ML20024A014
| ML20024A014 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 06/08/1983 |
| From: | DAIRYLAND POWER COOPERATIVE |
| To: | |
| Shared Package | |
| ML20024A012 | List: |
| References | |
| NUDOCS 8306150283 | |
| Download: ML20024A014 (51) | |
Text
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e 4.0 OPERATING LIMITATIONS 4.0.1-DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved. When these terms appear in capitalized type, the following definitions apply in these Technical Specifications.
ACTION ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
AVERAGE PLANAR EXPOSURE The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
AVERAGE PLANAR LINEAR HEAT GENERATION RATE The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the. number of fuel rods in the fuel bundle.
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
8306150283 830608 PDR ADOCM 05000409 P
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4.0 ~ OPERATING LIMITATIONS 4.0 1 DEFINITI0liS - (C_ont'd) 2 DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 pCi/ 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.
E - AVERAGE DISINTEGRATION ENERGY E shall be the average, weighted in proportion to the concentration 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.
EFFLUENT RELEASE BOUNDARY The Dairyland Power Cooperative property line within the 1109 ft. radius Exclusion Area is the EFFLUENT RELEASE BOUNDARY. See Figure 4/5.3.
FREQUENCY NOTATION The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table of Surveillance Frequency Notation.
GASEOUS RADWASTE SYSTEM A GASEOUS RADWASTE SYSTEM is a system designed to reduce radioactive gaseous effluents by collecting primary coolant offgases from the primary system and providing for delay,. holdup or filtering for the purpose of reducing the total radioactivity prior to release to the environment.
IDENTIFIED LEAKAGE t
I IDENTIFIED LEAKAGE shall be:
i Leakage into collection systems, such as pump seal or valve packing a.
l leaks, that are captured and conducted to a sump or collecting tank, l.
b.
Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY j
LEAKAGE.
LACBWR WP-4.12 27f
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4.0J0PERATING LIMITATIONS
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LIMITING CONTROL ROD PATTERN A LIMITING CONTROL ROD PATTERN shall be a pattern which results in the core being on a. thermal hydraulic limit, i.e.,
operating on a limiting value for AFLHGR, LHGR, or MCPR.
LINEAR HEAT GENERATION RATE LINEAR HEAT GENERATION RATE (LHGR) shall be the power generation in an arbitrary length of fuel rod, usually one foot. It is the integral of the heat flux-over.the heat transfer area associated with the unit length.
MEMBER (S) 0F THE PUBLIC MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally 4
associated with the utility. This category does not include employees of the utility, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or make deliveries. This category does include persons who use portions of the site for recreatioaal or other purposes not associated with the utility.
MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which exists in the core.
OFFSITE DOSE CALCULATION MANUAL (ODCM)
An OFFSITE DOSE CALCULATION MANUAL (0DCM) shall be a manual containing the methodology and parameters to be used for the calculation of offsite doses due to radioactive gaseous and liquid effluents and for the calculation of gaseous and liquid effluent monitoring instrumentation alarm / trip setpoints. It shall describe ~the radiological environmental monitoring program.
OPERABLE-OPERABILITY A system, subsystem, train, component or device shall be OPERABLE or have
' OPERABILITY when it is capable of performing its specified function (s) and when all necessary attendant instrumentation, controls, a normal and an emergency electrical power source, cooling or seal water, lubrication or other l
auxiliary equipment that are required for the system, subsystem, train, f
component or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL CONDITION - CONDITION An OPERATIONAL CONDITION, i.e. CONDITION, shall correspond to any one inclusive combination of power level and average reactor coolant temperature specified in Table of OPERATIONAL CONDITIONS.
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4.0 OPERATING LIMITATIONS 4.0.1 DEFIN1TIONS - (,C,o,nt,'d) _ _, _ _ _ _
n PARTIAL SCRAM A PARTIAL SCRAM signal shall cause the electric and hydraulic scram motors for 13 preselected control rod drive mechanisms to be actuated for control rod insertion. Full insertion of PARTIAL SCRAM control rods during POWER OPERATION shall render the reactor suberitical.
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 13 of the Safeguards Report, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.
PRESSURE SOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.
The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated solid wastes will be accomplished in such a way as to assure compliance with.10 CFR Part 20, 10 CFR Part 71 and Federal and State cegulations and other requirements governing the disposal of the radioactive waste.
RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant and reactor components of 165 MWt.
REPORTABLE OCCURRENCE A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specification 6.9.1.7 of Technical Specifications.
