ML20203F082
| ML20203F082 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 07/15/1986 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20203F055 | List: |
| References | |
| TAC-62137, TAC-62138, NUDOCS 8607300093 | |
| Download: ML20203F082 (96) | |
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Exhibit B Prairie Island Nuclear Generating Plant License Amendment Request Dated July 15, 1986 Proposed Changes Marked Up On Existing Technical Specification Pages Exhibit B consists of Existing Technical Specifications pages with the proposed changes written on those pages. Existing pages affected by this changes are listed below: TS-il TS.4.10-1 TS-vii Table TS.4.10-1 (Page 1 of 4) TS-viii Table TS.4.10-1 (Page 4 of 4) TS-ix Table TS.4.10-2 (Page 1 of 2) TS-x Table TS.4.10-2 (Page 2 of 2) TS.3.1-2A TS.5.1-1 TS.3.1-3 TS.5.1-2 TS.3.1-3A TS.5.2-3 TS.3.1-4 TS.S.2-4 TS.3.1-5 TS.S.3-1 TS.3.1-6 TS.S.4-1 TS.3.1-7 TS.S.5-1 TS.3.1-8 TS.S.5-2 TS.3.1-il TS.5.5-3 TS.3.1-12 TS.5.6-1 TS.3.1-13 TS.S 6-2 Table TS.3.1-1 TS.6.2-6 Table TS.3.1-2 TS.6.5-1 Figure TS.3.1-1 TS.6.5-4 Figure TS.3.1-2 TS.6.6-2 Figure TS.3.1-3 TS.6.7-2 Figure TS.3.1-4 TS.6.8-1 Figure TS.3.1-5 8607300093 860715 PDR ADOCK 05000282 p PDR r
r . TS-ti REV 73 0/23/05 TABLE OF CONTENTS (Continued) TS SEC. TION TITLE PAGE 2.0 SAFETY LIMITS AND LIMITING SAFETY SYS m SETTING TS.2.1-1 2.1 Safety Ltnit, Reactor Core TS.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure TS.2.2-1 2.3 Limiting Safety System Settings, Protective TS.2.3-1 Instrumentation A. Protective Instru=entation Settings for Reactor TS .2 .3 -1 Trip
- 3. Protective Instru=entation Settings for Reactor TS.2.3-3 Trip Interlocks C. Control Rod Withdrawal Stops TS.2.3-4 3.0 LIMITING CONDITIONS FOR OPERATION TS.3.1-1 3.1 Reactor Coolant System 15.3.1-1 A. Operational Components .
- 1. Codlant Pumps TS.3.1-1
- 2. Steam Generators TS.3.1-!A
- 3. Require =ents for Decay Heat Re= oval 3elow TS.3.1-1A 350*?
- 4. Pressurizer TS.3.1 7 -g 5.. Reactor Coolant Vent Syste= TS.3.1'5( g
, 3. Heatup and Cooldcun TS.3.1J%s C. Leakage ., TS:3.1-9 D. Maxi =us Coolant Activity TS.3.1-11 E. Max 1=um Reactor. Coolant Oxygen, Chloride TS.3.1-ti and Fluoride Concentration F. Minimum Conditions for Criticality TS.3.1-17 C. Mini =u= Conditions for RCS Te=perature Less TS.3.1-19 Than MPT H. Pri=ary Coolant System Pressure Isolation 75.3.1-21 Valver 3.2 Chemical and Volume Control System TS.3.2-1 3.3 Engineered Safety Features !S.3.3-1 A. Safety Injection and Residual Heat Removal TS.3.3-1 Syste=s
- 3. Contain=ent Cooling Syste=s TS.3.3-2 C. Component Cooling Water Syste= TS.3.3-4 D. Cooling Water System TS.3.3-5 3.4 S: cam and Power Conversion System TS.3.4-L A.1 Safety and Relief Valves TS.3.i-1 A.2 Auxiliary Feed System TS.3.1-1 A.3 Steam ExcInsion System TS.3.4-2 A.4 Radiochemistry TS.3.4-2
l ' i l 1 TS-vii REV-72 2/25/25 - TABLE OF C0EENTS P'ontinued) TS SEC ION TI"LE PACE 6.0 ADMINISTRATIVE CONTROLS TS.6.1-1
'6.1 Organization TS.6.1-1 6.2 Review and Audi: TS.6.2-1 A. Safety Audit Committee (SAC) TS.6.2-1
- 1. Membership 75.6.2-1
- 2. Qualifications TS.6.2-1
- 3. Meeting Frequency TS.6.2-2
- 4. Quorum TS.6.2-2
- 5. Responsibilities TS.6.2-2
- 6. Audi: TS . 6. 2-3
- 7. Authority TS.6.2-4
- 8. Records TS.6.2-4
- 9. Procedures TS.6.2-4 B. Operations Committee (OC) TS.6.2-5
- 1. Membership 75.6.2-5
- 2. Meeting Trequency TS.6.2-5
- 3. Quorum TS.6.2-5
- 4. Responsibilities TS.6.2-5
- 5. Authority TS.6.2-6 6.. Records .
TS.6.2-6
- 7. Procedures 75.6.2-6 6.3 Special Inspections and Audna TS.6.3-1 6.4 Safety Limi: Violation TS.6.4-1 6.5 Plant Operating Procedures TS.6.5-1 i
A. Plant Operations TS.6.5-1 B. Radiological ~75, f,6-'/ TS.6.5-1 C. Maintenance and Test TS.6.5-3 l g* ,pg, D. Process Control Progra= (PCP) TS.6.5-3
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[ .E. Offsite Dose Calculation Manual (OsCM) TS.6.5-4 pf. Te=porary Changes to Procedures TS.6.5-4 6.6 Plant Operating Records 75.6.6-1 A. Records Retained for Tive Years TS.6.6-1
- 3. Records Re:ained for the Life of the Plan: T5.6.5-1 6.7 Reporting Requirements T5.6.7-1 A. Routine Repor.s TS.6.7-1
- 1. Startup Repor: TS.6.7-1
- 2. Occupational Exposure Reor: TS.6.7-2 l 3. Monthly Operating Repor 73.6.7-2
- 4. Steam Cenerator Tube Inservice In:pection TS.6.7-2
- 5. Semiannual Radioactive Effluent Release "S.6.7-3 Repor:
- 6. Annual Summaries of Meteorological Data *S.6.7-3 l
- 7. Report of Safety and Relief Valve Failures TS.6.7-4 and Challenges B. Repor:able Events TS.6.7-4 i
l e 4
e , TS-viii REV 72 5/25/S5 TABLE OF C0!.TE!.TS (Continued) ; TS SECTION TITLE ' PAGE C. Environmental Reports TS.6.7-4
- 1. Annual Radiation Environmental !!onitoring TS.6.7-4 Reports
- 2. Environm c' A Special Reports TS.6.7-5
- 3. Other - _ronmental Reports TS.6.7-5 (non-radiological, non-aquatic)
D. Special Reports TS.6.7-5 5.3 Incire :::21 Quclific::i:: !!.5.3-1 s e e e O t l i 1 l 1 l l l l 1 I cqw -. , - - - - . , - ,c_ _ -.---
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e , TS-ixl TECHNICAL SPECIFICATIONS REV "' ' " ' "' '" LIST OF TABLES TS TA3LE TITLE 3.1-1 Unit 1 "-?ct r "----' *~';ha--- "'"' "%4"-*d4'* Ai 3.1 2 Uni; 2 " _ :;r 7;;;;l T::;h;;;; 22:: 'U;irr:di:::d) 3.5-1 Engineered Safety Features Initiation Instrument Limiting See 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 Syste=s 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instru=entation 3.9-2 Radioactive Gaseous Effluent Monitoring Instrumentation 3.14-1 Safety Related Fire Detection Instru=ents 3.15-1 Event Monitoring Instrumentation - Process & Containment 3.15-2 Event Monitoring Instrumentation - Radiation l 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Equipment Tests 4.1-23 Mini =um Frequencies for Sa=pling Tests 4.2-1 Special Inservice Inspection Requirements 4.4-1 Unic 1 and Unic 2 Penetration Designation for Leakage Tests 4.10-1 Radiation Environmental Monitoring Program (REMP) Sa=ple Collection and Analysis 4.10-2 4.10-3 REMP - Maxi =um Values for the Lower Limits of Detection REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sa=pling 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
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TS-x REV 77 i/3/06 _. _ _. APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE 2.1-1 Safety Limits, Reactor Core, Thermal 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 Systes Cooldown Limitations e_ s_ . e s,,e-s.. . _.c.,__....r.3_._..._._.
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2.1 1 : S:rri:: Lif: 3.1-1 DOSE EQUIVALENT I-131 Pri=ary Coolant Specific Activity L1=it S Versus Percent of RATED THEFF.AL POWER with the Pri=ary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Soundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents , 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Li=its 3.10-3 Insertion Limits 100 Step overlap with One 3 otto =ed Rod 3.10-4 Insertion Linits 100 Step Overlap with One Inoperable Rod 3.10-3 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thernal Power 3.10-7 V(Z) as a Function of Core Height l 4.4-1 Shield Building Design In-Leakage Race 6.1-1 NSP Corporation Organization Relationship to On-Site Operating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-Site Operating Group l l . _ .. . _ _ . . _ I f i
J o , 3 TS.3.1-14, REV 0^ 3/27/^4
- 5. Reactor Coolant Vent System
- a. A reactor shall not be made or maintained critical nor shall it be heated or maintained above 200 F unless reactor coolant vent system paths from both the reactor vessel head and pressurizer steam space are operable and closed except as specified in 3.1. A.S.b and 3.1. A.S.c below.
- b. During Startup Operation or Power Operation, any one of the following conditions of inoperability may exist for each unit until operability is restored:
- 1. Both of the parallel vent valves in the reactor vessel head vent path are inoperable.
- 2. Both of the parallel vent valves in the pressurizer vent path are inoperable.
- 3. The vent valve to the pressurizar relief tank discharge line is inoperable.
- 4. The vent valve to the containment atmospheric discharge line is inoperable.
If during Startup Operation or Power Operation any one of these conditions is not restored to an operable status within 30 days, - the reactor shall be placed in Ect Shutdown within 6 hours and in Cold Shutdown within the following 30 hours.
- c. With no reactor coolant vent system path operable, restore at lease one vent path to operable status within 72 hours or be in the Hot Shutdown condition within 6 hours and the Cold Shutdown condition within the following 30 hours.
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e . Y TS.3.1-1 r' REV 01 2/2/00 Basis When the boron concentration of the reactor coolant system is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the primary system volume in approximately one-half hour.
" Steam Generator Tube Surveillance", Technical Specification 4.12, identifies steam generator tube imperfections having a depth 50% of the 0.050-inch tube wall thickness as being unacceptable for power operation. The results of steam generator burst and tube collapse tests submitted to the staff have demonstrated that tubes having a wall thickness greater than 0.025-inch have adequate margins of safety against failurefuetoloadsimposedbynormal plant operation and design basis accidents.
Part A of the specification requires that both reactor coolant pumps be operat-ing when the reactor is critical to provide core cooling in the event that a loss of flow occurs. In the event of the worst credible coolant flow loss (loss of both pumps from 100% power) the minimum calculated DN3R remains well above 1.30. Therefore, cladding damage and release of fission products to the reactor coolant will not occur. Critical operation, except for low power ~ physics tests, with less chan two pu=ps is not planned. Above 10% power, an automatic reactor trip will occur il flow from either pu=p is lost. Below 10% power, a shutdown under administrative control will be made if flow from either pump is lost. The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients. Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point. Below 350*? and 450 psig in the reactor coolant system, the residual heat recoval system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated'at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequata defense against over pressurization of the reactor coolant system for reactor coolant temperatures less than 350*F. The combined capacity of both safety valve greater than the maximum surge rate resulting from complete loss of load.{ is
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TS.3.1-% *' REV 0; 3/27/o4 Basis (continued) The requirement that two groups of pressurizer heaters be operable provides assurance that at least one group will be available during a loss of offsite power to maintain natural circulation. Backup heater group "A" is normally supplied by one safeguards bus. Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus. Tests have confirmed the ability of either group to maintain natural circulation conditions. The pressurizer power operated relief valves (PORVs) operate to relieve reactor coolant system pressure below the setting of the pressurizar code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The PORVs are pneumatic valves operated by instrument air. They fail closed on loss of air or loss of power to their DC solenoid valves. The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses. The Specifications require that at least two methods of removing decay heat are available for each reactor. Above 350*F, both steam generators must be oper-able to serve this function. Below 350*F, either a steam generator.or a residual heat removal loop are capable of removing decay heat and any combina-tion of two loops is specified. If redundant means are not available, the reactor is placed in the cold shutdown condition. The reactor coolant vent system is provided to exhaust noncondensible gases
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from the reactor coolant systen that could inhibit natural circulation core , cooling. The operability of at least one vent path from both the reactor vessel head and pressurizer steam space ensures the capability exists to perform this function. The vent path from the reactor vessel head and the vent path from the pressurizer each contain two independently emergency powered, energi:e to open, valves in parallel and connect to a common header that discharges either to the contain-ment atmosphere or to the pressurizer relief tank. The lines to the contain=ent atmosphere and pressuri=er relief tank each contain an independently e=ergency powered, energize to open, isolation valve. This redundancy provides protec-tion from the failure of a single vent path valve rendering an entire vent path inoperable. An inoperable vent path valve is defined as a valve which cannot be opened or whose position is unknown. i A flow restriction orifice in each vent path limits the flow from an inadvettent actuation of the vent system to less than the flow of the reactor coolant =aseup system. References 1 FSAR, Section 14.1.9.
- 2. Testi=ony by J. Knight in the Prairie Island Public Hearing on January 23, 1975.
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TS . J .1-\ F l REV 21 11/1/70 B. HEATUP AND COOLDOWN Specification:
- 1. The Unit 1 and Unit 2 reactor coolant temperature and pressure and system heatup and cooldown
. rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS.3.1-1 and TS.3.1-2 fer th; dirst fm11 pre:r ::rvie pcried.
- a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.
- b. Figures TS.3.1-1 and TS.3.1-2 define limits to assure prevention of non-ductile failure only. For normal operation other inherent plant characteristics, e.g.,
pump heat addition and pressurizer heater capacity may limit the hqatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2. The limit lines shown in Figures TS.3.1-1, TS.3.1-2 shall be recalculated periodically using methods discussed in the Basis section.
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 700F.
t 4. The pressurizer heatup rate shall not exceed 1000F/hr and the pressurizer 00cidown rate shall not exceed 2000F/hr. The spray shall not be used if the te=perature difference between the pressurizer and the spray fluid is greater than 320"F. l Oasis
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Fracture Toughness Properties he fracture toughness properties of the ferritic m terials in the reactor coolant pressure boundary ar determined in accordance with Section III of the AS Boiler and Pressure vessel Code, 1972 Summer Adden a, Reference (1). Heatup and cooldown limit curves re calculated using the most limiting value of RT NDT determined as follows:
- 1. Det mine the highest RT of the material in th core region of th!DIeactor vessel using original values from Tables TS.3.1-1 and TS.31-2 and estimati g the radiation induced ART using Fig re TS.3.1-3. FastneutronkbT> l Mev) fluence at the 1/4 T and 3/4 T vessel locations a given as a function of full power service life in Figure TS.3.1-4.
