ML20071H641
| ML20071H641 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/11/1994 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20071H635 | List: |
| References | |
| GL-86-10, NUDOCS 9407190395 | |
| Download: ML20071H641 (41) | |
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LICENSE AMENDMENT REQUESTS DATED July 11, 1994 Fire Protection Technical Snecification Chances EXHIBIT B Operating License and Technical Specification Marked Up Pages DPR-42 Page 4
DPR-60 Page 4
Appendix A, Technical-Specification Pages TS-iv TS-vi TS-x TS-xi TS-xii TS.1-3 TS.3.14-1 TS.3.14-2 TS.3.14-3 TS.3.14-4 TABLE TS.3.14-1, page 1 TABLE TS.3.14-1, page 2 TABLE TS.3.14-1, page 3 TS.4.16-1 TS.4.16-2 TS.4.16-3 TS.4.16-4 TS.6.1-2 TS.6.1-3 TS.6.2-6 B.3.14-1 B.3.14-2 B.4.16-1 B.4.16-2 i
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9407190395 940711 i PDR -ADOCK 05000292 p PDR
(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 2781'7 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
" Prairie Island Nuclear Generating Plant Physical Security Plan," with revisions submitted through November 30, 1987; " Prairie Island Nuclear Generating Plant Guard Training and Qualification Plan," with revisions submitted through February 26, 1986; and " Prairie Island Nuclear Generating Plant Safeguards Contingency Plan,"
with revisions submitted through August 20, 1980.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
(4) EJre Protection The licenccc may precced uith and ic required te complete -
the mcdification: identified in P-ar:gr:ph: 3.1.1 threugh 3.1.21 ef the *?.C'c Fire Pretcetier Safety Evaluatien, dated September 6,19'9 for the facility according te the cchedule 1: Tchle 3.1. If any mcdificatienc cannet be ce=pleted er cchedule the licenccc ch:11 cubmit : reper4 exp4 ining the circumctanecc cnd propece, fer ct:ff apprevel, e revised cchedule.
Ir additien, the licenccc ch:11 cubmit the additienal in-for= tier identified ir Scot 4cnc 3.1 nd 3.2 cf the related Safety Feeluatien ir accordance zith the cchedule contained thereir In the event thece deter fer submitt:1 cannet be be met, the licenccc ch:11 cubmit :
repert, explaining the circumetenecc, together with a reviced cchedule.
The !!cenccc !c required te devcicp and impic=ent the adminictrative centrel identified ir Sectier 5 ef the related S2f-ety Evalu: tier ..ithir fcur centhe fre- the date ef thic endment.
NorthuhT5 Eatis^sTP60sEC6iiipsbyfsh'alEisfil siisiiti?shd Mainta.iMiOffictielljphovisionsjdfshh5NppiAved[fiYs p ro te c t iLon f pr6gtsnEsside shibidis;nd ne ferdn.csd fin %thd.
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. t ivouldjn+ot Commission-onlyfiflltho.- - .segchanges adve rs ely q a f f e c tithef;abil i tp? to!;iichieveland inainta ijij[safej shutdositfinithe2sje n$ o fshj firQ D. This license is effective as of the date of issuance and shall expire at midnight August 9, 2013.
FOR THE ATOMIC ENERGY COMMISSION
/s/ Roger S Boyd A Giambusso, Deputy Director for Reactor Projects Directorate of Licensing
Attachment:
Change No. 3 to Appendices A and B Date of Issuance:
April 5, 1974 Unit 1 l
4-(3) Physical Protection ,
The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made ,
pursuant to provisions of'the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817.and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
" Prairie Island Nuclear Generating Plant Physical Security Plan," with revisions submitted through November 30, 1987; " Prairie Island Nuclear Generating Plant Guard Training and Qualification Plan," with revisions submitted through February 26, 1986; and " Prairie Island ,
Nuclear, Generating Plant Safeguards Contingency Plan,"
with revisions submitted through August 20, 1980.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
(4) Fire Protection The licenece m2y precced eith crd ic required te cerplete the mcdification: identified ir Ferngr:ph 3.1.1 through 3.1.21 ef the "RC'c Fire Prctectier S2fety Sv21untien, dcted Septerber 5, 19'9 fer the facility ccccrding te the cchedule in Tcbic 3.1. !f ny ecdificatiene cannet-he completed er cchedule the licenece ch:11 cubmit 2 report ,
explcining the circurctnnecc 2nd prepece, for ct ff apprevnl, 2 reviced cchedule.- l Ir edditien, the licenccc'cbc11 cubmit the cdditienal in i ferm2 tion identified ir Sectienc 3.1 and 3.2 ef the !
related Snfety Evclu2 tier 1: ccccrdance zith the cchedule l centcired thereir' Ir the event there deter fer cubmitt:1 cannet be be met, the licencee ch:11 cubmit repert, explcining th circurctanecc, tegether eith a rev!ced cchedule.