PESTRICTED AREA A RESTRICTED AREA shall be any area within the exclusion boundary, access to which is controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials. See Figure 4/5.3.
SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical from its present condition assuming all control rods are fully inserted, except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, i.e. < 80*F, and Xenon free.
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4.0 OPERATING LIMITATIONS
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O SOLIDIFICATION SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping cnd burial ground requirements.
SOURCE CHECK A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of:
a.
A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals.
b.
The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant and reactor components.
UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.
UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area not controlled by the licensee for purposes of protection of MEMBERS OF THE PUBLIC from exposure to radiation and radioactive materials.
VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXilAUST TREATMENT SYSTEM is any system designed and installed to reduce radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through High Efficiency Particulate filters for the purpose of removing particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents.
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4.0 OPERATING LIMITATIONS 4.0.1 DEFINITIONS - (Cont'd) i SURVEILLANCE FREQUENCY NOTATION NOTATION FREQUENCY S
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
t D
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
W At least once per 7 days.
i M
At least once per 31 days.
Q At least once per 92 days.
SA At least once per 6 months.
A At least once per 12 months.
R At least once per 18 months.
S/U Prior to each reactor startup.
P Completed prior to each release.
N.A.
Not applicable.
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I LACBWR WP-4.12 (Pages 271 - 27x)
4.2.6.7 During operation with the nuclear instrumentation channels in 2 of 4 trip logic, at least three of nuclear channels 5, 6, 7 and 8, including their
~
automatic gain control subsystem channels, shall be OPERABLE. If nuclear channel 5 or 6 is inoperable, its scram contacts shall be placed in the trip position. If power is de-escalated, the tripped channel's output shall be bypassed prior to entering 1 of 2 logic, subject to Section 4.2.6.1 require-ments. If nuclear channel 7 or 8 is inoperable in a manner affecting the operability of its corresponding power-flow chancel, the power-flow channel shall be bypassed, pursuant to the time limitations of Section 4.2.6.1, and the scram contacts of the nuclear channel shall be placed in the trip position.
4.2.6.8 Safety channels directly backed up by an identical channel or channels may be bypassed for maintenance or testing. Safety channels in the partial scram circuit may be bypassed for maintenance or testing for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.2.6.9 Both reactor forced circulation pumps shall be automatically shut down by a high reactor pressure signal or by a low reactor water level signal.
Bases - The RPTS is a diverse and independent backup except for common current sensing loops to the normal scram system for rapid chutdown of the reactor.
To protect the primary system from an ATWS event in which either'MSIV closes at power, thus eliminating the main condenser as a heat sink, the recirculation pumps must be shut down to prevent damage to the primary system.
A rapid shut down of the recirculation pumps has the effect of causing an increase in the moderator voids in the reactor core. A substantial negative reactivity results and the power and pressure surges that might otherwise occur in the most limiting transient (MSIV closure) are substantially reduced.
With the recirculation pumps shut down, the reactor power will be reduced to a.
i steady state power level of less than 20% (based on natural circulation through the core).
4.2.7
- Deleted -
4.2.8 Spent Fuel Storage and Handling 4.2.8.1 Fuel elements and control rods shall be inserted or removed from the reactor vessel one at a time.
4.2.8.2 Irradiated fuel elements shall be stored underwater in spent fuel l
storage racks that are positioned on the bottom of the spent fuel storage well, or in an approved shipping cask.
4.2.8.3 During the handling of irradiated fuel elements that have been operated at power levels greater than 1 Mwt the depth of water in the reactor upper cavity and/or the spent fuel storage well shall be at least 2 ft above the active fuel.
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4.2.8.4 Irradiated fuel elements shall have decayed for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to placing them in the spent fuel storage well.
4.2.8.5 With the exception of a spent fuel shipping cask, the core spray bundle, the transfer canal shield plug and the other components and fixtures that are normally located and used within the spent fuel storage well, no objects heavier than a fuel assembly shall be handled over the spent fuel storage well.
i LACBWR WP-4.12 37
4/5.3 RADIOACTIVE EFFLUENTS 4/5.3.1 RADIOACTIVE LIQUID EFFLUENTS INSTRUMENTATION
]J_HITING COND1T10LF_0JL0f_CMT_1.0N 0
4.3.1.1 The following radioactive liquid effluent monitoring instrumentation channels shall be OPERABLE:
a.
Liquid Radwaste Effluent Line Monitor or Turbine Condenser Cooling Water Monitor, and b.
Liquid Radwaste Effluent Line Flow Meter APPLICABILITY: At all times when releasing liquid radioactive effluents.