- 2. Examine the dat for all other ferritic materials in the eactor coolant pressure boundary to assure that no other component
, will be limiting.
- 3. Initially, the effect f radiation of the -
RTNDT of the reactor ve el core region material is estimated usi g the curves shown in Figure TS.3.1-3. The a TNDT shown for the first full power servic period is factored into the heatup and cooldown urves provided. Values of ART NDT determined in this manner may be used until the results f om the material surveillance program, when evalu ed according i to ASTM E185, indicate that they a e inappropriate. At this time, the heatup and cooldo n curves must be recalculated. The length of the first full power service peri d has been chosen such that the limiting RTNDT at the 1/4 T vessel location nas a radiation induced ART on the or r of 100-1500F. The assumption of a rad!SIion induced hift of this magnitude assures that all other compcaents 'n the primary pressure boundary will be operated con-servatively, in accordance with code recommendations 1 (Reference (2)). e. See Alext fay / for veu; 7tr
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t \ tw 9@ rs*3.I~ [ 7 Bases Pressure / Temperature Limits Appendix G of 10 CFR Part 50, and the ASME Code require that the reactor coolant pressure boundary be designed with sufficient margin to insure that, when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a nonbrittle manner, the probability of rapidly propagating fracture is =inimited and the design reflects the uncertainties in determining the effects of irradiation on material properties. Figures TS.3.1.-1 and 2 have been developed (Reference 1) in accordance with these regulations. The curves are based on the properties of the most limiring material in either unit's reactor vessel (Unit I reactor vessel weld V-3) and are effective to 15 EFPY. The curves have been adjusted for possible errors in the pressure and-temperature sensing instruments. The curves define a region where brittle fracture will not occur and are determined from the material characteristics, irradiation effects, pressure stresses and stresses due to thermal gradients across the vessel wall. Heatup Curves During heatup, the thermal gradients in the reactor vessel wall produce thermal. stresses which vary from compressive at the inner wall to tensile at the outer wall. At the inner wall of the vessel, the thermal induced compressive. stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no ther=al stresses) represents a lower bound of all similar curves for finite heacup rates when the inner wall of the vessel is treated as the governing location. The heacup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The ther=al gradients established during heatup produce tensile stresses at the outer wall of the vessel. These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at che, outer wall of the vessel are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined. For the cases in which the outer wall of
-the vessel becomes the stress controlling location, each heatup rate of interest must be analyted on an individual basis. The heatup li=it curve is a composite curve prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heacup rate up
, to 60*F per hour. 1 In sert Fo v-vs. s. t, 7 rscw Pa5e
0- O TS.3.1'(d Heatup and Cooldown Limit Curves lowable pressure-temperature relationships for various h tup and cooldown rates are calculated using methods de 'ved from Non Mandatory Appendix G2000 in Section III of e'ASME Boiler and Pressure Vessel Code; and discu sed in detail in Reference (2). The app ach specifies that the allowable total stress intensit factor (Ky) at any time during heatup or cooldown c nnot be greater than that shown on the KIR curve ( eference 1) for the metal temperature at that time, rthermore, the approach applies explicit safety factor of 2.0 and 1.25* on stress intensity factors induce by pressure and thermal gradients , respectively. us, the governing equation for the heatup-cooldown alysis is:
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2K Im 1.2 kit 1KIR II) where: KIm is the stress 'ntensity factor caused by membrane (p essure) stress , kit is the stress int nsity factor caused by the thermal gra 'ents K IR is provided by the cc e as a function of temperature relative t the RTNDT Of the material. During the heatup analysis, Equation 1) is evaluated for two distinct situations. First, allowable pressure-temperature re ationships are developed for steady state (i.e., er rate of change of temperature) conditions assuming e presence of the code reference 1/4 T deep flaw at the ID of the pressure vessel. Due to the fact that, d ring i i heatup, the thermal gradients in the vessel wa tend to produce compressive stresses at the 1/4 T lo tion, the tensile stresses induced by internal pressure are
- The 1.25 safety factor on K 7. represents additional conservatism above Code requirements.
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Cooldown Curves During cooldown, the ther=al gradients in the reactor vessel wall produce thermal stresses which vary from tensile at the inner wall to compressive at the outer wall. The thermal induced tensile stresses at the inner wall are additive to the pressure induced tensile stresses which are already present. Therefore, the controlling location is always the inside wall. The cooldown limit curves were prepared utilizing the same type of analysis used to calculate the heatup curve except that the controlling location is always the inside wall. Criticality Limits Appendix G of 10 CFR Part 50 requires that for a given pressure, the reactor must not be made critical unless the temperature of the reactor vessel is 40*F above the minimum permissible temperature specified on the heacup curve and above the minimum permissible temperature for the inservice hydrostatic pressure test. For Prairie Island the curves were prepared, requiring that criticality =ust occur above the maxi =um permissible temperature for the inservice hydrostatic pressure test. ASME Code Section XI Inservice Test Limits The pressure temperature limits for the ASME Code Section XI Inservice Test Limits (hydrostatic pressure test) are less restrictive than the heatup and cooldown curves to allow for the periodic inservice hydrostatic test. These linits are allowed to be less restrictive because the hydrostatic test is based on a 1.5 safety factor versus the 2.0 safety factor built into the heacup i and cooldown curves and because the test is run at a constant temperature so the thermal stresses in the vessel are mini =al. f l i. , \
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- 1. USAR, Section 4.2
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- .3.1-7 omewhat alleviated. Thus, a pressure-temperature c rve based on steady state conditions (i.e., no th 1 stresses) represents a lower bound of all si ar curves for finite heatup rates when the 1/4 T loca 'on is treated as the governing factor.
The sec nd portion of the heatup analysis concerns the calc ation of pressure temperature limitations for the c e in which the 3/4 T location becomes the contro ing factor. Unlike the situation at the 1/4 T locati n, at the 3/4 T position (i.e., the tip of the 1/4 T eep O.D. flaw) the thermal gradients established du.'ng heatup produce stresses which are tensile in ratu ; and, thus, tend to reinforce the pressure stresses present. These thermal stresses are, of course, depende. t on both the rate of heatup and the time (or water temp *ature) along the heatup ramp. Furthermore, since t. thermal stresses as 3/4 T are tensile and increase 'th increasing heatup rate, a lower bound curve simil r to that described in the preceding paragraph cann t be defined. Rather, each heatup rate of interest m t be analyzed on an indi-vidual basis. Following the generation of p essure-temperature curves for both the steady state and inite heatup rate situations, the final limit curves are prod ced in the following fashion. First, a composite cur is constructed based on a point by point comparison of he steady state and finite heatup rate data. At any gi en temperature, the allowable pressure is taken to be th lesser of the two values ' taken from the curves under co ideration. The composite curve is than adjusted to al. w for possible errors in the pressure and temperature s nsing insttrments . The use of the composite curve becomes man tory in setting heatup limitations because it is po ible for conditions to exist such that over the cours of the heatup ramp the controlling analysis switches rom the O.D. to the I . D . location; and the pressure lim t must, at all times, be based on the most conservative se. The cooldown analysis proceeds in the same fashion s that for heatup, with the exception that the control ing location is always at 1/4 T. The thermal gradients induced during cooldown tend to produce tensile stress i l l 1 l J
TC.C.1 0 e t the 1/4 T location and compressive stresses at the 3 4 T position. Thus, the ID flaw is clearly the wo t case. As i the case of heatup, allowable pressure temperature relati ns are generated for both steady state and finite cooldo rate situations. Composite curves are then constru ed for each cooldown rate of interest. Again adjustmen s are made to account for pressure and temperatur instrumentation error. The use of t. composite curve in the cooldown analysis is necessary cause system control is based on a measurement of eactor coolant temperature, whereas the limiting pr sure is calculated using the material temperature at tip of the assumed reference flaw. During cooldown, e 1/4 T vessel location is at a higher temperature han the fhdd adjacent to the vessel I.D. This condition is, of course, not true for the steady-state situatic during cocidown result It follows that the aT induced in a calculated higher allowable KIR for finite cooldown atqs than for steady state under certain conditions .
. Because operation control s on coolant temperature, and cooldown rate may vary duri. the cooldown transient, the limit curves shown in Figure S.3.1-2 represent a composite curve consisting of the more for steady state and the speci. nservative values calculated c cooling rate shown.
Details for these calculations ar provided in Reference (2). Pressurizer Limits Although the pressurizer operates at te .perature ranges above those for which there is reason fo concern about brittle fracture, operating limits are pr vided to assure compatability of operation with the fatigue analysis performed in accordance with Code requireme. ts. y References (1) ASME Boiler and Pressure Vessel Code, Section III, 1972 Summer Addenda, Appendix G. (2) WCAP-7924. W.S. Hazelten, " Basis for Heatup and Cooldown Limit Curves," WCAP-7924, June 1972.
-. TS.3.1-11 REV % M/4/01 D. MAXIMUM COOLANT ACTIVITY Specification:
- 1. The specific activity of the primary coolant 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.
- 2. In specification 3.1.D.1 the following definitions apply:
(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 isetopic 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."
(b) E shall be the average (weighted in proportion to the concentra-tion of each radionuclide 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 half lives greater than 15 minutes, making up at le'ast 95% of the total' non-iodine activity in the coolant. -
- 3. If a reactor is above hot shutdown and RCS 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 but within the allowable limit (below and to the left of the line) shown on , Figure TS.3.1-f, operation may continue for up to 48 hours l provided that the cumulative operating time under these cir-l cumstances es not exceed 800 hours in any consecutive 12-month period, th the total cumulative operating time at a primary l co specific activity greater than 1.0 microcurie per gram N E EQUIVALENT I-131 exceeding 500 hours in any consecutive 0 6-mon eriod, a special report to the Commission shall be ( submitted hin 30 days indicating the number of hours above j this limit . l l (b) With the specific activt of the primary coolant greater than l 1.0 mierecurie per gram DOS QUIVALENT I-131 for more than 48 l hours during one continuous tim 'nterval or exceeding the j limit line shown on Figure TS.3.1-f, 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 c.o 500 F or less within 6 hours of detection. l l e 1
T: ' TS.3 1-12 REV 7: 5/23/05-
- 4. If a reactor is at or above cold shutdown:
(a) 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. A special report shall be submitted to the Commission within 30 days. This report shall centain 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 coolanc exceeded 1.0 microcurie per gram DOSE EQUIVALENT I-131.
Basis 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 race 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 1 Island site, such as site boundary location and meteorological conditions, were I 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 P Figure TS.3 1-J, accommodates possible iodine spiking phenomenon which may occur following p dEges in thermal power. Operation with' specific activity levels exceed-ing 10}microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure O3 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
i l l I f' TS.3.1-13
' REV 5 2 12 /'- / St Figure TS.3.1 increase the 2 hours thyroid dose at the site boundary by a factor 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 EQUIVALENT I-131 will allow suf ficient time for Commission evaluation of the circumstances prior to reaching the 800 hour limit.
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 requiremants 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. Information obtained on iodine spiking will be used to assess the parameters associated with' spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. l l t
?
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i i ! Table TS 3.1-1 IINIT NO. 1 REACTOR VESSEL TOUGitNESS DATA (UNIRRADIATED) ! TeanseerseI *I 50 ft ib/35 mits Material Cu P NDTT Lateral Espansion RT NDT Component Averase TransverseI *I (%) (%) ( Fl Temp. ( F1 (*F ) Upper Shelf (ft Ib) Closuse llead Dome A533 Gr. B, ' I 4 64 ICI P 4ICI 75 ICI ! llead Flange i A508 Cl. 3 4 , 12I l 4 bI 84 bI Vessel Flange A508 Cl. 3 4t ICI Id 4 4l I 77.5 Injection Norsies A508 C1. 3 II4Dl
-2 22El 97 bI intet and Outlet Norrie A508 Cl. 3 +5 395I 5 f'I 92 lCI 39 b Upper Shell A508 Cl. 3 4 4f?} 85 ICI Inter. Sheillbl A508 Cl. 3 0 06 0 013 *I4 14 14 143
) Lower Sheilltal A508 Cl. 3 0 07 0 014 4 45 4 134 i i Trans. Ring A508 Cl 3 +5 63 III 5 79 l Bottom flead WY A533 Gr. B, Cl. I 4 57I 'l 3 68.5 I'?I E a - Weldmen IDI Weld 0.13 0 017 0 10 0 78 5 [ flAZ lbl llAZ 0 , .i <,100 0 211 av NP q e. Sgancemen oriented nosmal to the meine wushme direct.on Eb w m b Bened on octuel evensverse date through the nueveellerwe proyam j c I .-mer s us.m
- Persiuse Temocratus e timets." Section 5 3 2 of Srsndard Reveaw Plan, NUFIEG 75/08 7, l'. 75, f>om sons tu-Jenal date i
l 1
.4 TABLE 'IE. 3.1-2 UNIT NO. 2 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED)
\ TransverseI *I 50 ft Ih/35 mds Meterial Cu P NDTT Lateral Espansion M NOT Average TransverseI *I Cornponent N Type (%) (%) Temp. (*F) (*F ) (*F) Upper Shelf (ft ab) l Closure flead Dome 5 52lCI 5 A533 Gr. C1. I 64ICI flead Flange 31 18 A508 Cl. 3 -31 87 ICI Vessel Flange A508 Cl. 3 22 18lc) 22 88 ICI Injecteon Nontes A508 Cf.3 -22 -114 ICI 22 97(c) I intet and Outlet Norrie A508 Cl. 3 l 50l ci 10 89 ICI Upper St. ell A508 Cl. 3 '13 41[c] 13 85ICI inter. Sheillbl A508 Cf. 3 0.075 0.010 -4 SG 4 l 112 Lower Shellib) A508 C1. 3 0.085 0.011 13 54 108 Trans. fling A508 Cl. 3 10 50 10 76lc) Bottom llead A533 Gr. B Cl.1 13 56 4 ICI 3 Weldmentlbl Weld 0.082 0.019 -31 -G 31 103 u Q IfAZ ibl lIAZ 31 35 31 117 ~ [
- e. Spec men ossensed noemet to the mopoe vwoekong desect.on CY
- b. Base <f on octual transvoese date through she nueveellance program [
y f0 c I .ws J e.s.aq Pressuee. Tem s eeaiues L.m.is." Section 5.3 2 of Srandard Review Pisa NUR EG 75/087, i' 16. 9 or s Hn n qu+ net date
VIGtJHE ni. 3.1-1 IJfilT 1 AtlD tJf4IT 2 RFACTOR COOLAfff SYSTI21 IIEATtJP LINITATIOft; ( Applicable for First,10 EFPY of Opernt.lon) 2500
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FIG'EI TS. 3 1 3 REV 31 11/1/78 400 C. ' COPPE4 8ASE. 0.25% wtLD An3 OVER 0.251 COPPER SASE. 0.205 wELO 0.20% COPPE' BASE 0.15% wtLD 200 - l 100 i C o 80 - [ 60 E g
- 0.15% COPPER B E. 0.10?. WELD
- 40 -
0.io: c0PPEt BASE. 0.05% wELO 20 - I I I I ' ' ! ! 10 10 88 2 4 6 8 10 2 4 8 10 20 FLUENCE (H/CH2> I NEV) EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RT NOT FOR REACTOR VESSEL STEELS EXPOSED TO 550*F TEMPERATURE l FIGURE':s.3.1-3
FIGUF2 TS.3 1-h RIV 31 11/1/78 l 3
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10 8 - 0 - slut 4 - c. 8 - 7_ g 2 - 5 5 m b lois _ b= 8 - 6 - 4 l 2 - 87 10 0 5 10 15 20 25 30 SERVICE LIFE (EFFECTIVE FULL POWER YEARS) FAST NEUTRON FLUENCE (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE FIGURE TS .3.1-4
Figura TS, o.3,.1,-g 3
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0 l { 20 30 40 50 60 70 80 90 100 i l PERCENT OF RATED THERMAL POWER 1 FIGURE 3.1- i . j DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus ! l Percent of RATED THERMAL POWER with the Primary Coolant Specific j Activity >1.0gCi/ gram Dose Equivalent 1131 1 ( l i
.'" N TS.4.10-1
! REV -10 10/21/n 4.10 RADIATION ENVIRONMENTAL MONITORING PROGRAM Apolicability Applies at all times to the periodic monitoring and recording of radioactive I ef fluents found in the plant environs.
