surveillance Requirements of RCS & Secondary Coolant Activities

ML20039B963
Person / Time
Site:	Prairie Island
Issue date:	12/04/1981
From:	Clark R Office of Nuclear Reactor Regulation
To:
Shared Package
ML20039B964	List: ML20039B963 ML20039B966 ML20039B969
References
NUDOCS 8112280118
Download: ML20039B963 (14)

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 52 Lict.nse No. DPR-42 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Northern States Power' Company (the licensee) date'd November 12, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is. reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii)' that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of-the publi.c; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 8112280118 811204' DR ADOCK 05000

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility' Operating License No. OpR-42 is hereby amended to read rs follows: (2) Technical Specifications t l The Technical Specifications contained in Appendices, A and B as revised through Amendment No. 52, are hereby incorporated in-the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION %2J2LL Robert A. Clark, Chief _' Operating Reactces Branch 33 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December 4, 1981 h 9 0 9 m

ps>2 ncog#o, 8 UNITED STATES E' c( ) NUCLEAR REGULATORY COMMISSION e , $ / WASHINGTON, D. C. 20555 k; +.... NORTHERN STATES POWER COMPANY j DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No. DPR-60 1. The Nuclear Regulatory Commission (the Commission) has foun,d t. hat: A. The application for amendment by Northern St'ates Power Company (the licensee) dated November 12, 1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The fccility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of. the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endar.gering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's ~ regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements' have been satisfied. l l l l. e .y ~ p y .w es s ,n-%

9 t 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 46, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications. 3. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION Robert A. Clark,-Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December 4, 1981 O i e a W I =, _. -

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO. 52 TO FACILITY OPERATING LICENSE NO. OPR-42' i AMENDMENT NO. 46 TO FACILITY OPERATING LICENSE NO. DPR-60 j DOCKET NOS. 50-282 AND 50-306 4 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by Amendment Nunber and contain vertical lines indicating the area of change. ~ Remove Pages Insert Pages 3' TS-iv TS-iv i ' TS.3.1-ll TS.3.1-11 ~ TS.3.1-12 TS.3.1-12 TS.3.1-13 TS.3.1-13 Figure TS.3.1-5 Figure TS.3.1-5 TS.3.4-2 TS.3.4-2 TS.3.4-3 TS.3.4-3 Table TS.4.1-28 Table TS.4.1-28 pages 1 and 2 e f i l i J-e F' 779

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TS-iv APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGUEES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Ther=al and Hydraulic Two Loop Operation 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 Effect of Fluence and Copper Content on Shif t of RT fr NDT Reactor Vessel Steels Exposed to 550*Te=per.ture 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function'of Full Power Service Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THER"AL PO'JER with Primary Coolant Specific Activity >1.0 pCi/ gram DOSE EQUIVALENT I-131 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step overlap with One Bottomed Rod 3.10-4 Iasertion Limits 100 Step overlap with one Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope for F = 2.21 0 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 Normalized Exposure Dependent Function BU(E ) for Exxon Nuclear Company Fuel 3.10-S V(Z) as a function of core height 4.4-1 Shield Building Design In-Leakage Rate 4.10-1 Prairie Island Nuclear Generating Plan't Radiation Environmental ~ Monitoring Program (Sample Location Map) 4.10-2 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to Oe-site Operating Organization 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group Prairie Island Unit 1 - Amendment No. 35, 44, 5 2 Prairie Island Unit 2 - Amendment No. 29, 38,4 6

. = d TS.3 1-11 D. MAXIMUM COOLANT ACiIYr]Y Specification: 1. The specific activity of the primary coolant shall be limited to: (a) Less than or equal to 10 microcuries per gram DOSE EQUIVALENT I-131, and (b) Less than or equal to 100/E microcuries per gran. 2. In Specification 3.1.D.1 the following definitions apply: 1 (a) DOSE EQUIVALENT I-131 is that concentration of I-131 (uCi/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actual.\\y present. The thyroid dose conversion f actors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites." (b) E shall be the average (weighted in proportion to the concentra-tion of each radi'onuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with hali lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant. 3 If a reactor is above hot shutdown and RCS te=perature 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 but within the allowable limit (below and to the lef t of the line) shown on Figure TS.3.1-5, operation may continue for up to 48 hours provided that the ' cumulative operating time under these cir-j cumstances does not exceed 800 hours in any consecutive 12-month period. With the total cumulative operating time at a primary coolant specific activity greater than 1.0 microcurie per gram. I DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive l 6-month period, a special report to the Commission shall be submitted within 30 days indicating the number of hours above l this limit. i (b) With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure TS.3.1-5, the affected reactor shall be shutdown and RCS temperature cooled to 500 F or less within 6 hours. (c) With the specific activity of the primary coolant greater than 100/E microcurie per gram, the affected reactor shall be shutdown and RCS temperature cooled to 5,00 F or less within 6 hours of detection. t Prairie Island Unit 1 - Amendment No. 5 2 i Prairie Island Unit 2 - Amendment No. 16 e 3 --m ,n

