ML20044H485
| ML20044H485 | |
| Person / Time | |
|---|---|
| Site: | Clinton  |
| Issue date: | 06/03/1993 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20044H481 | List: |
| References | |
| NUDOCS 9306090082 | |
| Download: ML20044H485 (7) | |
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INQEX BASES SECTION EAS1 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION.................
B 3/4 3-1 3/4.3.2 CONTAINMENT AND REACTOR VESSEL ISOLATION CONTROL SYSTEM...
B 3/4 3-2 i
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION...
B 3/4 3-3 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION......... B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION.
B 3/4 3-4 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION.........................
B 3/4 3-5 3/4.3.7 MONITORING INSThdMENTAT10N Radiation Monitoring Instrumentation......................
B 3/4 3-5 Seismic Monitoring Instrumentation........................
B 3/4 3-5 Meteorological Monitoring Instrumentation.................
B 3/4 3-5 Remnte Shutdown Monitoring Instrumentation................
B 3/4 3-6 Accident Monitoring Instrumentation.......................
B 3/4 3-6 Source Range Monitors...............................
B 3/4 3-6 Traversing in-Core Probe System...........................
B 3/4 3-6 Chlorine Detection System.
B 3/4 3-7 i
Loose-Part Detection System.............................
B 3/4 3-7 9306090082 930603 PDR ADOCK 05000461 P
PDR CLINTON - UNIT 1 xvii Amendment No. 6B DEC 2 41992
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INDEX BASES SECTION PAGE' INSTRUMENTATION (Continued)
Main Condenser Offgas Treatment System Explosive Gas
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Monitoring Instrumentation..............................
B 3/4 3-8 3/4.3.8 DELETED............................................
B 3/4 3 :
3/4.3.9 PLANT SYSTEMS AC-'JATION INSTRUMENTATION...............
B 3/4 3-8 N' CLEAR SYSTEM PROTECTION SYSTEM - SELF TEST SYSTEM....
B 3/4 3-9 3/4 3.10 J
Bases Figure B 3/4.3-1 Reactor Vessel Water Level.................
B 3/4 3-10 3/4.4 REACTOR COOLANT SYSTEM i
3/4.4.1 RECIRCULATION SYSTEM..................................
B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES...........................
B 3/4 4-3 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE t
Leakage Detection Systems..........................
B 3/4.4-4 Op e rati on al Le ak age...................................
B 3/4 4-4 3/4.4.4 CHEMISTRY.....................................
B 3/4 4-4a 3/4.4.5 SPECIFIC ACTIVlTY...................................
B 3/4 4-5 3/4.4.6 PRESSURE / TEMPERATURE LIMITS............................
B 3/4 4-6 l
3/4.4.7 MAIN STEAM LINE ISOLATION VALVES......................
B 3/4 4-7 l
3/4.4.8 STRUCTURAL INTEGRITY........................
B 3/4 4-7 3/4.4.9 RESIDUAL HEAT REM 0 VAL..........................
B 3/4 4-7 Bases Table B 3/4.4.6-1 Reactor Vessel Toughness Values...........
B 3/4 4-8 i
Bases figure B 3/4.4.6-1 Fast Neutron Fluence (E>l Mev) at I.D. surface as a Function of Service Life...........
B 3/4 4-10 Figure B 3/4.4.6-2 DELETED........................................
B 3/4 4-11 CLINTON - UNIT 1 xviii revised by letter' dated 6/3/93-I f
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REACTOR C00LAf1T SYSTEM
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I BASES 3/4.4.5 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant easure that the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thyroid and whole body doses resulting from a main steam line failure outside the containment during steady state operation will not exceed small-fractions of the dose guidelines of 10 CFR 100.
The values for the limits on specific activity represent interim limits based upon a parametric evaluation by the 11RC of typical site locations.
These values are conservative in that specific site parameters, such as site boundary location i
and meteorological conditions, were not considered in this evaluation.
The ACT10f1 statement permitting POWER OPERATI0t1 to continue for limited time l
periods with the pr' nary coolant's specific activity greater than 0.2 micro-curies per gram DOSE EQUIVALEt1T I-131, but less than or equal to 4.0 microcuries per gram DOSE EQUIVALEt1T I-131, accommodates possible iodine spiking which may v cur following changes in THERMAL POWER.
Closing the main steam line isolation valves prevents the release of activity to the environs should a steam line rupture occur outside containment.