Objective To provide for measurement of radiation levels and radioactivity in the site environs on a continuing basis. Soecification j A. Sample Collection and Analysis
- 1. The Radiation Environmental Monitoring Program described in Table 4.10-1 shall
' be conducted. Radioanalysis shall be conducted meeting the requirements of 4 Table TS.4.10-2. A map and a table identifying the locations of the sampling
- shall be provided in the Offsite Dose Calculation Manual (ODCM).
- 2. Whenever the Radiation Environmental Monitoring Program is not being conducted as specified in Table TS.4.10-1 the Annual Radiation Environmental
',, ss' Monitoring Report shall include a description of the reasons for, not, conducting the program as required and plans for preventing a recurrence.
jI e
- 3. Deviations are permitted from the required sampling schedule if sanples are unobtainable due to hazardous conditions, seasonable unavailability, j or to malfunction of automatic sampling equipment. If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period, a 4. With the level of radioactivity in an environmental sampling medium .. ,
i exceeding the reporting levels of Table 4.10-3 when averaged over j any calendar quarter, in lieu of any other report, prepare and submit to the Commission within 30 days from the end of the af fected calendar , quarter a Report pursuant to Specification 6.7.C.2(a). When more than one of' the radionuclides in Table 0.12-0 are detected in the sampling l < medium, this report shall be submitted ib.
' . 9./04
! concentration (1) , concentration (2) , ...> L.0 ' - * ' limit level (1) limit level (2) When radionuclides other than those in Table ' .10-2 are detected and are the result of plant effluents, this report shall be subnitted if i the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.9. A.2, 3.9.B.2, or 3.9.3.3.
. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be recorted and described in the Annual Radiological Environmental Monitoring Report.
6 { 'w. / b
'N
{ / T. TS.4.10-1 - (Page 1 of 4) PRAIRIE ISI.AND HlfCLEAR CENERATINC PLAttr RADIATI0tl ENVIRON!!EtiTAL HONITORING PROGRAM SAMPLE COLI.ECTION AND ANALYSIS Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample I.ocat ions ** Collection Frequency of Analysis
- 1. AIRIl0RNE Samples from 5 locations: Continuous Sampler Radiolodine analysis Radiolodine and 3 samples from offsite operation with weekly for I-131 Particulates locations (in different sample collection sectors) of the highest weekly Particulates calculated annual average Cross beta activity on ground level D/Q, each filter weekly *.
I sample from the vicinity Analyses shall be per-of a community having the formed more than 24 hours highest calculated annual fo!!owing filter change. average ground-level D/Q, and Perform gamma isotopic I sample from a. control loca- analysis on composite L io n S 2 0 ;;;! ! : ;!44tenee -"'
g
. (by location) sample
-4e the le as t-pr.av e la a r ut n f Sf e * ' quarterly. .
O OM Msession
- 2. DIRECT 32 TI.D stations established Quarterly Comma dose RADIATlotl with duplicate dosimeters quarterly placed at the following locattons:
- 1. Using the 16 meteoro- MH logical wind sectors 4 $.
as guidelines, an inner u7 g ring of stations in the . o ." general area of the site R .8' houndary is established I$H and an outer ring of U stations in the 4 to 5 mile . distance from the plant g site is estahlinhed. Because ee of inaccessibility, seven .
,' e sectors in the inner and -
e outer rings are not covered o
- . D If Cross beta activity in any indicator sample exceeds 10 times the yearly average of the control saisple, a garmaa isotopic analyslu is required. ,
- Sample locations are given on the figure and table in the ODCif.
s
s, T .
. . ,e *
~~.'
TAlli.E TS.4.10-1 , (Page 4 of 4) .* PRAIBIE ISt.AllD fillCI. EAR GEllERATING Pl.ANI' RADI ATION EllVIIt0NilEllTAI. MONITORIffG PROGRAtt SAflPl.E COI.I.ECTI0tl AND ANAI.YSIS 4 fluinher of Samples Exposure Pathway and Sampling and and/or Sample Sample I.ocations** Type and Frequency Collection Frequency of Analysis
- c. Food P oductu One sample of corn At time of harvest f rom ' ' "' my , nh? Camma isotopic analysis of edthle portion of
-f=: ne 2 2=;: l e each sample
- f. In 20 a44ui.
One sample of broad At time of harvest I-131 analysis of edihte leaf vegetation from highest D/Q garden portion of each sample and one sample from 10-20 mileg. .
, pddJ s hau.Lf AMVP amAs & .bn W *" gg u nH
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- a
** UH Sample locations are given on the figure and tahic in the ODCH. 9 N
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( ' ' i TABI.E TS.4.10-2 TABI.E NOTATION a - The I.I.D is the smallest concentration of radioactive saaterial in a sample that will be detected with 95% probability with 5% probability of f alsely concluding that a blank observation represents a "real" signal, 3 For a particular measurement system (uhich may include radiochemical separation): 4.66s ' LLD = F. . V . 2.2 2 . Y . exp( .\ 4 t ) Where: I.I.n is the apriori lower limit of detection as defined above (as picoeurie per unit mass or volume), s,' la the standard deviation of the background counting rate or of the counting rate of a blank sample In calculating the LLD for a radionuclide determined by gamma-ray as appropriate (as counts per minute). spectrometry, the background shall include t.he typical contributing of other radionuclides normally present inthesamples(e.g., potassium-40j.nmilksamples). Typical values of E. V, Y and at shall be used in the calculations. E is the counting efficiency (as counts per transformation), 2.22 is the number of transformation per minute per picocurie, {N Y is the fractional radiochemical yield (when applicable), O0 , d l \ la the radioactive decay constant for the particular radinnuclide, and G 's.
' UL l
At is the elapsed time between sample collection (or end of the sample collection period) and time of C? M j counting. . a se h - 1.I.0 for drinking water. e i c - Total for parent and daughter * " i d----Appl 4ete is n;;c;tel-lee &*pe-**alfel ::t 10 ;;- 0;:::44wo-analy :: c - Other-peake-whleh-are ,c:eurable : d !d :tif!:ble, te;; ether eith the ::d!:::: lid:: != T:51: 2.10-2 ch:41 - 0. j p j -he-ident444ed-end r e;r r t ed . v b [ O k O AC e . 4
.5.1-1 .
CV e
.3 5.0 DESIGN FEATURES 5.1 SITE The Prairie Island Nuclear Generating Plant is located on property owned by Northern States Power (NSP) Company at a site on the west bank of the Mississippi River, approximately 6 miles northwest of the city of Red Wing, Minne-sota. The minimum distance from the center line of either reactor to the site exclusion boundary is 715 meters, and the low population zone distance is 1-1/2 miles. The nearest population center of 25,000 or more pecple is South Saint Paul. These site characteristics comply with definitions in 10CFR100. 4 gg g The U.S. Army Corps of Engineers controls the land within the exclusion area that is 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. M z)}
s These specificadons usp atmospheric diffusion
> g/gg factors based on the4WNP staff evaluations. Its evaluation of accidental airborne releases is based on a relative concentration of 9.8 x 10-4 seconds per cubic meter at the site boundary.
Its evaluation of routine releases is based on ' a relative concentration of 1.5 x 10-5 seconds per cubic meter +M ,l The flood of record in 1965 produced a water surface elevation of +688 feet MSL at the site. The calculated probable maximum flood (PMF) level is +703.6 feet mean sea level (MSL), and the estimated wave runup could reach +706.7 feet MSL. (See Section 2.4.2 of this report,) Plant grade level is +695 feet MSL. Flood protection scructures 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 constitutes the concrete foundation walls for the various buildings. The top of this wall supports the i l l
m T .1-2 RGV (, 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 safety related f acilities . The cooling water pumps in the screenhouse are designed to operate up to a flood 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. The Technical Specification 6.5 A.7. requires an emergency procedure that will necessitate plant shutdown for flood water levels above +692 feet MSL at the plant site. The emergency pro-cedure will assure the proper erection of flood protection panels and assure an orderly shutdown -
, of.the plant and protection of safety related
,'} facilities. This procedure will provide for
/ progressive action levels to prev &nt the possibility of unsafe plant operation and will include requirements for periodic drills to test flood protection measures, such as erecting flood protection panels.
The plant is designed for a design basis earth - quake having a horizontal ground acceleration of 0.12g and an operational basis earthquake having a horizontal ground acceleration of 0.06g. An emergency procedure will be prepared in accordance with Specification 6.5 A.7 to define actions required for earthquak s, including plant shutdown and inspection if an o;erational basis earthquake is measured at the s Lte. References
. PSAR, Secti .
(2) FSA
- ion 12.7 (3 _R, Section 4 and 2.3.5
() sag
- SecNov 2' O '
\n /-
3'4 g 2- assiA' .Te80" -
.3, SEf , Sec gN A'3* 9 J z. 3. 6
TS .2-3 - RE/ annulus through certain penetrations in the event of leakage in their isolation valves. Such leakage would escape into a portion of the auxiliary building which is designed for minimum leakage and controlled access. The auxiliary building special ventilation system when actuated will draw all in-leakage air from this special ventilation zone and exhaust it through particulate and charcoal filters. C. Containment Svstem Functional Desicn Functionally, leakage from the primary containment to the annulus between the primary containment and the shield building is processed by recirculating annulus air through either of two redundant f an-filter trains, designated the shield building ven-tilation system (SBVS), to remove radioactivity prior to exhaust. The vast majority of the leakage that may bypass the annulus through lines penetrating containment is expected to enter the ABSVZ, where it is processed by exhausting zone air through either of two redundant fan-filter trains to remove radioactivity. A minor portion of the total leakage may occur so that it bypasses both the annulus and the ABSVZ. g NSP has used conservative values ir, evaluating shield building annulus air pressuz,e transients for the most severe loss-of-coolan accident (LOCA) and for a range of shield building leak rates (1% per day to 10% per day) assund e only one shield building fan-filter train operates 144t During the first 3 minutes following a LOCA, the steel primary containment temperature will increase and the resultant containment expansion and thermal tran-
- sient will raise the air pressure in the shield l
building annulus to approximately 7 inches of water above atmospheric pressure. From 3 minutes to 6 minutes af ter the LOCA, the shield building ven-l tilation system will reduce the annulus air pressure j to a negative value relative to air pressure outside the annulus. When the annulus air pressure reaches i 2 inches of water below atmospheric pressure, the recirculation damper in the SBVS will start to open. l About 24 minutes after the LOCA the air pressure will l reach a steady negative value, with the large re-circulation fan recirculating air through the filters ! and the small exhaust fan maintaining a negative pressure in the annulus. I
TS.5,2-4 REV -5 10/25/7'2 ry The fan filter trains in the ABSVZ will start automatically and the normal auxiliary building ventilation system will be isolated automatically within 1 minute following the LOCA. The fans in g the ABSVZ will maintain a negative pressure in is Zone. ( StNo $> 0 The 7tfHs has evaluate the L CA assuming the total primary containment leakage is 0.25% by weight per day, 0.1 wt / day of the primary contafnment leakage will bypass the annulus through lines penetrating containment and enter the ABSVZ, and that 0.01 wt%/ day gf the leakane will bypass both the annulus and the ABSVZ.(3), CO , (M , M The remainder is assumed to be recirculated through the shield building ventilation system to remove fission products prior to exhaust from the building.
> 4 In the *EC dose calculations'(3), C') , w, W conservative mixing assumptions are used. During the first few minutes when the shield building annulus pressure is pos-itive, direct leakage to the environs without mixing or filtration is assumed. During the time the ,
ventilation system is exhausting without recircu-
; lation, primary containment leakage is assumed to enter the SBVS inlet duct without mixing and to be filtered before being released to the environs.
During recirculation, leakage is assumed to enter the inlet duct without mixing and recirculated air from the outlet duct is assumed to mix in 50% of the annulus volume. The fraction of primary con-tainment leakage that goes to the ABSVZ is assumed to enter uhe ventilation system ducts without mixing e.nd to be filtered before discharge to the environs. The leakage values used in the AEC ctaff's dose analyses 4,
->(C , (5), (C form the bases for the limiting containment system leakage for containment system laakage tests in these Technical Specifications.
i i I References
. /SAR Appendix G j - FSAR, Section 5 and 14
-lP - SER, Section 1S SER, Supplement 1, issued later
- Si -
Letter from AEC to NSP dated November 29, 1973 j 6)
- Letter from AEC to NSP dated September 16, 1974 l i s-l 1
TS.S.3-1 . REV " 5/13/"I ,, s 5.3 REACTOR A. Reactor Core (Ra bre,reett- )
- 1. The reactor core contai s approximately 48 metric tons of uranium in the form of slightly en iched uranium dioxide pellets. The pellets are encapsulated in Zire loy-4 tubing to form fuel rods. The r eactor core is made up o 121 fuel assemblies. Each fuel assembly
. contains 179 fuel rods.
- 2. The average enrichment of the reload core is a nominal 2.90 weight per cent of U-235. The highest Uranium-235 loading is a nominal 39 grams of U-235 per axial centimeter of fuel assembly (average). )
- 3. In the reactor core, there are 29 full-length RCC assemblies that contain a 142-inglength of silver-indium-cadmium alloy clad with stainless stee .
B. Reactor Coolant System
@deceCe N
- 1. The design o~f the g etor coolant system complies with all applicable code requirements t g
- 2. All high pressure piping, components of the reactor coolant system and their supporting structures are designed to Class I requirements, and have been designed to withstand:
i n. The design seismic ground acceleration, 0.0' 6 g, acting in the horizontal and 0.04g acting in the vertical planes simultaneously, with stresses maintained within code
'llowable working stresses, a
- b. The maximum potential seismic ground acceleration 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.
- 3. The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.
C. Protection Systems
~
T The protection systems for the reactor and engineered safiity features are designed to applicable codes, including IEEE-279, dated 1068. The design includes a reactor trip for a high negative rate o ban e of neutron flux as measured by the excore nuclear instruments. The system is intended to gph a than one control rod g the Ifreactor only oneupon the abnormal control dropping rod is dropped, the of more core can be operated at full power for a short time, as permitted by Specification 3.10. References
'(
(2) F
, Se .2.3 2.1 and 3.2.3 (3 R. Ta .1-9 d
( , ion 7 g ushe., W 3A A 3- asM ' G\t. N. \ -Il h, ws AR., d #
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,s 5.4 ENGINEERED SAFETY FEATURES The engineered safety features include the containment system described in Specification 5.2, the emergency core cooling system, the containment air cooling system, the containment spray system, the post-accident combustible gas control system, emergency power supplies, component cooling water system, and the cooling water system.