TS.3.1-12 4. If a reactor is at or above cold shutdown: (a) With the specific activity of the pr'imary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT l-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. A reportable occurrence re-port shall be submitted to the Commission within 30 days. This report shall contain the results of the specific activity analyses together with the following information: 1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded, 2. Fuel burnup by core region, 3. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded, 4. History of de-gassing operations, if any, starting ~48 hours prior to the first sample in which the limit was exceeded, and 5. The time duration when the specific activity of the, primary coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT I-131. _3 asis The limitations on the specific activity.of the primary coolant ensure that the resulting 2 hour 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 WRC 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 l not considered in this evaluation. Specification 3.1.D.2, permitting power operation to continue for limited t'ime periods with the primary coolant's specific activity greater than 1.0 =1crocuries/ gram DOSE EQUIVALENT I-131, but within ~ the allowable limit shown on Figure TS.3.1-5, accommodates possible iodine spiking phencmenon which may occur following changes in thermal power.- Operation with specific activity levels exceed-ing 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure TS.3.1-5 must be restricted to no more than 800 hours per year (approximately 10 percent of the unit's yearly operating time) 'since the activity levels allowed by Prairie Island Unit 1 - Amendment No. 5 2 l Prairie Island Unit 2 - Amendment No. 4 6 s l i I l

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l ' Figure TS.3.1-5 increase the 2 hours thyroid dose at the site boundary by a f actor of up to 20 following a postulated steam generator tube rupture. ' The reporting of cumulative operating. time over 500 hours in any 6 month consecutive period with greater than 1.0 microcuries/ gram DOSE EQUlVALENT I-131 vill allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800 hour limit. i i 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 primar coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used t'o assess the parameters associated with j spiking phenomena. A reduction in frequency of isotopic analyses following 1 power changes may be permissible if justified by the, data obtained. i z. I 1 l ,7 Prairie Island Unit 1 - Amendment No. 52 i Prairie Island Unit 2 - Amendment No. 4 6 1 1 L j l I.....,,-,._,,,,....m,_ a. - ~ -

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TS.3.4-2 e. For Unit 1 operation motor operated valves MV32242 and MV32243 shall have valve position monitor lights operable and shall be locked in the open position by having the motor control center supply breakers manually locked open. For Unit 2, correspond-ing valve conditions shall exist. f. Essential features including system piping, valves, and inter-locks directly associated with the above components are operable. g. Manual valves in the above systems that could (if one is i= properly positioned) reduce flow below that assumed for accident analysis shall be locked in the proper position for emergency u.e. During power operation, changes in valve position will be under direct administrative control. 3. Steam Exclusion System Both isolation dampers in each ventilation duct that penetrates rooms containing equipment required for a high energy line' rupture outside of containment shall be operable or at least'one damper in each duct shall be closed. 4. Radiochemistrv The specific activity of the secondary coolant system for that reactor shall be < 0.10 uCi/gm DOSE EQUIVALENT I-131. 3. If, during startup operation or power operatien any of the conditions of Specification 3.4.A., except as noted below for 2.a, 2.b er 4 cannot be l met,. startup operations shall be discontinued and if operability cannot be restored within 48 hours, the affected reactor shall be placed in the cold shutdown condition using normal operating procedures. With regard to Specifications 2a or 2b, if a turbine driven AFW pump is not operable, that AFW pump shall be restored to operable status within 72 hours or the aff ected reactor shall be cooled to less than 350*F within the next 12 hours. If a motor driven AFW pump is not operable, that AFW pump shall be restored to operable status within 72 hours or one unit shall be cooled to less than 350*F within the next 12 hours. If 4. is not met, the affected reactor shall be placed in hot standby within 6 hours and cold shutdown within the following 30 hours. Basis A reactor shutdown from power requires removal of decay heat. Decay heat removal requirements are normally satisfied by the steam bypass to the condenser and by continued feedwater flow to the steam generators. Normal feedwater flow to the steam generators is provided by operation of the turbine-cycle feedwater system. Prairie Island Unit 1 - Amendment No. 17, 46, 5 2 Prairie Island Unit 2 - Amendment No. 11, 40, t6 O