U+a-licenece ir required te devcicp nd !=plerent th-2deinictrative centrelc identified ir Sectier 6 cf the rel tel Scfety Evcluntic ^ eithin fcur monthe frer the date cf thic ==end=cnt.
Northe rn' S ta te's ' Pose r 'Companp'^ sh'all' impliment"and i maintainin effect all provisionsof4 th's approved' fire l
protection ' program as described and" referenced'in the'^ ;
Updated Safety Analysis Report' for the ' Prairie Island I Nuclear Generating Planth Unit's,1,and 2,:and.as'appropEd~
in Safety Ev'aluation Reports dated September 6,1979; April,4', 1980, December 29,'1980 and (,.',dat'e for this application approval to be inserted,by the NRC, . . )
, subject to,tl3 e following provision:,
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D. This license is effective as of the date of issuance and shall expire at midnight October 29, 2014.
FOR THE AT)MIC ENERGY COMMISSION
/s/ A Giambusso A Giambusso, Deputy Director for Reactor Projects Directorate of Licensing Date of Issuance:
October 29, 1974 Unit 2 l
TS-iv I REV 102 9/29/92 TABLE OF CONTENTS (Continued)
TS SECTION TITLE PACE 3.10 control Rod and Power Distribution Limits TS.3.10-1 A. Shutdown Margin TS.3.10-1 B. Power Distribution Limits TS.3.10-1 C. Quadrant Power Tilt Ratio TS.3.10-4 D. Rod Insertion Limits TS.3.10-5 E. Rod Misalignment Limitations TS.3.10-6 F. Inoperable Rod Position Indicator Channels TS.3.10-6 G. Control Rod Operability Limitations TS.3.10-7 H. Rod Drop Time TS.3.10-7 I. Monitor Inoperability Requirements TS.3.10-8 J. DNB Parameters TS.3.10-8 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 A Control Room Special Ventilation System TS.3.13-1 3.14 Deleted 3.16 Fire Detection cod Protection Sycte=c TS.3.14-1
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Fire Detection Inctrumentatien TS.3.15-1 S. Fi+a-Suppreccier "eter Sycter TS.3.15 1 C. Sprey end Sprinkler Syctem: TS.3.15-2 D. Cerbe: Dic. side Sycter TS.3.14 3 E. Fire Hece Statienc TS.3.15-3 r Yard Hydrent Here Heuccc TS.3.15 ^
C. Penetretier Fire Scrriere TS.3.15 ^
3.15 Event Monitoring Instrumentation TS.3.15-1 A. Process Monitors TS.3.15-1 B. Radiation Monitors TS.3.15-1 C. Reactor Vessel Level Instrumentation TS.3.15-2 i
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TS-vi REV 99 7/9/92 TABLE OF CONTENTS (Continued)
TS SECTION TITLE PACE 4.12 Steam Generator Tube Surveillance TS.4.12-1 1 A. Steam Generator Sample Selection and TS.4.12-1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection .
C. Inspection Frequencies TS.4.12-3 D. Acceptance Criteria TS.4.12-4 E. Reports TS.4.12-5 4.13 Snubbers TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 4.16 Deleted
'.15 Fire Detectier end Prctectier Syct .c TS.'.16 1
^ Fire Detcetier Inctr=c.tetier TS.'.15 1 S. Fire Suppreccier "eter Sycter TS.'.15 1 C. Sprey cnd Sprinkler Sycter: TS . ' .15 - 3 D. Cerber Dicy.ide Syrte- TS.'.16-3 E Fire Mere Stetien: TS.'.15-3 i F Fire Hydrent Mece Meuccc TS . ' .16 '-
C. Penetrctier Fire Scrrierc TS . ' .16 '-
4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Gaseous Effluents TS.4.17-2 ,
C. Solid Radioactive Waste TS.4.17-4 D. Dose from All Uranium Fuel Cycle Sources TS.4.17-4 4.18 Reactor Coolant Vent System Paths TS.4.18-1 A. Vent Path Operability TS.4.18-1 ;
B. System Flow Testing TS.4.18-1 ;
4.19 Auxiliary Building Crane Lifting Devices TS.4.19 J J
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TS-x REV 94 3/20/91 ,
TABLE OF CONTENTS (continued) !