These systems are designed to applicable industry codes, the 45MF General Design Criteria in Appendix A AC to 10CFR50, and %EG Safety Guides. Particular features for the Prairie Island plant include the following:
- 1. Several of the features are shared between the two units , including the onsite diesel generators, the cooling water system, and the motor-driven pumps of the auxiliary feedwater system. Shared systems are designed to mitigate the effects of an accident in one unit and simultaneousiv, provide for a hot shutdown in the other unit.s~ j; 3
- 2. The emergency cooling water pumps are driven by diesel engines. These diesel engines are designed and will be tested to the same reliabillty criteria as those for the diesel generators that supply emergency electrical power g
- 3. Thecoolgngwatersystemisautomaticallydivice into two FEt redundant loops by motor-operated valves which are actuated by a safety injection signal. Branches serving non-safety related equipment in the turbine rooms are isolated from the class I cooling water loops by motor-operated valves. Line breaks in these branches are sensed and the motor-operated valves are sc; o ed by -
instruments which monitor coincident .igh flow and low pressure in the class I cooling ater loops. References #C i
$13, FSAR Table 1.2-2 (2). FSAP Ecctic.; ^
l (A SM, S3 /O 4 e
, - - . - . . - . . ~ , , . - - - - -,,_,-..-_,,,,.,,---.-.,.,,.%..-.--,.m ,---,e.- .r ., . - - r - - - -*- w--- -----=~---- e-=vW- '
/0 CM /s,r(s $00 md 'IO, /0 Cf'C kUO y5 5_1 fp 50. 3 ca., appn h A nwt.T k u pr /0 CMC ladSd, ssse/ 9'O CfW An d A 0 5.5 RADIOACTIVE WASTE SYSTEMS u, I m e l ' * /'# " #
The design objective for the liqu'd radwaste system is to process the waste so that dis arges will approach essentially cero under normal erating conditions d
* " I
~
=
rail fracti~ of lacrn rart 20 limicc under design basis conditions. The design objective for the gaseous waste system is to release a small fraction of 10 CTP ?;rt 20- limits. Holdup time for the gaseous waste storage system is designed for the plant lifetime. The design objective of the s611d radwaste system is to package solid waste in accordance with applicable govern-ment regulations for offsite shipment. A. Accidental Releases The auxiliary building is designed as Class I (seismic) in areas containing the auxiliary building special ven-tilation system and radwaste storace area. The rad-
~
waste building is also designed to Class I standards .+1+ All radwaste tanks, filters, and equipment are either contained in a Class I (seismic) building or in ( G specially constructed areas to provide a substantial degree of control of the wastes should a liquid
) radwaste tank rupture.
B. Routine Releases
- 1. Licuid Wastes q The evaluation of the processing of li uid L) 4tt- wastes is described 2c fcilcuc-in the ISAR. M The liquid wastes from the non-aerated and aerated wastes systems (includine the laundry, hot shower,
' and decontamination station system) zerc cctir ted
) 2c"521,2:0 ;;11cnc per jccr,- processed throuch @
an evaporato:5 a demineralicert and 100% released to the environment. The, reactor coolant in the g gd chemical and volume control system (CVCS) -wee ' #
~
concidered to essentially -be- recycled 6nd lecc than i millicuric,0cr; car (rCi/yr' rcicaccc- *
-thrcuch ucctc dicpecal by the CVCS monitor tanic.
The blowdown treatment system ucc cvaluated at 40 h #88 %## -- a total blowdown rate of -10';pm, treate_d bv s F y;f' 4 demineralication, and-discharged /2t a concentration cf 10" MicrecuriccN per cubic centi =cter (uci/ce' The total activity eleased is estimated to be 3.2 curies per ye . (Ci/yr) , excluding tritium,
TS.5.5-2
= )CEV i
e and 820 Ci/yr of tritium to be released from the liquid radwaste system for both units. The liquid blowdown system is estimated to release 0.199 Ci/yr excluding tritium. Non-aerated drains from components within the reactor coolant system and a portion of the coolant letdown stream used for boron management will be processed through the chemical and volume control system (CVCS). Aerated drains from the floor drains, aerated equipment drains and leaks, decontamination drains, laboratory
' and sample drains, will be processed through the waste evaporator system or the aerated drain c;;pcie;e.-demineralizer treatment system.
I The laundry-shower water and certain decontam-l ination solutions will be treated in a special coagulation tank facility. All equipment in the liquid radwaste system is common to both i
. units except the steam generator blowdown flash I tanks, reactor coolant drain-tanks, and drain '
! tank pumps.
r
- 2. Gaseous Wastes z-asa sa-o -
a i The evaluation of the gaseous wastes is described
- f:ll:.:: in the' M The gas decay tanks are designed to hold gases for the plant life-
; time, but it was estimated that the release of one decay tank per year would account for unpre-dictable influent sources. This tank was estimated at 800 Ci/yr of Kr-85 (other isotopes decayed to negligible amounts) for the two units. The
- estimated gaseous release from containment purging is 478 Ci/yr of Xe-133 for the two units.
The total is thus 1,278 Ci/yr for the two units. A potential source of radioactivity is the atmospheric steam dump system during large power trantients. These releases are expected to be infrequent and small. Specification 3.9 requires an inventory of the activity released from this source to assure compliance with effluent release limits.
'4
a . ; TS.5.5-3 REV -1: ElEl';~! pfecl i- du
- 3. Solid Wastes Gofur ju s Z> - !
r~-- isc llaneous materials, such as paper, rags, and las ware, will be ___z______ int: 55 gallen drrr-. Spent resins and concentrates f the waste evaporator will be solidified in 55- ,.f * (gy n,g,0.) gallon drums or other regulation containers # 7 and stored in a shielded area prior to ship-ment offsite for burial. The total annual ship-ment is estimated to be 100 200 bazzela of
/I hicuid concentrates and miscellaneous materials IOcd d gg[ej and 250-350 cubic feet of spent resins for the two units.
The storage facilities and packaging and shipping are designed according to 10 CFR Part 71 and 49 CFR 170-199. 4 C. Process and Effluent Radiological Monitoring Svstem The process radiation monitoring system is designed to provide information on radioactive concentrations in certain systems, leakage from one system to another, ' gag 7 3) and radioactive concentrations released to the environment'.'f3t~ The monitoring includes: containment or containment purge vent, shield building vent and auxiliary building vent monitoring separately for particulate, gas and iodine; , steam generator blowdown liquid monitor; liquid waste disposal system effluent discharge line monitor; condenser air ejector gas monitor; radwaste treatment building vent monitor; control room ventilation supply monitor; spent fuel pool air exhaust monitor; residual heat removal cubicle air exhaust monitor; containment fan cooling water and component cooling system monitors; and discharge canal effluent monitor. The area radiation monitoring system is designed to provide information on radiation levels in various areas of the plant for personnel protection and for qualitative information of a systems condition such as a failed demineralizer. Ten channels monitor areas in the control room, containment, radwaste building, and auxiliary building. These monitoring systems will detect, indicate, annunciate, and/or record the concentrations or levels of activity to verify compliance with 10CFR Part 20 and keep radiation levels as low as practicable. References rSAR, Appe d' B /. USA 4, C dB "~ # ~ _ (2) FSA , ion 11.1 p, MS44, fec/M 92! Section 1.3 gg 4 /7 6 *)' 2. 3
TS.S.6-1 REV 70 0/20/05 f -s 5.6 FUEL HANDLING A. Criticality Consideration The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (seismic) structures. The spent fuel pi: has a stainless steel liner to ensure against loss of water The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed loca-tions. The fuel is stored vertically in an array with the center-to-center distance between assemblies sufficient to assure Keff 1 0.95 even if unborated water were used to fill the pit. In addition, fuel in the storage pool shall have a U-235 loading of 1 39.0 grams of U-235 per axial centimeter of fuel assembly (average). The criticality considerations as they relate to the dropping of a spent fuel cask (i.e., heavy load) drop onto the racks has been evaluated. The maximum Keff has been calculated to be 0.949 at a water /UO2 ratio of a . 2.0 with a boron concentration of.1800 ppm. __ B. Seent Fuel Storage Structure / [}$f44v 't ) / The spent fuel storage pool is en:losed with a reinforced concrete build-
' {
ing having 12- to 18-inch thick walls and roof.44Y The pool and pool enclosure are Class I (seismic) structures that afford protec: ion against loss of integrity from postulated tornado =issiles. The storage compar:- ments and the fuel transfer canal are connec:ed by fuel transfer slots that can be closed off with pneumatically sealed gates. The bottoms of the slots are above the tops of the active fuel in the fuel assemblies which will be stored ver:1cally in specially constructed racks. The spen: fuel pool has a reinforced concrete bottom slab nearly 6 feet thick and has been designed to minimize loss of water due to a dropped cask accident. In addition, the spent fuel cask will have an impac: limiter attached or a crash pad will be in place in the pool which will have the capability to absorb energy of i= pact due to a cask drop. This will result in no strue: ural da= age taking place to the pool which would resul: in significant leakage from the pool. Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies. C. Fuel Handling The fuel handling systa= provides the means of transporting and handling fuel from the time it reaches the plant in an unirradiated condition until 1: leaves after post-irradiation cooling. The system consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system.
)
TS.S.6-2 l RE'.' i 5/20/05-Major components of the fuel handling system are the manipulation crane, the spent fuel pool bridge, the auxiliary building crane, the fuel I transfer system, the spent fuel storage racks, the spent fuel cask, and j the rod cluster control changirg fixture. The reactor vessel stud
! tensioner, the reactor vessel head lifting device, and the reactor '
internals lifting device are used for preparing the reactor for refueling and for assembling the reactor after refueling. ' Upon arrival in the storage pit, spent fuel will be removed from the transfer system and placed, one assembly at a time, in storage racks , using a long-handled manual tool suspended from the spent fuel pit bridge crane. After sufficient decay, the fuel will be loaded into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane. l The load drop consequences of a spent fuel cask for Prairie Island have i .been evaluated. It is not possible, due to physical constraints, for a cask to be dropped into the large pool (pool no. 2). A load path has
- been defined which provides fer safe movement of the cask. Travel inter-t locks and mechanical stops prevent cask movement outside of this path.
The only safety-related equipment that can be impacted directly during a j cask drop along this path is the fuel sected in the small pool (pool No. 1). The consequences of this drop have been evaluated and found to meet the NRC staff criteria contained in NUREG-0612 if at least 50 days j have elapsed since reactor shutdown for fission gas release considera-l tions and the pool water contains at least 1800 ppm boron for criticality considerations. While 50 days was determined adequate, a minimum decay ] period of 5 years has been incorporated ~into these technical specifica-tions to provide additional margin in meeting the criteria specified in NUREG-0612 for fission gas releases, while not restricting the plant's
'; operational flexibility. A cask impact limiter or crash pad prevents significant structural damage to the pool floor.
l The spent fuel cask will be lowered 66 feet from the auxiliary building to the railroad car for offsite transportation. Specification 3.8 will l limit this loading operation so that if the cask drops 66 feet, there
; will not be a significant release of fission products from the fuel in j the cask.
D. Seent Fuel Storare Cacacity l j The spent fuel storage facility is a two-comoartment pool that. if I completely filled with fuel storage racks, provides up to 1582 storage locations. The southeast corner of the small pool (pool No. 1) also ] serves as the cask lay down area. During times when the cask is being i used, four racks are removed from the small pool. With the four storage racks in the southeast corner of pool 1 removed, a total of 1386 storage locations are provided. To allow insertion of a shipping cask, total storage is limited to 1386 assemblies, not including those assemblies I which can be returned to the reactor. Reference #- - - ' -( l ' ThR, %,.:i:: ^
. /. MM, Y88
- i J
j 1
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TS.6.2-6 REV-70 0/25/0! t
- f. Investigations of all Reportable Events and events requiring Special Reports to the Com=ission.
- g. Drills on emergency procedures (including plant evacuation) and adequaev '-
of communication with off site support grouns. ' (ages h as exe**fN *-
- h. All procedures required by these Technical Specifications, including 58'0
- 0
implementing procedures of the Emergency Plan, and the Security Plan, shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two years.
- 1. Special reviews and investigations, as requested by the Safety Audit Commit tee.
- 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). ,,
- 5. Authority .
. The OC shall be advisory to the Plant Manager. In the event of a disagree-ment between the recommendations of the OC and the Plant Manager, the ' course determined by the Plant Manager to be the core conservative vill be followed.
A written summary of the disagreement will be sent to the General Manager Nuclear Plants 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 caterial reviewed. The =inutes shall be distributed to each me=ber of the OC the Chairman and each me=ber of the Safety Audit Committee, the General Manager Nuclear Plants and others designated by the OC Chairman or Vice Chairman.
- 7. Procedures A written charter for the CC shall be prepared that contains :
- a. Responsibility and authority of the group
- b. Content and method of submission of presentations to the Operations Committee
- c. Mechanism for scheduling mattings
- d. Provision for meeting agenda 1
\
y TS.6.5-1 REV ',1 2/ 2 /03-- l 6.5 PLANr OPERATING PROCEDURES Detailed written procedures, including the applicable checkoff lists and instructions, covering areas listed below shall be prepared and followed. These procedures and changes thereto, except as specified in TC 0.0.0. , shall be reviewed by the Operations Committee and approved by'a member of plant management designated by the Plant Manager. A. Plant Operations
> T S____
C>.5_ S d G- 1. Integrated and system procedures for normal startup, operation and shutdown of the reactor and all systems and components involving nuclear safety of the facility.
- 2. Fuel handling operations -
- 3. Actions to be taken to correct specific and foreseen potential or actual malfunction of systems or components including responses to alarms, primary system leaks and abnormal reactivity changes and including follow-up actions required after plant protective system actions have initiated.
- 4. Surveillance and testing requirements that could have an effect on nuclear safety.
l S. '..v 1...n.nev ca.d aea ef :L .ee ai:; Aen.
. Implementing procedures of the Facility Emergency Plan, including procedures for coping with emergency conditions involving potential or actual releases of radioactivity.
h Implementing procedures of emergency plans for coping with earthquakes and floods. ~he flood emergency plan shall require plant shutdown for water levels at the site higher than 692 feet above MSL. 7 / Implementing procedures of the fire protection program. Implementing procedures for the Process Control Program g [ and Offsite Dose Calculation Manual including quality control measures. Drills on the procedures specified in A.3. abose, shall be sencucted as a part of the retraining program. B. Radiological Radiation control procedures shall be maintained and made available to all plant personnel. These procedures shall show permissible radiation exposure and shall be consistent with the requirements of 10C7R20. This radiation protection program shall be organized to t meet the requirements of 10CFR20.
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. 'Offsite Dose Calculation Manual (ODCM) 1 M h The CDCM shall be approved by the Caser. salon prLor to inu.aA 1 ~/ h -
implementation. Changes to the ODCM shall satisfy the following requirements:
- 1. Shall be submitted to the Commission with the Semi-Annual Radio-active Effluent Report for the period in which the change (s) were made effective. This submittal shall contain: .
- a. sufficiently detailed infor=ation to totally support the rationale for the enange without benefit of additional or supplemental infor'sation. Information submitted should consist of a package of those pages of the CDC:1 to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s).