TS.3.4-3 The ten =ain stern safety valves have a total combined rated capability of 7,74 5,000 lbs/hr. The total full power steam flow is 7,094,000 lbs/hr; therefore, the ten main steam ggfety valves will be able to relieve the total steam flow if necessary. In the unlikely event of co=plete loss of offsite electrical power to either or both reactors, continued removal of decay heat would be ascured by avail-ability of either the stea=-driven auxiliary feedwater pu=p or the motor-driven auxiliary feedvater pump associated with each reactor, and by steam discharge to the atmosphere through the main steam safety valves. One auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from one reactor. The motor-driven auxiliary feedvater pump for each reactor can be made available to the other reactor. The minimum amount of water specified for the condensate storage tanks is sufficient to re=ove the decay heat generated by one reactor in the first 24 hours of shutdown. Essentially unlimited replenishment of the condensate storage supply is available from the intake structures through the cooling water system. The two power-operated relief valves located upstream of the =ain stea= isola-tion valves are required to re=ove decay heat and cool th followingahighenergylineruptureoutsidecontainment.yggeactordown Isolation da=pers are required in ventilation ducts that penetrate those roo=s contain-ing equip =ent needed for the accident. The li=1tations on secondary syste= specific activity ensure that the resultant off-site radiation dose will be it=ited to a s=all fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0-GPM primary to secondary tube leak in the steam generator of the affected stea= line. These values are consistent with the assu=ptions used in the accident analyses. References (1) FSAR, Section 10.4 (2) FSAR, Appendix I Prairie Island Unit 1 - A=endment No. 46, 5 s Prairie Island Unit 2 - Amendment No. 40,4 6 w .,m

~. Table TS.4.1-28 Page 1 of 2 l TABLE TS.4.1-2B '4INIMUM FREQUENCIES FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Re ference 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 Radicchemistry lI determi' ation 1/6 months (1) (when at power) n 4 RCS Isotopic Analysis for Iodine a) Once per. 4 hours, Ehenever Including I-131, I-133, and I-135 the specific activity ex-ceeds 1.0 uCi/ gram LOSE EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours following a 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. RCS Chemistry (Cl*, F*/ 02) 5/ Week 8. RCS Boron Concentration *(3) 2/ Week (4) 9.2 9. RWST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week

11. Caustic Standpipe NaOH Concentration Monthly 6.4

12. Accumulator Boron Concentration Monthly 6

13. Spent Fuel Pit Boron Concentration Monthly 9.5.5 Prairie Island Unit 1 - Amend =ent No.its, 5 2 Prairie Island Unit 2 - Amendment No.19, 4 6 9

Table TS.4.1-23 Page 2 of 2 ( TABLE TS.4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS FSAR Section TEST FREQUENCY Reference 14. Secondary Coolant Gross Beta-k'eekly Gamma activity 15. Secondary Coolant Isotopic 1/6 months (5) Analysis for DOSE EQUIVALMT I-131 concentration 16. Secondary Coolant Chemistry pH 5/ week (6) A=monia 5/ week (6) Sodium 5/ week (6) NOTES: 1. Sample to be taken after a mini =um of 2 EFPD and 20 days of power operation ~ have elapsed since reactor was last suberitical for 48 hours or longer. 2. To determine. activity of corrosion products' having a half-life greater than 30 minutes. l 3. See Specification 3.8 for requirements during refueling. 4. The' maximum interval between analyses shall not' exceed 5 days. 5. If activity of the samples is greater than 10% of the limit in Specification 3.4.A.4, the frequency shall_be once per month. i l 6. The maximum interval between analyses shal.1 not exceed 3 days. l

• See Specification 4.1.D.

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