ACTION:
With neither activity monitor OPERABLE, suspend release of liquid a.
radioactive effluent without delay. Effluent releases may be resumed provided that at least two independent samples are analyzed and that at least two technically qualified members of the staff independently.
verify the release rate calculations.
If channels are not operable for more than 30 continuous days, explain in the next Semi-Annual Effluent Report pursuant to Specification 6.9.3.a.
- b. 'With the flow meter not OPERABLE, effluent releases via this pathway may continue provided the flow rate is estimated at '. east once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> during actual releases. Pump curves may be used to estimate flow.
c.
The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1 are not applicable.
S.gR{ELLLANC_E_R,EAUI_RE_M_ERS_
5.3.1.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown on Table 5.3.1.1.
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m TABLE 5.3.1.l' RADIOACTIVE LIQUID EFFLUENT MONITORING' INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL SURVEILLANCE CHANNEL SOURCE FUNCTIONAL CHANNEL REQUIREMENT INSTRUMENT CHECK CHECK TEST CALIBRATION CONDITIONS a.
Liquid Radwaste Effluent Line Monitor P
M*
Q(1)
R(3) b.
Turbine Condenser Cooling Water Monitor D
M*
Q(1)
R(3) c.
Liquid Radwaste Effluent Line Flow Meter D(2)
NA NA R(4)
- Background radiation may be used as source check.
- During applicable conditions.
(1). The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist:
a.
Instrument indicates measured levels at the alarm setpoint.
~
b.
Instrument indicates a downscale (circuit failure) failure.
(2) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on days in which continuous, periodic, or batch releases are made.
(3) The CHANNEL CALIBRATION shall include the use of a known liquid radioactive source positioned in a.
reproducible geometry with respect to the sensor and emitting gamma radiation with the fluences and energies in the ranges measured by the channel during normal operation.
(4) The CHANNEL CALIBRATION will be in accordance with the manufacturer's recommended procedure.
LACBWR WP-4.12 39
O RADIOACTIVE LIQUID EFFLUENTS CONCENTRATION LIMITING CONDITION FOR OPERATION 4.3.1.2 The concentration of radioactive material released in liquid effluents at any time to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited to the concentrations specified in 10 CFR Part 20, Appendix B, Table II, Column 2, issue of December 30, 1930, for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10-4 pCi/ml total activity.
APPLICABILITY: At all times.
ACTION: With the concentration of radioactive material released to UN-RESTRICTED AREAS exceeding the above limits, without delay restore concentration to within the above limits.
The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE RE,Q,U,IREMENTS 5.3.1.2 The radioactivity content of each batch of radioactive liquid waste to be discharged shall be determined prior to release by sampling and analysis in accordance with Table 5.3.1.2. The results of pre-release analyses shall be used to assure that the concentration at the point of release is maintained within the limits of Specification 4.3.1.2.
RADIOACTIVE LIQUID EFFLUENTS DOSE M I,TJNG CONDITION FOR OPERATION _,
4.3.1.3 The dose or dose commitment to a KEMBER OF THE PUBLIC from radioactive materials in liquid effluents released to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited:
a.
During any calendar quarter to f 1.5 mrems to the total body and to f 5 mrems to any organ, and b.
During any calendar year to 1 3 mrems to the total body and to f 10 mrems to any organ.
APPLICABILITY: At all times.
ACTION: With the calculated dose from the release of radioactive materials in liquid effluents exceeding the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to assure that subsequent releases shall be in compliance with the above limits.
The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE REQQIREMENTS 5.3.1.3 Dose Calculations:
Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the Offsite Dose Calculation Manual (0DCM) at least once per 31 days.
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g TABLE 5.3.1.2 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LIQ'JID HINIMUM TYPE OF RELEASE SAMPLING ANALYSIS ACTIVITY TYPE FREQUENCY FREQUENCY ANALYSIS (d)
Waste Tank P
P Principal Batch Gamma Releases (a)
Emitters (c)
I-131 One Batch /M M
Dissolved and Entrained Gases (gamma emitters)
P M
H-3 Composite (b)
Gross Alpha P
Q g
Composite (b)
Sr-89, Sr-90 i
Fe-55 l'
a.
A batch release is the discharge of liquid wastes of a discrete volume.
Prior to sampling for analyses, each batch shall be isolated, and then thoroughly mixed, to assure representative sampling.
b.
A composite sample is one made up of individual samples which are proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquid released.
c.
The principal gamma emitters for which the LLD specification will apply are exclusively the following radionuclides:
Mn-54, Fe-59, Co-58, Co-60, Zn-62, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.
d.