- b. a determination that the change vill not reduce the accuracy or reliability of dose calculations or setpoint deter =inations; and
' c. documentation of the fact that the change has been reviewed and found . acceptable by the Operations Committee.
- 2. Shall become ef fective upon review and acceptance by the Operations Com=ittee.
Temporary Changes to Procedures yd,Q ffee rey,dwec) g wA C: J'. ~ Tenporary changes to procedures described in A,3,0,D.-and-Tr above, which do not change the intant of the original procedure may be made with the concurrence of two indiriduals holding senior operator licenses. Such changes shall be documented, reviewed by the Operations Com '.ttee and approved by a member of plant management designated by the Plant Manager within one month, f V ej M I jD/ g,,,, es r'o sewn k jrrocedv'cJ N* N f op,. fk Co,,,.:&e sk // k an N W ? w e _ u ,.. , u V/4-ftiole t I
TS.6.6-2 REV 50 10/16/61
- 6. Plant radiation and contamination surveys.
- 7. Changes made to the plant as it is described in the Final Safety Analysis Report, reflected in updated, corrected and as-built drawings.
- 8. Cycling beyond normal ILaits for those components that have been designed to operate safely for a limited number of cycles beyond such limits.
- 9. Reactor coolant system in-service inspections.
- 10. Minutes of meetings of the Safety Audit Committee.
!! ?ecerde ef En fir ----t rl Qurlifiertier rhi:5 ;;; ;.....J
" >cr the pre-ici tr cf perr;rr;5 5.S .
44,- Records of the service lives of all safety-related snubbers, including the date at which the service life commences and associated installation and maintenance records. 11 t I
. -. . . - . . . - - , , . . - - - . . - . . . _ . . _ . _ . . . - . , , _ - . . - _ _ , . - - , , _ , - - , , , , - , , - - - . , - - , _ _ . - - . . . - - - , ~ . - - - - _ - . _ -
TS.6.7-2 REV 75 10/11/05 Occuestional Excesure Reeort. An annual report of 2. occupational exposure covering the previous calendar year shall be submitted prior to March 1 of each year. The report should tabulate on an annual basis the nu=ber of sta:1on, utility and other personnel (including con-
- actors) receiving exposures greater than 100 =re=/yr and their associated =an-re: exposure according to work and job functions, e.g., reactor operations and surveillance, inservice inspection, routine =aintenance, special =aintenance (describe
=aintenance) , vaste processing, and ref ueling. The dose assign =e.n: to various duty functions =ay be esti=ates based on pocket dosi=e:er, TLD, or fil: badge =easure=ents. S=all exposures totalling less than 20% of :he individual :otal dose need no: be accounted for. In the aggregate, a: leas: 80%
of the total whole body dose received fro = external sources shall be assigned to specific =ajor work fune: ions.
- 3. Monthlv Ooeratine Reeert. A =enchly report of operating statistics and shutdown experience covering the. previous =onth shall be sub=i::ed by the l$th of :he f ollowing =onth.-:: the Offi:: cf M nag:: n: an d : g = '.n c.1f : i , ' . -
R ;ula:: ;: 0 ;;i::ica,*a:ning:;;, 20 20~j3. C. kel f - D it. A c, N S * # "
- g"" j g3 tj o c. lea r- Rep
- bh W#**9 3 ge' 2.o55 C o M 9;* % Qu%hn >
M W (1) This repor: supple =en:s the require =ents of 10 CFR 20, See: ion 20.407. If 10 CFR 20, Sec:1on 20.407 is revised te include such inforna:1on, this Specifica: ion is unnecessary.
* .5.0 1 2:*/ 10 10/01/20 Environmental Qualification A. no later than June 30, 1982 all safety-related electrical equi #
ment 'n the facility shall be qualified in accordance with th provisi of: Division of Operating Reactors " Guideline .or Evaluating 'ronmental Qualification of Class IE E .rical Equipment in Ope *ing Reactors" (DOR Guidelines )- or, NURIG-0588
" Interim Staff Posit on Environmental Qua ' teation of Safety-Related Electrical Equip t", December . Copies of these documents are attached to Or for F tfication of Licenses DPR-42 and DPR-60 dated October 24, 19 B. By no later than Decembe- , 1980, comple and audicable records must be available an incained at a centra cation which describe the environmenta ualification method used for a safety-related electrical e .pment in sufficient detail to document e degree of complian with the DOR Guidelines or NURIG-0588. Therea. r, such rece should be updated and maintained current as equipment '
aced, further tested, or otherwise further qualified. 9 l l l l l
o .. Exhibit C Prairie Island Nuclear Generating Plant License Amendment Request Dated July 15, 1986 Revised Technical Specification Pages Exhibit C consists of the proposed Technical Specifications pages with the changes shown in Exhibit B incorporated. The proposed pages are listed below: TS-il Table TS.4.10-1 (Page 4 of 4) TS-vii Table TS.4.10-2 (Page 1 of 2) TS-viii - Table TS.4.10-2 (Page 2 of 2) TS-ix TS.5.1-1 TS-x TS.S.1-2 . TS.3.1-3* TS.5.2-3 TS.3.1 4 T5.5.2 4 TS.3.1-5 TS.5.3-1 TS.3.1-6 TS.5.4-1 TS.3.ll7 TS.S.5-1 TS.3.1 8 TS.S.5-2 TS.3.1-11 TS.S.5-3 TS.3.1-12 TS.S.6 1 TS.3.1-13 TS.S.6 2 Figure TS.3.1-1** TS.6.2-6 Figure TS.3.1-2 TS.6.5-1 Figure TS.3.1 3 TS.6.5 4 TS,4.10-1 TS.6.6 2 Table TS.4.10 1 (Page 1 of 4) TS.6.7-2
- Pages TS.3.1-2A, 3, 3A, 4, 5 and 6 are being renumbered to TS.3.1 3 through 8, existing pages TS.3.1 7 and 8 are being deleted.
** Tables TS.3.1-1 and 2 and Figures TS.3.13 and 4 are being deleted; existing Figure TS.3.1-5 is being renumbered to TS.3.1 3.
1
o . I TS-11 REV i TABLE OF CONTENTS (Continued) TS SECTION TITLE PAGE 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING TS.2.1-1 2.1 Safety Limit, Reactor Core TS.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure TS.2.2-1 2.3 Limiting Safety System Settings, Protective TS.2.3-1 Instrumentaiton A. Protective Instrumentation Settings for Reactor TS.2.3-1 Trip B. Protective Instrumentation Settings for Reactor TS.2.3-3 Trip Interlocks C. Control Rod Withdrawal Stops TS.2.3-4 3.0 LIMITING COVDITIONS FOR OPERATION 3.1 Reactor Coolant System TS.3.1-1 A. Operational Components TS.3.1-1
- 1. Coolant Pumps TS.3.1-1
- 2. Steam Generators TS.3.1-1A
- 3. Requirements for Decay Heat Removal Below TS.3.1-1A 350'F
- 4. Pressurizer TS.3.1-2
- 5. Reactor Coolant Vent System TS.3.1-3 B. Heatup and Cooldown TS.3.1-6 C. Leakage TS.3.1-9 D. Maximum Coolant Activity
- TS.3.1-11 E. Maximum Reactor Coolant Oxygen, Chloride TS.3.1-14 and Fluoride Concentration F. Minimum Conditions for Criticality TS.3.1-17 C. Minimum Conditions for RCS Temperature Less TS.3.1-19 Than MPT H. Primary Coolant System Pressure Isolation TS.3.1-21 Valves 3.2 Chemical and Volume Control System TS.3.2-1 3.3 Engineered Safety Features TS.3.3-1 A. Safety Injection and Residual Heat Removal TS.3.3-1 Systems B. Contairment Cooling Systems TS.3.3-2 C. Component Cooling Water System TS.3.3-4 D. Cooling Water System TS.3.3-5 3.4 Steam and Power Conversion System TS.3.4-1 A.1 Safety and Relief Valves TS.3.4-1 A.2 Auxiliary Feed System TS.3.4-1 A.3 Steam Exclusion System TS.3.4-2 A.4 Radiochemistry TS.3.4-2
TS-vii REV TABLE OF CONTENTS (Continued) TS SECTION TITLE PAGE l 6.0 ADMINISTRATIVE CONTROLS TS.6.1-1 6.1 Organization 6.2 TS.6.1-1 Review and Audit TS.6.2-1 A. Safety Audit Committee (SAC) TS.6.2-1
- 1. Membership TS.6.2-1
- 2. Qualifications TS.6.2-1
- 3. Meeting Frequency TS.6.2-2
- 4. Quorum TS.6.2-2
- 5. Responsibilities TS.6.2-2
- 6. Audit TS.6.2-3
- 7. Authority TS.6.2-4
- 8. Records TS.6.2-4
- 9. Procedures TS.6.2-4 B. Operations Committee (OC)
TS.6.2-5
- 1. Membership TS.6.2-5
- 2. Meeting Frequency TS.6.2-5
- 3. Quorum TS.6.2-5 4., Responsib111 ties TS.6.2-5
- 5. Authority TS.6.2-6
- 6. Records TS.6.2-6
- 7. ?rocedures TS.6.2-6 6.3 Special Inspections and Audits TS.6.3-1 6.4 Safety Limit Violation 6.5 TS.6.4-1 Plant Operating Procedures TS.6.5-1 A. Plant Operations TS.6.5-1 B. Radiological , TS.6.5-1 C. Maintenance and Test TS.6.5-3 D. Process Control Program (PCP)
TS.6.5-3 E. Offsite Dose Calculation Manual (ODCM) TS.6.5-4 F. Security TS.6.5-4 G. Temporary Changes to Procedures l TS.6.5-4 6.6
- Plant Operating Records TS.6.6-1 A. Records Retained for Five Years TS 6.6-1 i B. Records Retained for the Life of the Plant TS.6.6-1 l 6.7 Reporting Requirements TS.6.7-1 A. Routine Reports TS.6.7-1 i
- 1. Startup Report TS.6.7-1
- 2. Occupational Exposure Report TS.6.7-2
- 3. Monthly Operating Report TS.6.7-2 i
- 4. Steam Generator Tube Inservice Inspection TS.6.7-2
- 5. Semiannual Radioactive Effluent Release TS.6.7-3
, Report 1
- 6. Annual Summaries of Meteorological Data TS.6.7-3
- 7 Report of Safety and Relief Valve Failures TS.6.7-4 and Challenges 4
B. Reportable Events TS.6.7-4 i
TS-viii REV TABLE OF CONTENTS (Continued) TS SECTION TITLE PAGE C. Environmental Reports TS.6.7-4
- 1. Annual Radiation Environmental Monitoring TS.6.7-4 Reports
- 2. Environmental Special Reports TS.6.7-5
- 3. Other Environmental Reports TS.6.7-5 (non-radiological, non-aquatic)
D. Special Reports TS.6.7-5 a
TS-ix REV 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 Instrumsntation 3.9-2 Radioactive Gaseous Effluent Monitoring Instrumentation 3.14-1 Safety Related Fire Detection Instruments 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 4.4-1 Unit 1 and Unit 2 Penetration Designation for Leakage Tests 4.10-1 Radiation Environmental Monitoring Program (REMP) Sample Collection and Analysis 4.10-2 REMP - Maximum Values for the Lower Limits of Detection 4.10-3 REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Rquirements 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-x REV APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits Reactor Core, Thermal 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 l 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit i Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 Ci/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 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 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 V(Z) as a Function of Core Height 4.4-1 Shield Building Design In-Leakage Rate 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating Organizations 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group
I TS. 3.1-3 l REV l
- 5. Reactor Coolant Vent System
- a. A reactor shall not be made or maintained critical nor shall it be heated or maintained above 200*F unless reactor coolant vent system paths from both the reactor vessel head and pressurizer steam space are operable and closed except as specified in 3.1.A.S.b and 3.1.A.S.c below.
- b. During Startup Operation or Power Operation, any one of the following conditions of inoperability may exist for each unit i until operability is restored:
i
- 1. Both of the parallel vent valves in the reactor vessel
- head vent path are inoperable.
}
- 2. Both of the parallel vent valves in the pressurizer vent path are inoperable.
i
- 3. The vent valve to the pressurizer relief tank discharge line e is inoperable.
l 4. The vent valve to the containment atmospheric discharge line
, is inoperable. ,
j If during Startup Operation Power Operation any one of these conditions is not restored to an operable status within 30 days, the reactor shall be placed in Hot Shutdown within 6 hours and in Cold Shutdown within the following 30 hours.
- c. With no reactor coolant vent system path operable, restore at least one vent path to operable status within 72 hours or be in
, the Hot Shutdown condition within 6 hours and the Cold Shutdown condition within the following 30 hours. i l l 1 . i l I i 4 rs----- - - r--- , y e.-_.y.-.t- ,w--,.----.mm_,--c.m, , - - . . . , _,.. . . - . ---- - .._ -r-_,,-%--,,,-_.----_.w-.., , _ _ -..-. . - - - , . - , - - - . , , ,
l j i TS.3.1-4 l REV Basis When the boron concentration of the reactor coolant system is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uniform boron concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the equivalent of the primary system volume in approximately one-half hour.
" Steam Generator Tube Surveillance", Technical Specification 4.12, identifies steam generator tube imperfections having a depth 50% of the 0.050-inch tube wall thickness as being unacceptable for power operation. The results of steam generator burst and tube collapse tests submitted to the staff have demonstrated that tubes having a wall thickness greater than 0.025-inch have adequatemarginsofsafetyagainstfailurefuetoloadsimposedbynormal plant operation and design basis accidents.
Part A of the specification requires that both reactor coolant pumps be operat-ing when the reactor is critical to provide core cooling in the event that a loss of flow occurs. In the event of the worst credible coolant flow loss (loss of both pumps from 100% power) the minimum calculated DNBR remains well above 1.30. Therefore, cladding damage and release of fission products to the. reactor coolant will not occur. ' Critical operation, except for low power physics tests, with less than two pumps is not planned. Above 10% power, an automatic reactor trip will occur if flow from either pump is lost. Below 10% power, a shutdown under administrative control will be made if flow from either pump is lost. The pressurizer is needed to maintain acceptable system pressure during normal plant operation, including surges that may result following anticipated transients. Each of the pressurizer safety valves is designed to relieve 325,000 lbs per hour of saturated steam at the valve set point. Below 350*F and 450 psig in the reactor coolant system, the residual heat removal system can remove decay heat and thereby control system temperature and pressure. If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve relief pressure would be less than half the valves' capacity. One valve therefore provides adequate defense against over-pressurization of the reactor coolant system for reactor coolant temperatures less than 350*F. The combined capacity of both safety valve greater than the maximum surge rate resulting from complete loss of load.g is
TS.3.1-5 REV Basis (continued) The requirement that two groups of pressurizer heaters be operable provides , assurance that at least one group will be available during a loss of offsite power to maintain natural circulation. Backup heater group "A" is normally supplied by one safeguards bus. Backup heater group "B" can be manually transferred within minutes to the redundant safeguards bus. Tests have confirmed the ability of either group to maintain natural circulation conditions. The pressurizer power operated relief valves (PORV's) operate to relieve reactor coolant system pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The PORV's are pneumatic valves operated by instrument air. They fail closed on loss of air or loss of power to their DC solenoid valves. The PORV block valves are motor operated valves supplied by the 480 volt safeguards buses. The Specifications require that at least two methods of removing decay heat are available for each reactor. Above 350*F, both steam generators must be operable to serve this function. Below 350*F. either a steam generator or a residual heat removal loop are capable of removing decay heat and any combination of two loops is specified. If redundant means are not avail-able, the reactor is placed in the cold shutdown condition. The reactor coolant vent system is provided to exhaust noncondensible gases from the reactor coolant system that could inhibit natural circulation core cooling. The operability of at least one vent path from both the reactor vessel head and pressurizer steam space ensures the capability exists to perform this function. The vent path from the reactor vessel head and the vent path from the pressurizer each contain two independently emergency powered, energize to open, valves in parallel and connect to a common header that discharges either to the containment atmosphere or to the pressurizer relief tank. The
- lines to the containment atmosphere and pressurizer relief tank each contain an independently emergency powered, energize to open, isolation valve.