TS BASES SECTION TITLE PAGE ;
f 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety. Limit, Reactor Core B.2.1-1 2.2 Safety Limit, Reactor' Coolant System Pressure B.2.2-1 2.3 Limiting Safety System Settings, Protective. B.2.3-1 i Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1
- B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 ,
D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 l 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4 1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B'.3.6 1 -
3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-l~
A. Liquid Effluents B.3.9-1 B. Gaseous Effluents B.3~.9-2 C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 ,
E. & P. Effluent Monitoring Instrumentation B.3.9-5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin .
B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 G. Control Rod Operability Limitations B.3.10-9.
H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.ll-1 3.12 Snubbers B.-3.12-1 3.13 . Control Room Air Treatment System B.3.13-1 3.14 Deleted 3.15 Fire Detection and Protection Sycterc E.3.16-1 3.15 Event Monitoring Instrumentation B.3.15-1 l
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TS-xi REV 99 7/9/92 TABLE OF CONTENTS (continued)
TS BASES SECTION TITLE PACE j 4.0 BASES FOR SURVEILLANCE REQUIREMENTS !
4.1 Operational Safety Review B.4.1-1 I 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 ;
and Valves Requirements 4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves 4.4 Containment System Tests B.4.4-1 4.5 Engineered Safety Features B.4.5-1 4.6 Periodic Testing of Emergency Power Systems B.4.6-1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Radiation Environmental Monitoring Program B.4.10-1 A. Sample Collection and Analysis B.4.10-1 B. Land Use Census B.4.10-1 C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test B.4.ll-1 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 Deleted
^ 16 FLre Detectien and Protection Syrter:
. E.^.16 1 4.17 Radioactive Effluents Surveillance B.4.17-1 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1
TS-xii REV 107 7/29/93 TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System
3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring instrumentation 3.2^-1 Screty Related Fire Detection Incevumente 3.15-1 Event Monitoring instrumentation - Process & Containment 3.15-2 Event Monitoring instrumentation - Radiation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.10-1 Radiation Environmental Monitoring Program (REMP)
Sample Collection and Analysis 4.10-2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillanco Requirements 4.17-2 Radioactive Caseous Effluent Monitoring instrumentation-Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition
TS.1-3 REV 91 10/27/89 DECREE OF INSTRUMENTATION REDUNDANCY DEGREE OF INSTRUMENTATION REDUNDANCY is defined as the difference between the number of OPERABLE channels and the minimum number of channels which when tripped will cause an automatic shutdown.
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 is that concentration of I-131 (uC1/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,
" Calculation of Distance Factors for Power and Test Reactor Sites".
E-AVERAGE DISINTEGRATION ENERGY 5 shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95%
of the total non-iodine activity in the coolant.
F4W m M ss40" "^TER cYSTEM The FIRE SUFPRESSION "ATER SYSTEM conci-stc ef * 'leter courecc; pumpc; and dictributler piping . tith acccciated ccctienclicing Icelation valvec. Such valver include yard hydrant valrec, and the firct velve chend of the ; ter flev clare device oc ecch cprinkler, hece ctandpipe, er cpray cyct - ricer.
CASEOUS RADUASTE TREATMENT SYSTEM The CASEOUS RADVASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
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- g. . .
. TS.6.1-2 I REV 105 5/4/93
- 3. At least two licensed operators shall he present in the control room during a reactor startup, a scheduled reactor shutdown, and during recovery from a reactor trip. These operators are in addition to those required for the other reactor.
- 4. An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor.
- 5. All refueling operations shall be directly supervised by a licensed Senior Reactor Operator or a Senior Reactor Operator ,
Limited to Fuel Handling who has no other concurrent respons-ibilities during this operation.
i 6 fire brignde ef et 1cact fi 2: membere ch:11 he mnintained er cite et 211 tirer.* The fire brig 2dc chall net include the ein ec=here cf the minimum chift c r et. fer c2fc chutocu cf the
.__-_._m_.,__._
6{ J. The General Superintendent Plant Operations shall be formerly licensed or hold a current license.
7] 3. At least one member of plant management holding a current Senior i Reactor Operator license'shall be assigned to the plant operations group on a long term basis (approximately two years). This individual shall not be assigned to a rotating shift.