Methods of calculating the Lower Limits of-Detection (LLD) shall be contained in plant procedures.
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- 4/5.3.2 RADIOACTIVE GASEOUS EFFLUENT
. INSTRUMENTATION LI_MITING CONDITI_0N_F_01_0_PEJRTION _
4.3.2.1 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 4.3.2.1 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 4.3.2.2 are not exceeded. The noble gas instrumentation alarm setpoint will be determined and adjusted in accordance with the methodology and parameters in the ODCM.
' APPLICABILITY: As shown in Table 4.3.2.1.
ACTION:
a.
With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than that required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable, or change the setpoint so it is acceptably conservative.
b.
With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION required by Table 4.3.2.1. Exert best efforts to return the instruments to OPERABLE status within 30 days, and if unsuccessful, explain in the next Semi-Annual Radioactive Effluent Release Report _ pursuant to 6.9.3.a why the inoperability was not corrected in a timely manner.
c.
The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1 are not applicable.
SURVEILLANC_EAQUIfEjiEXT_S_
5.3.2.1 Each radioactive gaseous effluent monitoring instrumentation
. channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION operations at the frequencies shown in Table 5.3.2.1.
2.
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TABLE 4.3.2.1 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION A MINIMUM CHANNELS APPLICABLE INSTRUMENT OPERABLE CONDITIONS ACTION 1.
Reactor Containment Building Ventilation Monitor System a.
Particulate Activity 1
B Monitor b.
Gaseous Activity-1 B
Monitor c.
Sampler Flow Rate 1
C
- Measuring Device 2.
Stack Monitor System a.
Noble Gas' Activity Monitor 1
D b.
Iodine Sampler 1
E
- c. _ Particulate-Sampler 1
E d.-
Sampler Flow Rate-1 C
Measuring Device
- When Containment Building Ventilation System is in operation.
- At all times, unless alternate monitoring is available.
A.
For post-accident instrumentation, refer to Section 4.5.1.
B.
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases through this pathway may continue for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, as long as stack monitors are OPERABLE.
C.
With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
D.
With the number'of channels OPERABLE less than required by the Minimum Channels i
OPERABLE requirement, effluent releases via this pathway may continue provided alternate monitoring is available or grab samples are taken at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and these samples are analyzed for gross activity within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
E. - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided alternate monitoring is available meeting the requirements of Table 5.3.2.1 or samples are continuously collected with auxiliary sampling equipment.
LACBWR-WP-4.12 43
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TABLE 5.3.2.l' RADIOACTIVE CASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL SURVEILLANCE CHANNEL SOURCE FUNCTIONAL CHANNEL REQUIREMENT INSTRUMENT CHECK CHECK TEST CALIBRATION (4) CONDITIONS 1.
Reactor Containment Building Ventilation Monitor System a.
Particulate Activity Monitor D
M Q(1)
R Q(3) 1)
R b.
Gaseous Activity Monitor D
M Q(
R c.
Sampler Flow Rate Measuring Device D
N/A 2.
Stack Monitor System a.
Noble Gas Activity Monitor D
M Q(2)
R b.
Iodine Sampler D
M Q(2)
R c.
Particulate Sampler D
N/A Q(2)
R d.
Sampler Flow Rate Measuring Device D
N/A Q(3)
R During applicable conditions per Table 4.3.2.1.
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if any of the following conditions exist:
a.
Instrument indicates measured levels at or above the alarm setpoint.
b.
Instrument indicates a downscale failure (provides control room annunciation alarm only).
c.
Instrument indicates a circuit failure (provides control room annunication alarm only).
(2) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if'any of the following conditions exist:
a.
Instrument indicates measured level above the alarm setpoint on one channel.
b.
Instrument indicates a failure by a Low Flow and Low Count Rate signal.
c.
Instrument controls in Maintenance mode.
d.
Instrument indicates a parameter circuit failure.
(3) The CHANNEL FUNCTIONAL TEST shall also demonstrate that the control room local alarm occurs if the flow instrument indicates measured levels below the minimum and/or above the maximum alarm setpoint.
(4) The CRANNEL CALIBRATION shall be conducted in accordance with plant procedures.
LACBWR 44 UP-4.12
RADIOACTIVE GASEOUS EFFLUENTS INSTANTANEOUS DOSE RATE LI MI.T_I NG_C_0JiD_III 0]L_FO R,0_P E_RA_T_IO N 4.3.2.2 The dose rate due to radioactive materials released in gaseous effluents to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited to the following:
a.