This redundancy provides protection from the failure of a single vent path valve rendering an entire vent path inoperable. An inoperable vent path valve is defined as a valve which cannot be opened or whose position is unknown, l A flow restriction orifice in each vent path limits the flow from an ( inadvertent actuation of the vent system to less than the flow of the reactor j coolant makeup system. References
- 1. FSAR, Section 14.1.9
- 2. Testimony by J Knight in the Prairie Island Public Hearing on January 28, 1975.
1 1 l
I
. TS.3.1-6 l REV B. HEATUP AND COOLDOWN Specification:
- 1. The Unit 1 and Unit 2 reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures TS.3.1-1 and TS.3.1-2. l
- a. Allowable combinations of pressure and temperature for specific
]j temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation.
- b. Figures TS.3.1-1 and TS.3.1-2 define limits to assure preven-tion of non-ductile failure only. For normal operation other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
- 2. The limit lines shown in Figures TS.3.1-1, TS.3.1-2 shall be re-calculated periodically using methods discussed in the Bases section.
^
- 3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the vessel is below 70*F.
< 4. The pressurizer heatup rate shall not exceed 100*F/hr and the pressurizer cooldown rate shall not exceed 200*F/hr. The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320*F. i
TS.3.1-7 ' REV Bases Pressure / Temperature Limits Appendix G of 10 CFR Part 50, and the ASME Code require that the reactor coolant pressure boundary be designed with sufficient margin to insure that, when stressed under operating, maintenance, testing, and postulated accident conditions, the boundary behaves in a nonbrittle manner, the probability of rapidly propagating fracture is minimized and the design reflects the uncertainties in determining the effects of irradiation on material properties. Figures TS.3.1.-1 and 2 have been developed (Reference 1) in accordance with these regulations. The curves are based on the properties of the most limiting material in either unit's reactor vessel (Unit I reactor vessel weld W-3) and are effective to 15 EFPY. The curves have been adjusted for possible errors in the pressure and temperature sensing instruments. The curves define a region where brittle fracture will not occur and are determined from the meterial characteristics, irradiation effects, pressure stresses and stresses due to thermal gradients across the vessel wall. Heatup Curves During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall. At the inner wall of the vessel, the thermal induced t compressive stresses tend to alleviate the tensile stresses induced by the internal pressure. Therefore, a pressure-temperature curve based on steady state conditions (i.e., no thermal stresses) represents a lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location. The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location. The thermal gradients established during heatup
- produce tensile stresses at the outer wall of the vessel. These stresses l are additive to the pressure induced tensile stresses which are already l
present. The thermal induced stresses at the outer wall of the vessel are dependent on both the rate of heatup and the time along the heatup ramp; ( therefore, a lower bound curve similar to that described for the heatup of l the inner wall cannot be defined. For the cases in which the outer wall of the vessel becomes the stress controlling locatic , sach heatup rate of
- interest must be analyzed on an individual basis. The heatup limit curve i is a composite curve prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour.
1 i l l l l -. - _
TS.3.1-8 REV Bases (continued) Cooldown Curves During cooldown, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from tensile at the inner wall to compressive at the outer wall. The thermal induced tensile stresses at the inner wall are additive to the pressure induced tensile stresses which are already present. Therefore, the controlling location is always the inside wall. The cooldown limit curves were prepared utilizing the same type of analysis used to calculate the heatup curve except that the controlling location is always the inside wall. Criticality Limits Appendix G of 10 CFR Part 50 requires that for a given pressure, the reactor must not be made critical unless the temperature of the reactor vessel is 40*F above the minimum permissible temperature specified on the heatup curve and above the minimum permissible temperature for the inservice hydrostatic pressure test. For Prairie Island the curves were prepared, requiring that criticality must occur above the maximum permissible temperature for the inservice hydrostatic pressure test. ASME Code Section XI Inservice Test Limits The pressure temperature limits for the ASME Code Section XI Inservice Test Limits (hydrostatic pressure test) are less restrictive than the heatup and cooldown curves to allow for the periodic inservice hydrostatic test. These limits are allowed to be less restrictive because the hydrostatic test is based on a 1.5 safety factor versus the 2.0 safety factor built into the heatup and cooldown curves and because the test is run at a constant temperature so the thermal stresses in the vessel are minimal. Reference
- 1. USAR, Section 4.2
. - - - - - _ _ .~ . - - ._ . . . _ _ _ _ . _ , _ - - - _
TS.3.1-11 . REV D. MAXIMUM COOLANT ACTIVITY
- 1. The specific activity of the primary coolant 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.
- 2. In Specification 3.1.D.1 the following definitions apply:
(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 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."
- (b) E shall be the average (weighted in proportion to the concentra-tion of each radionuclide 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 half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
- 3. .If a reactof is above hot shutdown and RCS 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 but within the allowable limit (below and to the lef t of the line) shown on Figure TS.3.1-3, operation may continue for up to 48 hours l provided that the cumulative operating time under these cir-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 DOSE EQUIVALENT I-131 exceeding 500 hours in any consecutive 6-month period, a special report to the Commission shall be submitted within 30 days indicating the number of hours above this limit. (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.3.1-3, the affected reactor shall be l 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 500*F or less within 6 hours of detection. i i l (
TS.3.1-12 REV
- 4. If a reactor is at or above cold shutdown:
(a) 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 require-ments of item 4a of Table 4.1-2B until the specific activity of the primary coolant is restored to within its limits. A special report 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. Basis 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 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 c.ontinue 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 l occur following changes in thermal power. Operation with specific activity levels exeeding 1.0 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure TS.3.1-3 must be restricted to no more than 800 hours l per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by Figure TS.3.1-3 increase the 2 hours thyroid l
TS.3.1-13 . REV dose at the site boundary by a factor 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 EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800 hour limit. 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. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained. S 9
FIGURE TS.3.1-1 UNIT 1 and UNIT 2 REACTOR COOLANT SYSTEM HEAT UP LIMITATIONS (Applicable for First 15 EFPY of Operation ) 2500 ..,._. . . , . _ . ., _ . _ , . _ . , . _ , . _ . .. . ._r ,. _ ,.__,__ __ . . . ., .,. , . _ _ . . _ , . _ _ , . _ _ , . _ _ _ , . _ , . _ _ , . _ _ , ... j . ,, ._. .. , . , . _ . _ . , . _ _, ., __ . _ _ . _ t.. _q. p_j._; .
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..........u. . 6_a._a_ 4_. ... 6--" ASME Code Seeflon XI 4- d- *- -*-- --d-- --*-- --6--'-
. . ; _ _ ; _ . ; . .;. . .-a...._.__.....p.j._j._;__ ..)._;.. p.j__ __;_ E--E ~ Inservice Test Limits T 'tv4D4 - -E--E .--4-- --l' ' 'r-' r--E-
~ -- "-~ ~ ~ ' ^ '
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FIGURE TS.3.1-2 UNIT 1 and UNIT 2 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS (Applicable for First 15 EFPY of Operation ) 2500 .. .. .... . . ... . _ , . _ ,.... . - t.. ,,
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TS.4.10-1 REV , 4.10 RADIATION ENVIRONMENTAL MONITORING PROGRAM Applicability Applies at all times to the periodic monitoring and recording of radioactive effluents found in the plant environs. Objective To provide for measurement of radiation levels and radioactivity in the site environs on a continuing basis. Specification A. Sample Collection and Analysis
- 1. The Radiation Environmental Monitoring Program described in Table 4.10-1 shall be conducted. Radioanalysis shall be conducted meeting the requirements of Table TS.4.10-2. A map and a table identifying the locations of the sampling shall be provided in the Offsite Dose Calculation Manual (ODCM).
- 2. Whenever the Radiation Environmental Monitoring Program is not being conducted as specified in Table TS.4.10-2 the Annual Radiation Environmental Monitoring Report shall include a description of the reasons for not conducting the
, program as required and plans for preventing a recurrence.
- 3. Deviations are permitted from the required sampling schedule if samples are unobtainable due to hazardous conditions, seasonable unavailability, or to malfunction of automatic sampling equipment. If the latter occurs, every effort shall be made to complete corrective action prior to the end of the next sampling period.
- 4. With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 4.10-3 when averaged over any calendar quarter, in lieu of any other report, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Report pursuant to Specification 6.7.C.2(a). When more than one of the radionuclides in Table 4.10-3 are detected in the sampling l medium, this report shall be submitted if:
concentration (1) , concentration (2) + * * * >l .0 limit level (1) limit level (2) When radionuclides other than those in Table 4.10-3 are l detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.9.A.2, 3.9.B.2, or 3.9.B.3. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radio-logical Environmental Monitoring Report.
TABLE TS.4.10-1 (Pags 1 of 4) PRAIRIE ISLAND NUCLEAR CENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis
- 1. AIRBORNE Samples from 5 locations: Continuous Sampler Radioiodine analysis Radiciodine and 3 samples from offsite operation with weekly for I-131 Particulates locations (in different sample collection sectors) of the highest weekly Particulate:
calculated annual average Gross beta activity on ground level D/Q, each filter weekly *. I sample from the vicinity Analyses shall be per-of a community having the formed more than 24 hours highest calculated annual following filter change, average ground-level D/Q, and Perform gamma isotopic 1 sample from a control loca- analysis on composite tion specified in the ODCM (by location) sample quarterly.
- 2. DIRECT 32 TLD stations established Quarterly Gamma dose RADIATION with duplicate dosimeters quarterly placed at the following locations:
- 1. Using the 16 meteoro-logical wind sectors as guidelines, an inner ring of stations in the pig general area of the site r.
M boundary is established and an outer ring of Cl stations in the 4 to 5 mile - distance from the plant r. site is established. Because j' of inaccessibility, seven sectors in the inner and ;g outer rings are not covered $m
~
- If Gross beta activity in any indicator sample exceeds 10 times the yearly average of the control sample, a gamma isotopic analysis is required. %
** Sample locations are given on the figure and table in the ODCM. d5
TABLE TS.4.10-1 (Page 4 of 4) PRAIRIE ISLAND NUCLEAR GENERATING PLANT RADIATION ENVIRONMENTAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS Number of Samples Exposure Pathway and Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis
- c. Food Products One sample of corn from any At time of harvest Gamma isotopic analysis field that is irrigated by of edible portion of water into which liquid plant each sample wastes have been discharged ***
One sample of broad leaf At time of harvest I-131 analysis of edible vegetation from highest D/Q portion of each sample garden and one sample from 10-20 miles w .a Qi% N u 9 n
- Sample locations are given on the figure and table in the ODCM.
- As determined by methads outlined in the ODCM. lQ
- Sample locations are given on the figure and table in the ODCM.
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rl 0 ( e/ 1 ti 4 ( 5 0 5 0 5 5 1 aC 0 1 3 1 1 1 Wp 0 ( 0 2 a s t s e C a i b o b 7 L s C N 3 - y s 0 - d 1 a l s n e 6, n r 7 , B a o M F Z Z 1 4 0 n r g H 4 8 5 6 5 5 9 3 1 3 1 4 1 A 3 5
TABLE *TS.4.10-2 I TABLE NOTATION l l a - The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal. For a particular measurement system (which may include radiochemical separation): LLD = E.V. 2.22 . Y . exp(-A A t) Where: LLD is the apriori lower limit of detection as defined above (as picocurie per unit mass or volume), s, is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute). In calculating the LLD for a radionuclide determined by gamma-ray , spectrometry, the background shall include the typical contributing of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples). Typical values of E, V, Y and At shall be used in the calculations. E is the counting efficiency (as counts per transformation), 2.22 is the number of transformation per minute per picocurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and ma at is the elapsed time between sample collection (or end of the sample collection period) and time of Q @i counting. p a b - LLD for drinking water. ? c - Total for parent and daughter d - These LLDs apply only where "131I analysis" is specified. Es e - Where " Gamma Isotopic Analysis" ig4 spec 6 Mn,ggied,5gheLggspeggficaggonapplies Fe, Co, Co. Zn, Zr-Nb, ,s jg7the{gglowingrygfonuclides: 7 Cs, Cs, and Ba-La. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported. % w Mg
TS.S.1-1 REV 5.0 DESIGN FEATURES 5.1 SITE The Prairie Island Nuclear Generating Plant is located on property owned by Northern States Power (NSP) Company at a site on the west bank of the Mississippi Ri'rer, approximately 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 715 meters, and the low population zone distance is 1-1/2 miles. The nearest population center of 25,000 or more people is South Saint Paul. These site characteristics comply with defi-nitions in 10CFR100 (Reference 1). l The U S Army Corp of Engineers controls the land within the exclusion area that is 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 l staff evaluations. Its evaluation of accidentgl airborne releases is based on a relative concentration of 9.8 x 10 seconds per cubic meter at the site boundary. Its evaluation _gf routine releases is based on a relative concentration of 1.5 x 10 seconds per cubic meter (Reference 3). l The flood of record in 1965 produced a water surface elevation of +688 feet MSL at the site. The calculated probable maximum flood (PMF) level is
+703.6 feet mean sea level (MSL), and the estimated wave runup could reach
+706.7 feet MSL. (See Section 2.4.2 of this report.) Plant grade 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 constitutes 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 safety related facilities. The cooling water pumps in the screenhouse are designed to operate up to a flood 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. The Technical Specification 6.5 A.7. requires an emergency procedure that will necessitate plant shutdown for flood water levels above +692 feet MSL at the plant site. The emergency procedure will assure the proper
. TS.5.1-2 REV erection of flood protection panels and assure an orderly shutdown of the plant and protection of safety related facilities. This procedure will provide for progressive action levels to prevent the possibility of unsafe plant operation and will include requirements for periodic drills to test flood protection measures, such as erecting flood protection panels.
The plant is designed for a design basis earthquake having a horizontal ground acceleration of 0.12g and an operational basis earthquake having a horizontal ground acceleration of 0.06g. An emergency procedure will be prepared in accordance with Specification 6.5.A.7 to define actions l required for earthquakes, including plant shutdown and inspection if an operational basis earthquake is measured at the site.
- e.