D. Each member of the plant staff shall meet or exceed the minimum l qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the General Superintendent Radiation Protection who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Manager who shall have a bachelors degree or equivalent in a' scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Superintendent Plant Operations who shall meet the requirements of ANSI N18.1-1971, except that NRC license requirements are as specified in Specification 6.1.C.7.
The training program shall be under the direction of a designated member of Northern States Power management. j i
'l IFire Erigade ecmpecitier may be ler: th n the minimur requirement: fer I peried cf t! c net te eneced 2 'curc ir crder te cecc=redate unexpected cbc^nce of Fir Erignde :::bcrc previded irredicte cctier ic tcher te recterr the Fire Erigade te eithir the minimum requirc=entc.
l
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.I
. .TS.6.1-3 REV 105 5/4/93 l
9 E. Deleted E. ' training pregre: fer the fire brigede ch:11 Ec = intained under the directier of 2 decignated rerber cf Siertherr St te Peuer n negerent. ,
Thic pregr r chall meet the requircrent of Sectier 27 cf the "FPA Ccde 1976 .zith the except!cr cf training ccheduling. Fire brigade trcining chc21 he ccheduled ne cet fert' ir the trcining pregrn: j F. Administrative procedures shall be developed and implemented to limit the !
working hours of unit staff who perform safety-related functions; e.g.,
senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel. Procedures shall include the following provisions:
- 1. Adequate shift coverage shall be maintained without routine heavy ;
use of overtime. The objective shall be to have operating personnel work a nominal 40-hour week while the plant is operating. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary bcsis, the following guidelines shall be followed:
- a. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> i straight excluding shift turnover time. >
- b. Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in 1 any seven day period, all excluding shift turnover time. Individuals should not be required to work more than 15 consecutive days without "
two consecutive days off.
- c. A break of at least eight hours including shift turnover time should be allowed between work periods.
I
- d. Except during extended shutdown periods, the use of overtime should i be considered on an individual basis and not for the entire staff on a shift.
- e. Shift Emergency Coordinator (SEC) on-site rest time periods shall not be considered as hours worked when determining the total work time for which the above limitations apply.
j m
p
TS.6.2-6 REV 105 5/4/93
- f. Investigations of all Reportable Events and events requiring Special Reports to the Commission.
- g. Drills on emergency procedures (including plant evacuation) and adequacy of communication with offsite support groups.
- h. All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan, and the Security Plan (except as exempted in Section 6.5.F), shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two years.
Maintenance work requests and their associated procedures shall be reviewed per the requirements of Section 6.2.C.
- 1. Special reviews and investigations, as requested by the Safety Audit Committee.
- j. Review of investigative reports of unplanned releases of radioactive material to the environs.
- k. All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (ODCM).
- 1. The review of safety evaluations, when safety evaluations are required by 10 CFR Part 50, Section 50.59, for procedures or procedure changes to verify that such actions do not constitute an unreviewed safety question.
?
?m' ' t Fife 7 Pr6t6Etif6nTPi'6biss"and^Tisplss6htfdf pfd6Ed6fEs?'ddditli6Wubliii
~ ----- ffalt 93
~ -
- 5. Authority The OC shall be advisory to the Plant Manager. In the event of a disagreement between the recommendations of the OC and the Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the Vice President Nuclear Generation and the Chairman of the SAC for review.
- 6. Records Minutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the Vice President Nuclear Generation and others designated by the OC Chairman.
- 7. Procedures A written charter for the OC shall be prepared that contains:
- a. Responsibility and authority of the group
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LICENSE AMENDMENT REQUESTS DATED July 11, 1994 Fire Protection Technical Specification Chances EXHIBIT C operating License and Technical Specification Revised Pages DPR-42 Page 4
DPR-60 Page 4
Appendix A, Technical Specification Pages TS-iv TS-vi TS-x TS-xi TS-xii TS.1-3 TS.6.1-2 TS.6.1-3 TS.6.2-6 l
'I
4-(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:
" Prairie Island Nuclear Generating Plant Physical Security Plan," with revisions submitted through November 30, 1987; " Prairie Island Nuclear Generating Plant Guard Training and Qualification Plan," with revisions submitted through February 26, 1986; and " Prairie Island Nuclear Generating Plant Safeguards Contingency Plan,"
with revisions submitted through August 20, 1980.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
(4) Fire Protection Northern States Power Company shall implement and maintain in effect all provisions of the approved fire protection program as described and referenced in the Updated Safety Analysis Report for the Prairie Island Nuclear Generating Plant, Units 1 and 2, and as approved in Safety Evaluation Reports dated September 6, 1979, April 4, 1980, December 29, 1980 and (...date for this application approval to be inserted by the NRC...)