ACT10f1 statement C requires a reactor coolant sample be taken and an isotopic analysis for iodine be performed if, during steady state operation, offgas levels increase by more than 10,000 microcuries per second in one hour at release rates less than 75,000 microcuries per second, or increase by more than 15% in one hour at release rates greater than 75,000 microcuries per second.
These required isotopic analyses are intended to support determination of the cause for the increase in offgas radiation levels, such l
as the onset of leakage from a fuel pin (s) which will lik 'y result in an increase in the reactor coolant specific activity.
However, several evolutions have been identified which result in a predictable, known and i
temporary increase in the indicated offgas levels such that the indicated levels may exceed those specified in the action statement.
These evolutions are placing a condensate polisher in service, temporarily turning the offgas pre-treatment monitor sample pump off, swapping steam jet air ejectors and regenerating the offgas desiccant dryer (s). Although the noted evolutions have no effect on reactor operation, they do affect " steady-state operation" of the offgas or offgas process radiation monitoring system with respect to Technical Specification 3/4.4.5.
These evolutions do not require an offgas i
sample to be taken.
However, it is prudent to verify that the offgas pre-i treatment process radiation monitor readings return to expected levels within four hours following the identified evolutions (or as soon as possible l
following the steam jet air ejector swap or desiccant dryer regeneration process).
This will confirm that there were no other causes for the indicated increase in the radioactivity rate.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.
CLIllT0!1 - UtilT 1 6 3/4 4-5 revised by letter dated 6/3/93 m
i REACTOR COOLANT SYSTEM BASES 3/4.4.6 PRESSURE / TEMPERATURE LIMITS All components in the reactor coolant system are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations.
The various categories of load cycles used for design purposes are provided in Section 3.9 of the USAR.
During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.
The operating limit curves of Figure 3.4.6.1-1 are derived from the fracture toughness requirements of 10 CFR 50 Appendix G and ASME Code Section Ill, Appendix G.
The curves are based on the RT and stress intensity factor-uoy information for the reactor vessel components.
Fracture toughness limits and the basis of compliance are more fully discussed in USAR subsection 5.3.1.5 entitled " Fracture Toughness."
The reactor vessel materials have been tested to determine their initial RT,o7 The results of these tests are shown in Table B 3/4.4.6-1.
Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT,37 of the core beltline region.
Therefore, an adjusted refe,ence temperature, based upon the fluence, nickel content and copper content of the material in question, can be predicted using Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.
The pressure / temperature limit curve, figure 3.4.6.1-1, curves A, B, and C, includes an assumed shift in RT for the conditions at 12 Effective Full l
ya7 Power Years.
The actual shift in RT, of the vessel material will be estab-lished periodically during operation by removing and evaluating, in accordance with ASTM E185 and 10 CFR 50, Appendix H, irradiated reactor vessel material specimens installed near the inside wall of the reactor vessel in the core area.
The irradiated specimens can be used to predict reactor vessel material transition temperature shift.
Flux wires which were removed after the first fuel cycle and will be removed at later intervals with the surveillance speci-mens are analyzed and provide an improved neutron fluence estimate for the reactor vessel.
This data is then used te modify Bases Figure B 3/4.4.6-1 and predictions of reactor vessel material tr sition temperature. shift per Regu-latory Guide 1.99, Revision 2. The optr ig limit curves of Figure 3.4.6.1-1 have been and will be adjusted, as reqc ed, on the basis of the specimen data and the recommendations of Regulatory Guide 1.99, Revision 2.
The pressure-temperature limit lines shown in Figures 3.4.6.1-1, curves C and A for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50 for reactor criticality and for inservice leak and hydrostatic testing.
CLINTON - UNIT 1 B 3/4 4-6 revised by letter dated 6/3/93
c.
REACTOR COOLANT SYSTEM BASES 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES Double isolation valves are provided on each of the main steam lines to minimize the potential leakage paths from the containment in case of a line break. Only one valve in each line is required to maintain the integrity of the containment; however, single failure considerations require that two valves be OPERABLE.
The surveillance requirements are based on the operating history of this type valve. The maximum closure time has been selected to contain fission products and to ensure the core is not uncovered following line breaks. The minimum closure time is consistent with the assumptions in the safety analyses to prevent pressure surges.
3/4.4.8 STRUCTURAL INTEGRITY The inspection programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant.
Components of the reactor coolant system were dea igned to provide access to permit inservice inspections in accordance with Sectian XI of the ASME Boiler and Pressure Vessel Code 1975 Edition and Addenda through Winter 1975.