- References
- 1. USAR, Section 2.2.1
- 2. USAR, Section 3.4.5
- 3. SER, Sections 2.3.4 and 2.3.5
TS.5.2-3 REV I annulus through certain penetrations in the event of leakage in their isolation valves. Such leakage would escape into a portion of the auxiliary building which is designed for minimum leakage and controlled access. The auxiliary building special ventilation system when actu-ated will draw all in-leakage air from this special ventilation zone ' and exhaust it through particulate and charcoal filters. C. Containment System Functional Design Functionally, leakage from the primary containment to the annulus between the primary containment and the shield building is processed by recirculating annulus air through either of two redundant fan-filter trains, designated the shield building ventilation system (SBVS), to remove radioactivity prior to exhaust. The vast majority of the leakage that may bypass the annulus through lines penetrating containment is expected to enter the ABSVZ, where it is processed by exhausting zone air through either of two redundant fan-filter trains to remove radioactivity. A minor portion of the total leakage may occur so that it bypasses both the annulus and the ABSVZ. NSP has used conservative values in evaluating shield building annulus air pressure transients for the most severe loss-of-coolant accident (LOCA) and for a range of shield building leak rates (1% per day to-10% per day) assuming only one shield building fan-filter train operates (Reference 1). During the first 3 minutes following a LOCA, l the steel primary containment temperature will increase and the resultant containment expansion and thermal transient will raise the air pressure in the shield building annulus to approximately 7 inches of water above atmospheric pressure. From 3 minutes to 6 minutes after the LOCA, the shield building ventilation system will reduce the annulus air pressure to a negative value relative to air pressure outside the annulus. When the annulus air pressure reaches 2 inches of water below atmospheric pressure, the recirculation damper in the SBVS will start to open. About 24 minutes after the LOCA the air pressure will reach a steady negative value, with the large recircu-lation fan recirculating air through the filters and the small exhaust fan maintaining a negative pressure in the annulus. The fan filter trains in the ABSVZ will start automatically and the normal auxiliary building ventilation system will be isolated auto-matically within 1 minute following the LOCA. The fans in the ABSVZ will maintain a negative pressure in this zone. i i
. TS.5.2-4 REV The NRC has evaluated the LOCA assuming the total primary containment l leakage is 0.15% by weight per day, 0.1 wt%/ day of the primary contain-ment leakage will bypass the annulus through lines penetrating contain-ment and enter the ABSVZ, and that 0.01 wt%/ day of the leakage will bypass both the aanulus and the ABSVZ (References 3, 4, 5, 6). The l remainder is assumed to be recirculated through the shield building ventilation system to remove fission products prior to exhaust from the building.
In the NRC dose calculations (References 3, 4, 5, 6) conservative l mixing assumptions are used. During the first few minutes when the shield building annulus pressure is positive, direct leakage to the environs without mixing or filtration is assumed. During the time the ventilation system is exhausting without recirculation, primary containment leakage is assumed to enter the SBVS inlet duct without mixing and to be filtered before being released to the environs. During recirculation, leakage is assumed to enter the inlet duct without mixing and recirculated air from the outlet duct is assumed to mix in 50% of the annulus volume. The fraction of primary containment leakage that goes to the ABSVZ is assumed to enter the ventilation system ducts without mixing and to be filtered before discharge to the environs, g , . The leakage values used in the NRC staff's dose analyses (References 3, 4, 5, 6) form the bases for the limiting containment system leakage for containment system leakage stest in these Technical Specifications. References
- 1. USAR, Appendix G
- 2. USAR, Sections 12.5, 14
- 3. SER, Section 15
- 4. SER, Supplement 1, issued later
- 5. Letter from AEC to NSP dated November 29, 1973
- 6. Letter from AEC to NSP dated September 16, 1974
TS.S.3-1 REV 5.3 REACTOR A. Reactor Core
- 1. The reactor core contains approximately 48 metric tons of uranium in the form of slightly enriched uranium dioxide pellets. The pellets are encapsulated in Zircaloy-4 tubing to form fuel rods.
The reactor core is made up of 121 fuel assemblies. Each fuel l assembly contains 179 fuel rods (Reference 1). I
- 2. The average enrichment of the reload core is a nominal 2.90 weight per cent of U-235. The highest Uranium-235 loading is a nominal 39 grams of U-235 per axial centimeter of fuel assembly (average).
- 3. In the reactor core, there are 29 full-length RCC assemblies that contain a 142-inch length of silver-indium-cadmium alloy clad with stainless steel (Reference 2). l B. Reactor Coolant System
- 1. The design of the reactor coolant system complies with all appli-cable code requirements (Reference 3). l
- 2. All high pressure piping, components of the reactor coolant system and their supporting structures are designed to Class I requirements, and have been designed to withstand:
- a. The design seismic ground acceleration, 0.06g acting in the horizontal and 0.04g acting in the vertical planes simultane-ously, with stresses maintained within code allowable working stresses,
- b. The maximum potential seismic ground acceleration, 0.12g, acting in the horizontal and 0.08g acting in the vertical planes simultaneously with no loss of function.
- 3. The nominal liquid volume of the reactor coolant system, at rated operating conditions, is 6100 cubic feet.
C. Protection Systems The protection systems for the reactor and engineered safety features are designed to applicable codes, including IEEE-279, dated 1968. The design includes a reactor trip for a high negative rate of change of neutron flux as measured by the excore nuclear instruments (Reference 4). The only one control rod is dropping of more than one control rod (Reference 4). If only one control rod is dropped, the core can be operated at full power for a short time, as permitted by Specifica-tions 3.10. References
- 1. USAR, Section 3.4.2 3. USAR, Table 4.1-11
- 2. USAR, Section 3.5.2 4. USAR, Section 7.1
1 I
, TS.S.4-1 I REV 5.4 ENGINEERED SAFETY FEATURES The engineered safety features include the containment system described in Specification 5.2, the emergency core cooling system, the containment air cooling system, the containment spray system, the post-accident com-bustible gas control system, emergency power supplies, component cooling water system, and the cooling water system. These systems are designed to applicable industry codes, the NRC General Design Criteria in Appendix A to 10CFR50, and NRC Safety Guides. Particular features for the Prairie Island plant include the following:
- 1. Several of the features are shared between the two units, including the onsite diesel generators, the cooling water system, and the motor-driven pumps of the auxiliary feedwater system. Shared systems are designed to mitigate the effects of an accident in one unit and simultaneously provide for a hot shutdown in the other unit (Reference 1).
- 2. The emergency cooling water pumps are driven by diesel engines. These diesel engines are designed and will be tested to the same reliability criteria as those for the diesel generators that supply emergency l electrical power (Reference 2). I e #
- 3. The cooling water system is automatically divided into two (2) redundant loops by motor-operated valves which are actuated by a safety injection signal. Branches serving non-safety related equip-ment in the turbine rooms are isolated from the class I cooling water loops by motor-operated valves. Line breaks in these branches are sensed and the motor-operated valves are actuated by instruments which l monitor coincident high flow and low pressure in the class I cooling water loops.
Reference
- 1. FSAR, Table 1.2-2
- 2. USAR, Section 10.4
a . TS.5.5-1 REV 5.5 RADI0 ACTIVE WASTE SYSTEMS The design objective for the liquid radwaste system is to process the waste so that discharges will approach essentially zero under normal operating conditions and will meet the requirements of 10 CFR Parts 20 and 70, 10 CFR Part 50 Section 50.36a, Appendices A and I to 10 CFR Part 50, and 40 CFR Part 190 under design basis conditions. The design objective for the gaseous waste system is to release a small fraction of 10 CFR Parts 20 and 70, 10 CFR Part 50 Section 50.36a, Appendices A and I to 10 CFR Part 50, and 40 CFR Part 190 limits. Holdup time for the gaseous vaste storage system is designed for the plant lifetime. The design objective of the solid radwaste system is to package solid waste in accordance with applicable government regulations for offsite shipment. A. Accidental Releases The auxiliary building is designed as Class I (seismic) in areas containing the auxiliary building special ventilation system and radwaste storage area. The radwaste building is also designed to Class I standards (Reference 1). All radwaste tanks, filters, and l equipment are either contained in a Class I (seismic) building or in specially constructed areas to provide a substantial degree of control of the wastes should a liquid radwaste tank rupture. B. Routine Releases
- 1. Liquid Wastes The evaluation of the processing of liquid wastes is described in the USAR (Reference 2). The liquid wastes from the non-aerated and aerated wastes systems (including the laundry, hot shower, and decontamination station system) are normally processed through an evaporator (if required), a demineralizer treatment system, and 100% released to the environment. The boron in the reactor coolant in the chemical and volume control system (CVCS) is essentially i recycled. The blowdown treatment system, at a total blowdown rate of 60 to 200 gpm per unit, is treated by demineralization and reclaimed, or discharged. The total activity released is estimated to be 3.2 curies per year (Ci/yr), excluding tritium, and 820 Ci/yr of tritium to be released from the liquid radwaste system for both units. The liquid blowdown system is estimated to release 0.199 !
C1/yr excluding tritium, i i Non-aerated drains from components within the reactor coolant , system and a portion of the coolant letdown stream used for boron l management will be processed through the chemical and volume i control system (CVCS). Aerated drains from the floor drains, aerated equipment drains and leaks, decontamination drains, i
TS.5.5-2 REV l laboratory and sample drains, will be processed through the waste evaporator system or the aerated drain demineralizer treatment l system. The laundry-shower water and certain decontamination solutions will be treated in a special coagulation tank facility. All equipment in the liquid radwaste system is common to both units except the steam generator blowdown flash tanks, reactor coolant drain tanks, and drain tank pumps.
- 2. Gaseous Wastes The evaluation of the gaseous wastes is described in the USAR (Reference 2). The gas decay tanks are designed to hold gases for the plant lifetime, but it was estimated that the release of one decay tank per year would account for unpredictable influent sources. This tank was estimated at 800 Ci/yr of Kr-85 (other isotopes decayed to negligible amounts) for the two units. The estimated gaseous release from containment purging is 478 Ci/yr of Xe-133 for the two units. The total is thus 1,278 Ci/yr for the two units.
A potential source of radioactivity is the atmospheric steam dump , system during large power transients. These. releases are expected to be infrequent and small. Specification 3.9 requires an inved-tory of the activity released from this source to assure compliance with effluent release limits.
- 3. Solid Wastes Miscellaneous materials, such as paper, rags, and glassware, will be compacted for disposal when practical (Reference 2). Spent l resins and concentrates from the waste evaporators will be solidified if required in 55-gallon drums or other regulation containers and l stored in a shielded area prior to shipment offsite for burial.
The total annual shipment is estimated to be 4000 cubic feet of solid or solidified concentrates and miscellaneous materials and 250-350 cubic feet of spent resins for the two units. The storage facilities and packaging and shipping are designed according to 10CFR Part 71 and 49CFR 170-199, C. Process and Effluent Radiological Monitoring System i The process radiation monitoring system is designed to provide informa-tion on radioactive concentrations in certain systems, leakage from one system to another, and radioactive concentrations released to the environment (Reference 3). The monitoring includes: containment or l containment purge vent, shield building vent and auxiliary building vent monitoring separately for particulate, gas and iodine; steam generator blowdown liquid monitor; liquid waste disposal system effluent discharge line monitor; condenser air ejector gas monitor; i
TS.S.5-3 REV radwaste treatment building veut monitor; control room ventilation supply monitor; spent fuel pool air exhaust monitor; residual heat removal cubicle air exhaust monitor; containment fan cooling water and component cooling system monitors; and discharge canal effluent monitor. The area radiation monitoring system is designed to provide informa-tion on radiation levels in various areas of the plant for personnel protection and for qualitative information of a systems condition such as a failed demineralizer. Ten channels monitor areas in the control room, containment, radwaste building, and auxiliary building. The monitoring systems will detect, indicate, annunciate, and/or record the concentrations or levels of activity to verify compliance with 10CFR Part 20 and keep radiation levels as low as practicable, i l i 1
- References
- 1. USAR, Chapter 12
- 2. USAR, Section 9.1.1 l
! 3. USAR, Section 9.2.3 l
. TS.5.6-1 REV 5.6 FUEL HANDLING i
d A. Criticality Consideration The new and spent fuel pit structures are designed to withstand the anticipated earthquake loadings as Class I (seismic) structures. The
- spent fuel pit has a stainless steel liner to ensure against loss of water (Reference 1). l I
The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations. ] The fuel is stored vertically in an array with the center-to-center i distance between assemblies sufficient to assure K <0.95 even if unborated water were used to fill the pit. InaddIb[on5fuelinthe l storage pool shall have a U-235 loading of < 39.0 grams of U-235 per axial centimeter of fuel assembly (average). The criticality considerations as they relate to the dropping of a spent fuel cask (i.e. , heavy load) drop onto the racks has been evalu-
. ated. The maximum K has been calculated to be 0.949 at a water /UO 2
- ratioofa2.0withIboronconcentrationof1800 ppm.
j B. Spent Fuel Storage Structure e e The spent fuel storage pool is enclosed with a reinforced concrete i building having 12- to 18-inch thick walls and roof (Reference 1). l The pool and pool enclosure are Class I (seismic) structures that i afford protection against loss of integrity from postulated tornado missiles. The storage compartments and the fuel transfer canal are j 4 connected by fuel transfer slots that can be closed off with
- pneumatically sealed gates. The bottoms of the slots are above the e
tops of the active fuel in the fuel assemblies which will be stored vertically in specially constructed racks. The spent fuel pool has a reinforced concrete bottom slab nearly 6 feet thick and has been designed to minimize loss of water due to a i dropped cask accident. In addition, the spent fuel cask will have an impact limiter attached or a crash pad will be in place in the pool l which will have the' capability to absorb energy of impact due to a
- l cask drop. This will result in no structural damage taking place to j the pool which would result in significant leakage from the pool.
Piping to the pool is arranged so that failure of any pipe cannot drain the pool below the tops of the stored fuel assemblies. C. Fuel Handling . The fuel handling system provides the means of transporting and i handling fuel from the time it reaches the plant in an unirradiated condition until it leaves after post-irradiation cooling. The system s consists of the refueling cavity, the fuel transfer system, the spent fuel storage pit, and the spent fuel cask transfer system. f __,-_,_..-._m _ _,, _ ,_- . _ , . , _ _ . , _ . _ _ _ _ , _ _ _ . _ . - , . _ . . . __ _ ._. . , , _ . _ _ _ , _ . _ _ , -
l t TS.5.6-2 REV 1 i Major components of the fuel handling system are the manipulation crane, the spent fuel pool bridge, the auxiliary building crane, the fuel transfer system, the spent fuel storage racks, the spent fuel 4 cask, and the rod cluster control changing fixture. The reactor vessel stud tensioner, the reactor vessel head lifting device, and the reactor internals lifting device are used for preparing the reactor for refueling and for assembling the reactor after refueling. i , Upon arrival in the storage pit, spent fuel will be removed from the transfer system and placed, one assembly at a time, in storage racks using a long-handled manual tool suspended from the spent fuel pit ! bridge crane. After sufficient decay, the fuel will be loaded into shipping casks for removal from the site. The casks will be handled by the auxiliary building crane. The load drop consequences of a spent fuel cask for Prairie Island , have been evaluated. It is not possible, due to physical constraints, j for a cask to be dropped into the large pool (pool no. 2). A load path i has been defined which provides for safe movement of the cask. Travel interlocks and mechanical stops prevent cask movement outside of this , path. The only safety-related equipment that can be impacted directly l during a cask drop along this path is the fuel stored in the small pool (pool no. 1). The consequences of this drop have been evaluated and l j' found to meet the NRC staff criteria contained in NUREG-0612 if at ,
- least 50 days have elapsed since reactor shutdown for fission gas release considerations and the pool water contains at least 1800 ppm j boron for criticality considerations. While 50 days was determined adequate, a minimum decay period of 5 years has been incorporated into these technical specifications to provide additional margin in meeting the criteria specified in NUREG-0612 for fission gas releases, while not restricting the plant's operational flexibility. A cask impact
- limiter or crash pad prevents significant structural damage to the pool floor.