subject to the following provision:
The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
l Unit 1 ,
)
i
I (3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and j safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "
" Prairie Island Nuclear Generating Plant Physical Security Plan," with revisions submitted through November 30, 1987; " Prairie Island Nuclear Generating Plant Guard Training and Qualification Plan," with revisions submitted through February 26, 1986; and " Prairie Island Nuclear Generating Plant Safeguards Contingency Plan,"
with revisions submitted through August 20, 1980.
Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.
(4) Fire Protection Northern States Power Company shall implement and maintain in effect all provisions of the approved fire protection program as described and referenced in the Updated Safety Analysis Report for the Prairie Island Nuclear Generating Plant, Units 1 and 2, and as approved in Safety Evaluation Reports dated September 6,1979, April 4, 1980, December 29, 1980 and (...date for this application approval to be inserted by the NRC...)
subject to the following provision:
The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in che event of a fire.
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Unit 2
. TS-iv REV 102 9/29/92 TABLE OF CONTENTS (Continued)
TS SECTION TITLE PACE 3.10 Control Rod and Power Distribution Limits TS.3.10-1 A. Shutdown Margin TS.3.10-1 B. Power Distribution Limits TS.3.10-1 C. Quadrant Power Tilt Ratio TS.3.10-4 D. Rod Insertion Limits TS.3.10-5 E. Rod Misalignment Limitations TS.3.10-6 F. Inoperable Rod Position Indicator Channels TS.3.10-6 C. Control Rod Operability Limitations TS.3.10-7 H. Rod Drop Time TS.3.10-7 I. Monitor Inoperability Requirements TS.3.10-8 J. DNB Parameters TS.3.10-8 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 A Control Room Special Ventilation System TS.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation TS.3.15-1 A. Process Monitors TS.3.15-1 B. Radiation Monitors TS.3.15-1 C. Reactor Vessel Level Instrumentation TS.3.15-2 i
1 i
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. TS-vi I REV'99' 7/9/92 i
TABLE OF CONTENTS (Continued) :
1 i
TS SECTION TITLE PACE 4.12 Steam Generator Tube Surveillance TS.4.12-1 1 A. Steam Generator Sample Selection and TS.4.12-1 i Inspection l B. Steam Generator Tube Sample Selection TS.4.12-1 l and Inspection C. Inspection Frequencies TS.4.12-3 D. Acceptance Criteria TS.4.12-4 E. Reports TS.4.12-5 4.13 Snubbers TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.lS-1 4.16 Deleted 4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Gaseous Effluents TS.4.17-2 C. Solid Radioactive Waste TS.4.17-4 D. Dose from All Uranium Fuel Cycle Sources TS.4.17-4 4,18 Reactor Coolant Vent System Paths TS.4.18-1 A. Vent Path Operability TS.4.18-1 B. System Flow Testing TS.4.18-1 4.19 Auxiliary Building Crane Lifting Devices TS.4.19-1
c TS-x REV 94 3/20/91 TABLE OF CONTENTS (continued)
IS BASES SECTION TITLE PACE ,
2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit, Reactor Core B.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3,1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4-1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6 1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 B. Caseous Effluents B.3.9-2 ;
C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 E. & F. Effluent Monitoring Instrumentation B.3.9-5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10-9 F. Inoperable Rod Position Indicator Channels B.3.10-9 G. Controi Rod Operability Limitations B.3.10-9 H. Rod Drop Time B.3.10-10 l I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 l 3.12 Snubbers B.3.12-1 l 3.13 Control Room Air Treatment System B.3.13-1 l 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1 l
i O
l TS-xi REV 99 7/9/92 TABLE OF CONTENTS (continued)
TS BASES SECTION TITLE PAGE 4.0 BASES FOR SURVEILLANCE REQUIREMENTS 4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 and Valves Requirements 4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves 4.4 Containment System Tests B.4.4 1 4.5 Engineered Safety Features B.4.5-1 4.6 Periodic Testing of Emergency Power Systems B.4.6-1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Radiation Environmental Monitoring Program B.4.10-1 ,
A. Sample Collection and Analysis B.4.10-1 B. Land Use Census B.4.10-1 C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test B.4.11-1 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 Deleted 4.17 Radioactive Effluents Surveillance B.4.17 1 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1
- TS-xii REV 107 7/29/93 TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3 Instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating Conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring instrumentation 3.15-1 Event Monitoring instrumentation - Process & Containment 3.15-2 Event Monitoring instrumentation - Radiation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirenents 4.10-1 Radiation Environmental Monitoring Program (REMP)
Sample Collection and Analysis 4.10-2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval ;
4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Mc ratoring instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling..,nd Analysis Program 4.17-4 Radioactive Gaseous Waste Samplinj and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating ;
Plant (Per Unit) )
5.5-2 Anticipated Annual Release of Radioactive Nuclides in '
Caseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Crew Composition i
TS.1-3 REV 91 10/27/89 DEGREE OF INSTRUMENTATION REDUNDANCY DEGREE OF INSTRUMENTATION REDUNDANCY is defined as the difference between the number of OPERABl.E channels and the minimum number of channels which when tripped will cause an automatic shutdown.