The inservice inspection program for ASME Code Class 1, 2 and 3 components will be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the NRC pursuant to 10 CFR Part 50.55a(g)(6)(i).
3/4.4.9 RESIDUAL HEAT REMOVAL A single shutdown cooling mode loop provides sufficient heat removal capability for removing core decay heat and mixing to assure accurate temperature indication; however, single failure considerations require that two loops be OPERABLE or that alternate methods capable of decay heat removal be demonstrated and that an alternate method of coolant mixing be in operation.
CLINTON - UNIT 1 B 3/4 4-7 revised by letter dated 6/3/93 e
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PLANT SYSTEMS BASES 3'4.7.4 SNUBBE;S (Continued) etc.
The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.
These records will provide statistical bases for future consideration of snubber service life.
The requirements for the maintenance of recorc's and the snubber service life review are not intended to affect plant operation.
3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
This limitation will ensure that leakage from byproduct, source, and special nuclear material sources will not exceed allowable intake values.
Sealed sources are classified into three groups according to their use, with surve11-lance requirements commensurate with the probability of damage to a source in that group.
Those sources which are frequently handled are required to be tested more often than those which are not.
Sealed sources which are continu-ously enclosed within a shielded mechanism, i.e., sealed sources within radia-tion monitoring devices, are considercd to be stored and need not be tested unless they are removed from the shielded mechanism.
3.4.7.6 MAIN TURBINE BYPASS SYSTEM The main turbine bypass system is required to be OPERABLE consistent with the assumptions of the feedwater controller failure analysis in FSAR Chapter 15, 3/4.7.7 L10UID STORAGE TANKS The tanks listed in this Specification include all tnose outdoor storage tanks that art not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.
Restricting the quantity of radioactive material contained in each of the specified tanks provides assurance that in the event of an uncontrolled release of the contents from any of these tanks, the.resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 2, l
Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.
l CLINTON - UNIT 1 B 3/4 7-3 Amendment No. 69 MAR 2 91993
PLANT SYSTEMS a
BASES 3/4.7.8 MAIN CONDENSER OFFGAS MONITORING 3/4.7.8.1 0FFGAS - EXPLOSIVE GAS MIXTURE Although there should normally be more than sufficient steam flow to the steam jet air ejectors to ensure adequate dilution of hydrogen (and thus prevent the offgas from attaining hydrogen levels in excess of the flammability limit),
this specification is provided to ensure that the ctncentration of potentially explosive gas mixtures contained in the offgas holdtp system is monitored and maintained below the flammability limit of hydrogen.
Maintaining the concentration of hydrogen below its flammability limit provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.
3/4.7.8.2 0FFGAS - NOBl.E GAS RADI0 ACTIVITY RATE Restricting the gross radioactivity rate of noble gases from the main condenser provides reasonable assurance that the total body exposure to an individual at the exclusion area boundary will not exceed a small fraction of the limits of 10 CFR Part 100 in the event this effluent is inadvertently discharged directly to the environment without treatment.
This specification implements the requirements of General Design Criteria 60 and 64 of Appendix A to 10 CFR Part 50.
The surveillance requirements provide adequate assurance that changes in the gross radioactivity rate of noble gasses from the main condenser are identified and monitored to ensure the Technical Specification limits are not exceeded.
The Technical Specification requires continuous monitoring of the offgas recombiner effluent, as well as periodic isotopic analysis (at least once per 31 days) of a representative sample of gasses taken at the discharge of the offgas recombiner.
In addition, an isotopic analysis must also be performed within four hours following an increase in the nominal steady state fission gas release of greater than 50%, after factoring nut increases due to changes in thermal power levels.
The required isotopic analysis is intended to support determination of the cause for the increase in offgas radiation levels, such as the onset of leakage from a fuel pin (s) which will likely result in an increase in the reactor coolant specific activity.
- However, there are two evolutions, swapping of the steam jet air ejectors and regeneration of the offgas system desiccant dryers, which are known to result in a predictable and temporary increase in the indicated offgas radioactivity rate.
Since the increase is due to an evolution (s) known to cause such an increase and not due to an actual increase in "the nominal steady state fission gas release from the primary coolant", isotopic analysis of a gas sample is not required for these two specific evolutions.
However, it is prudent to ensure that the offgas radiation level (radioactivity rate) returns to previous or expected levels within four hours or as soon as possible following the swap or regeneration process.
This will confirm that there are no other causes for the indicated increase in the radioactivity rate.
CLINTON - UNIT 1 B 3/4 7-4 revised by letter dated 6/3/93
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