The spent fuel cask will be lowered 66 feet from the auxiliary building l to the railroad car for offsite transportation. Specification 3.8 will limit this loading operation so that if the cask drops 66 feet, there will not be a significant release of fission products from the fuel in ,i the cask. i D. Spent Fuel Storage Capacity The spent fuel storage facility is a two-compartment pool that, if ' completely filled with fuel storage racks, provided up to 1582 storage locations. The southeast corner of the small pool (pool no. 1) also j serves as the cask lay down area. During times when the cask is being l used, four racks are removed from the small pool. With the four storage racks in the southeast corner of pool 1 removed, a total of i 1386 storage locations are provided. To allow insertion of a shipping cask, total storage is limited to 1386 assemblies, not including those assemblies which can be returned to the reactor. I j Reference l
- 1. USAR, Section 10.1
. . . . . _ , , _ _ _ _ _ _ . . _ - - . . . _ _ . _ , . _ , . . _ . ~ . _ . , _ , , _ _ _ , _ _ , , _ _ . . , , , . . . . _ , _ _
TS.6.2-6 REV
- f. All events which are required by regulations or Technical Specifi-cations to be reported to the NRC in writing within 24 hours.
- 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.
- 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).
- 5. Authority The OC shall be advisory to the Plant Manager. In the event of a disagreement between the rec'6mmendations of the OC and the ilant 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 General manager Nuclear Plants 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 distribured to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General manager Nuclear Plants and others designated by the OC Chair =an or Vice Chairman.
- 7. Procedures A written charter for the OC shall be prepared that contains:
- a. Responsibility and authority of the group
- b. Content and method of submission of presentations to the Operations Committee
- c. Mechanism for scheduling meetings
- d. Provision for meeting agenda
TS.6.5-1 REV 6.5 PLANT OPERATING PROCEDURES Detailed written procedures, including the applicable checkoff lists and instructions, covering areas listed below shall be prepared and followed. These procedures and changes thereto, except as specified in TS 6.5.F and G, shall be reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager. A. Plant Operations
- 1. Integrated and system procedures for normal startup, operation and shutdown of the reactor and all systems and components involving nuclear safety of the facility.
- 2. Fuel handling operations
- 3. Actions to be taken to correct specific and foreseen potential or actual malfunction of systems or components including responses to alarms, primary system leaks and abnormal reactivity changes and including follow-up actions required after plant protective system actions have initiated.
- 4. Surveillance and testing requirements that could have an ,
effect on nuclear safety.
- 5. Implementing procedures of the Facility Emergency Plan, including procedures for coping with emergency conditions involving potential or actual releases of radioactivity.
- 6. Implementing procedures of emergency plans for coping j with earthquakes and floods. The flood emergency plan shall require plant shutdown for water levels at the site higher than 692 feet above MSL.
- 7. Implementing procedures of the fire protection program. l
- 8. Implementing procedures for the Process Control Program !
and Offsite Dose Calculation Manual including quality control measures. Drills on the procedures specified in A.3. above, shall be conducted as a part of the retraining program. B. Radiological Radiation control procedures shall be maintained and made available to all plant personnel. These procedures shall show permissible radiation exposure and shall be consistent with the requirements of 10CFR20. This radiation protection program shall be organized to meet the requirements of 10CFR20.
TS.6.5-4
- REV E. Offsite Dose Calculation Manual (ODCM)
The ODCM shall be approved by the Commission prior to initial implementation. Changes to the ODCM shall satisfy the following requirements:
- 1. Shall be submitted to the Commission with the Semi-Annual Radio-active Effluent Report for the period in which the change (s) were made effective. This submittal shall contain:
- a. sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with a revision date, together with appropriate analyses or evaluations justifying the change (s).
- b. a determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and
- c. documentation of the fact that the change has been reviewed and found acceptable by the Operations Committee.
- 2. Shall become effe$tive ipon review and acceptanc'e by the Operations Committee.
F. Security Procedures shall be developed to implement the requirements of the Security Plan and the Security Contingency Plan. These implementing procedures, with the exception of those non-safety related proce-dures which govern work activities exclusively applicable to or performed by security personnel, shall be reviewed by the Opera-tions Committee and approved by a member of plant management designated by the Plant Manager. Security procedures not reviewed by the Operations Committee shall be reviewed and approved by the Supervisor, Security and Services. G. Temporary Changes to Procedures Temporary changes to Operations Committee reviewed procedures described ia A,B,C,D,E and F above, which do not change the intent of the original procedure may be made with the concurrence of two individuals holding senior operator licenses. Such changes shall be documented, reviewed by the Operations Committee and approved by a member of plant management designated by the Plant Manager within one month. Temporary changes to security procedures not reviewed by the Operations Committee shall be reviewed by two (2) individuals knowledgeable in the area affected by the procedure.
TS.6.6-2 REV Plant radiation and contamination surveys.
- 7. Changes made to the plant as it is described in the Final Safety Analysis Report, reflected in updated, corrected and as-built drawings.
- 8. Cycling beyond normal limits for those components that have been designed to operate safely for a limited number of cycles beyond such limits.
- 9. Reactor coolant system in-service inspections.
s 10. Minutes of meetings of the Safety Audit Committee.
- 11. Records of the service lives of all safety-related snubbers, s including the date at which the service life commences and associated installation and maintenance records.
e i
l
- TS.6.7-2 REV
- 2. Occupational Exposure Report. An annual report of occupational exposure covering the previous calendar year shall be submitted prior-to March 1 of each year.
The report should tabulate on an annual basis the number of station, utility and other personnel (including con-tractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g., reactor opertions and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling. The dose assignment to various duty functions may be estimates 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 shall be assigned to specific major work functions.
- 3. Monthly Operating Report. A monthly report of operating statistics and shutdown experience covering the previous month shall be submitted by the 15th of the following month to the Director, Office of Resource Management, US Nuclear Regulatory Commission, Washington, DC 20555.
t # (1) This report supplements the requirements of 10CFR20, section 20.407. If 10CFR20, Section 20.407 is revised to include such information, this Specification is unnecessary.
1 Exhibit D License Amendment Request Dated July 15, 1986 5.0 RADIATION ENVIRONMENTAL MONITORING PROGRAM 5.1 Samoling Table 5.1-1 and Figures 5.1-1, 5.1-2 and 5.1-3 specify the current sampling locations based on the latest land use census. If it is learned from an annual census, that milk animals or gardens are present at a location which yfelds a calculated, thyroid dose greater - than those previously sampled, the new milk animal or garden locations resulting in higher calculated doses shall be added to the surveillance program as soon as practicable. Sample locations (except the control) having lower calculated doses may be dropped from the program at the end of the growing season (October 31) to keep the total number of sample locations constant. The plant routinely discharges liquid radioactive waste into the Mississippi River. An annual land use survey will be conducted to determine whether any crops are irrigated with water taken from the Mississippi River between the plant discharge canal and a point 5 miles downstream. If edible crops are being irrigated from Mississippi water, appropriate samples will be collected and analyzed per Technical Specification Table TS4.10-1. 5.2 Interlaboratory Comoarison prooram Analyses shall be perfomed on radioactive samples supplied by the EPA crosscheck program. This program involves the analyses of samples provided by a control 'aboratory and compariton of results with those of the control laboratorj, as well as with other laboratories which receive portions of the same samples. Media used in this program (air, milk, water, etc.) shall be limited to those found in the radiation environ-mental monitoring program. The results of analyses performed as a part of the crosscheck program shall be included in the Annual Radiation Environmental Monitoring Report. o,.. .
\
s 5-1 Rev
s LICENSE AMENDMENT REQUEST DATED JULY 15, 1986 EXHIBIT E Revised Assessment of Pressurized Thermal Shock Reference Temperature in Accordance with 10 CFR Part 50 Section 50.61. Prairie Island Units No. I and No. 2 Ref: (a) Letter dated January 10, 1986, D M Musolf, NSP, to Director of NRR, NRC, " Assessment of Pressurized Thermal Shock Reference Temperature in Accordance with 10 CFR Part 50, Section 50.61" (b) Letter dated April 25, 1986, D M Musolf, NSP, to Director of NRR, NRC, " Summary Technical Report of Analysis of Capsule from Unit 1 Reactor Vessel Radiation Surveillance Program" Reference (a) provided projected values of pressurized thermal shock reference temperature (RTPTS) at the inner vessel surface of reactor beltline materials at Prairie Island Nuclear Generating Plant Units No. I and No. 2. Reference (b) is the summary report of the analysis of the third surveillance capsule removed from the reactor vessel of Unit No. 1. The analysis was performed by the Westing-house Eleqttric Corporation. The cakculated reactor vessel wall fast neutron fluence increased approximately 21% over earlier values due to improved transport methodology as described in Section 6 of the report. Wall fast neutron fluence values for Unit No. 2 are the same as Unit No. I since the vessel designs and fuel loading patterns are the same. The purpose of this report is to revise the Prairie Island projected values of pressurized thermal shock reference temperature at the inner vessel surface to reflect the 21% increase in calculated fluence. Calculations originally performed and described in Reference l (a) have been revised using the higher wall fluence reported l in Reference (b). The following revised figures and tables f are attached: l ! 1) Revised Fast Neutron Flux Radial Distribution Page 3-1 of Reference (a)
- 2) Revised RTPTS Values for 60 EFPYs
( Tables 1 and 2 of Reference (a) l
- 3) Revised Plots of RTPTS vs. EFPY l Figures 1 and 2 of Reference (a) -
l l 1
EXHIBIT E The revised RTPTS values for Prairie Island Units No. I and No. 2 are as follows: Time Unit No. 1 Unit No. 2 15 EFPY 141 F 171 F License 1b4 F 189 F Expiration 60 EFPY 178 F 222 F Calculations were performed for 15 effective full power years (EFPY) since this number corresponds to the EFPY used in the proposed heatup and cooldown curves. Estimates of RTPTS for the end of the current operating license (June 25, 2008 for both units) are based on an estimated 25.9 EFPY for Unit No. 1 and 26.4 EFPY for Unit No. 2. Calculations have been performed for 60 EFPY to conservatively bound any plant life extension plans. All values of RTPTS remain well below the screening criteria established in 10 CFR Part 50, Section 50.61, of 270 F.for forgings and 300 F for circumferential welds. e i
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. EXHIBIT E 10 3I -
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, Calculated Azimuthal Cistribution of Maximum Fast (E>1.0 Mev) ! Neutron Flux Within tiie Reac*or Vessel Surveillance l Caosute Geometry 1650 MWt 1 e l l
1 ISLAND PTS SCO E E tt l NG - IOCFR50.62 ? PGAIRIE-UN11 1 60 YEARS OF OPERATION 4 _10_ . . _ _ . _ _ _D_ E_S_ C_ . _ R_ I P_ T .I O tt . . . . - _ _ _ . . _ _ ._H_AR_G 2_. ..E.ON_2__. __ _ . R. 's_ P_ _ T S_ _ _. H_D_ T .S_R_C_ _ W / O_ _C O.P.P.E _R. _ H_ W /A_ O R_ G_ _ _IC_K_L_ _ _ _ _ EE_ O_ N_ _ _1_ _ . i _ _ _ F_ L_ U_ _X _..F.L_U_
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H.D _ __TT __ .._ _9_4_5 _ FORG 4.80E401 1 24E900 0.0 4.54E*02 1.24E+02 I INTER SHELL9.75E+jHG s9 1.40E901 MEASURED 6.00E-02 7 20E-01 5.ISE+10 1.16E902 0.0 4.36E402 1.56E+02 2 LOWER SHELL 9.75E689 FORCING -4.00E600 MEASURED 7.00E-02 6.60E-01 4.80E+01 5.15E+30 1.65E+02 3.40E401 3.85E902 1 65E90: 3 NOZ-INTER SHELL 3.03E+39 WELD 0.0 2269/1580 GEHERIC 1 70E-On 3.50E-On 5.90E901 1.60Eelo 3.40E401 3.8SE+02 1.23E+02 4 HOZ-INTER SHELL 3.03E+39 WELD 0.0 3049/1880 GENERIC I.35E-05 8.00E-02 5.90E901 1.23E+02 1.60E+Io 3 40E+0! 4.74E902 1 78E+02 5 INTER-LOW SHELL 9.75E619 WELD 0.0 3752/1230 GENERIC 3 40E-01 3 70E-01 5.90E901 1 78E402 5.35E610 3 46E902 3.40E401 4.74E402 1 46E402 6 INTER-LOW SHELL 9.75E+39 WELD 0.0 3049/1230 GENERIC 1.35E-OS 8.00E-02 5.90E+0! 5.35E+10 I g ISLAND PTS SCREENING - 10CFR50.42 V 03 PRAIRIE-UNil 2 60 YEARS OF OP E R A T I Ott H ' d EO _ _R_ T P_ _T S_ . . I _I D_. _ _ _ _ _ _D_ E_ S_ C_ R_ .I P. _T _l o_ t_a _ _ - _ _ _ _ _ _ _ _ _ M. 1_4_/ O_ _H_.I CK.L_E_ . A.
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. - N_ - _T 1.27E402 0.0 4.36E402 1.27E402 1 INTER SHELL9.75E+19 FORGING -4.00E600 t1E ASUR E D 7.50E-02 7.50E-On 4.80E401 5.ISE+10 1.36E*02 0.0 4.34E402 1 36E+02 2 LOWER SHELL FORGING ~6.00E+00 t1EASURED 8.50E-02 7.00E-05 4.80E401 5.15E+IO 9.75Ee19 1 44E*02 3 40E901 3.85E+02 1 44E902 3 NOZ-INTER SHELL WELD 0.0 1752/1263 GENERIC 1.40E-03 1.40E-01 5.90E601
- 1.60E+IO 3.03E+I9
- 1 78E+02 3.40E+0! 3.85E402 8.78E902 i
4 HOZ-INTER SHELL WELD 0.0 3049/1263 GENERIC 1.90E-03 1.30E-01 5.90E601 1.60E630 3.03E+I9 5.90E601 1.26E602 3.40E908 4.74E902 1.26E+02 5 INTER-LOW SHELL 9.75E+19 WELD0.0 2721/1263 GENERIC 9.00E-02 1.30E-01 5.15E+30 5.90E401 2.22E402 3.40E903 4 74E902 2 22E902
& INTER-LOW SHELL 9.75E+I9 WELO0.0 3049/1263 GENERIC 1.90E-On I.30E-Ol i 5.15E+10 N
EXHIBIT E 10 CFR PRRT 50, SECTION 50.61, SCREENING PRRIRIE ISLRNO UNIT = 1 o. 8 n c SCREENING CRITERION - 300 DEGREES F
$- LEGEND o - INTER SHELL FORGING C. c - LOWER SHELL FORGING 5- + - N0Z-INTER SHELL WELD 22S9/1180 x - N0Z-INTER SHELL WELO 3049/1180 o - INTER-LOW SHELL WELO 1752/1230 7 7- INTER-LOW SHELL WELD 3049/1230 0
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EylilBIT E 10 CFR PART 50, SECTION 50.61, SCREENING PRRIRIE ISLRNO UNIT = 2 C. 8 m a. SCREENING CRITERION - 300 DEGREES F IC- LEGEND o - INTER SHELL FORGING C.
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