DOSE EQUIVALENT I-131 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 .
actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,
" Calculation of Distance Factors for Power and Test Reactor Sites".
E-AVERAGE DISINTEGRATION ENERGY E shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95%
of the total non-iodine activity in the coolant.
CASEOUS RADVASTE TREATMENT S"! STEM The GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluente by collecting primary coolant system offgases from the primary system a.Td providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.
., TS.6.1-2 REV 105 5/4/93
- 3. At least two licensed operators shall he present in the control room during a reactor startup, a scheduled reactor shutdown, and during recovery from a reactor trip. These operators are in addition to those required for the other ;
reactor.
- 4. An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor.
- 5. All refueling operations shall be directly supervised by a licensed Senior Reactor Operator or a Senior Reactor Operator Limited to Fuel llandling who has no other concurrent respons-ibilities during this operation.
- 6. The General Superintendent Plant Operations shall be formerly licensed or hold a current license.
- 7. At least one member of plant management holding a current Senior Reactor Operator license shall be assigned to the plant operations group on a long term basis (approximately two years) . This individual shall not be assigned to a rotating shift.
D. Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (1) the General Superintendent Radiation Protection who shall meet or exceed the qualificationa af Regulatory Guide 1.8, September 1975, and (2) the Shift Manager who shall havn a bachelore degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Superintendent Plant Operations who shall meet the requirernents of ANSI N18.1-1971, except that NRC license requirements are as specified in Specification 6.1.C.7.
The training program shall be under the direction of a designated member of Northern States Power management.
s e TS.6.1-3 REV 105 5/4/93 E. Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,
senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel. Procedures shall include the following provisions:
- 1. Adequate shif t coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a nominal 40-haur week while the plant is operating. However, in the event that unfo.eseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:
- a. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight excluding shift turnover time.
- b. Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in any seven day period, all excluding shift turnover time. Individuals should not be required to work more than 15 consecutive days without two consecutive days off.
- c. A break of at least eight hours including shift turnover time should be allowed between work periods,
- d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.
- e. Shift Emergency Coordinator (SEC) on-site rest time periods shall not be considered as hours worked when determining the total work time for which the above limitations apply.
2- TS.6.2-6 ;
, REV 105 5/4/93
- f. Investigations of all Reportable Events and events requiring Special Reports to the Commission.
- g. Drills on emergency procedures (including plant evacuation) and adequacy of communication with offsite support groups.
- n. All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan, and the Security Plan (except as exempted in Section 6.5.F), shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two years.
Maintenance work requests and their associated procedures shall be reviewed per the requirements of Section 6.2.C.
- 1. Special reviews and investigations, as requested by the Safety Audit Committee.
- j. Review of investigative reports of unplanned releases of radioactive material to the environs.
- k. All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (ODCM).
- 1. The review of safety evaluations, when safety evaluations are required by 10 CFR Part 50, Section 50.59, for procedures or procedure changes to verify that such actions do not constitute an unreviewed safety question.
- m. Fire Protection Program and implementing procedures and the submittal of recommended changes to the Safety Audit Commmittee.
- 5. Authority The OC shall be advisory to the Plant Manager. In the event of a disagreement between the recommendations of the OC and the Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed. A written summary of the disagreement will be sent to the Vice President Nuclear Generation and the Chairman of the SAC for review.
- 6. Records Minutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the Vice President Nuclear Generation and others designated by the OC Chairman.
- 7. Procedures A written charter for the OC shall be prepared that contains:
- a. Responsibility and authority of the group I