The dose rate limit for noble gases shall be < 500 mrems/yr to the total body and j[ 3000 mrems/ year to the skin, and b.
The dose rate limit for 1-131, 1-133,-H-3 and for all radioactive materials in particulate form with half lives greater than 8 days, shall be j[ 1500 mrems/ year to any organ.
APPLICABILITY: At all times.
ACTION:
a.
With the dose rate (s) exceeding the above limits, without delay decrease the release rate to within the above limit (s).
b.
Specifications 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE RE_QUIREMENTS
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5.3.2.2.1.The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methods and
-procedures of the ODCM.
5.3.2.2.2 The dose rate due to I-131, I-133, H-3 and for all radioactive materials in particulate form with half lives > 8 days in gaseous effluents i
shall be determined to be within the above limits in accordance with the method and procedures of the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 5.3.2.2.
t t
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LACBWR WP-4.12 45
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d TABLE 5.3.2.2 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM
.r Gaseous
- Sampling, sMinimum' Analysis Release Type Frequency Frequ'ency Type of Activity Analysis (d)(e)
A. Stack Continuous (b) g(a)
I-131 Effluents Charcoal Sample I-133 Continuous \\D>
W(al Composite Particulate' Sample Principal Gamma Emitters (c)
Continuous (DJ-Q CoQposite e
Particulate Sample S r-89,90 Continuous (D)
M Composite i
Particulate Sample Gross Alpha Continuous (D)
Noble Gas Noble Gases Monitor Gross Beta and Gamma
~
M M
Principal Gamma Emitters (c)
-Marinelli or Bomb H-3481 TABLE NOTATION:
(a) Filter and cartridge samples shall be changed at least weekly and analyses shall be completed.within 7 days after removal from sampler.
- i l
(b) The ratio of the sample flow rate to the sampled stream. flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications' 5.3.2.2.1, 5.3.2.2.2, 5.3.2.3-and 5.3.2.4.
i d
(c) The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:
Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, co-58, Co-60, Zn-65, Tc-995, Cs-134, Cs-137,-Ce-141 and Ce-144 for particulate caissions. This list does not mean that only these
+
nuclides are to be considered. Other gamma peaks that are identifiable and measureable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Spec. 6.9.3.a.
(d) When upper cavity is flooded, stack tritium grab samples will be taken at least once per'7 days.
(e) Lower Limits of Detection (LLD) are determined in.accordance with plant procedures.
LACBWR I
WP-4.12 46
ADI0 ACTIVE GASEOUS EFFLUENTS DOSE, NOBLE GASES-LIMITING CONDITION FOR OPERATI_0N 4.3.2.3 The air dose to a MEMBER OF THE PUBLIC due to noble gases released in gaseous effluents to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited to the following, (See Figure 4/5.3):
a.
During any calendar quarter, to < 5 mrads for gamma radiation and j[ 10 mrads for beta particle radiation; and b.
During any calendar year, to j[ 10 mrads for gamma radiation and j[ 20 mrads for beta particle radiation.
APPLICABILITY: At all times.
ACTION:
a.
With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit.to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit (s) and defines the corrective actions to be taken to reduce the releases of radioactive noble gases in gaseous effluents so that the cumulative dose during the calendar year is within (10) mrad for gamma radiation and (20) mrad for beta radiation.
b.
The provisions of Specification 3.0.3, 3.0.4 and 6.9.1 are not-applicable.
SURVEILLANCE REAUI RM_ENTS 5.3.2.3 Dose Calculations: Cumulative dose contributions for the current calendar quarter and current calendar year shall be determined in accordance with.the Offsite Dose Calculation Manual (ODCM) at least once every 31 days.
I f
i 1
-LACBWR 1WP-4.12 47
RADIOACTIVE GASEOUS EFFLUENTS DOSE, RADIONUCLIDES OTHER TRAN NOBLE GASES LIMITING CONDITION FOR OPERATION 4.3.2.4 The dose to a MEMBER OF THE PUBLIC from I-131, I-133, H-3, and all radionuclides in particulate form with half-lives greater than 8 days, in gaseous effluents released to areas beyond EFFLUENT RELEASE BOUNDARY shall be limited to the following (See Figure 4/5.J):
a.
During any calendar quarter to < 7.5 mrems to any organ, and i
b.
During any calendar year to < 15 mrems to any organ
. APPLICABILITY: At all times.
ACTION:
a.
With the calculated dose from the release of I-131, I-133, H-3 and all radionuclides in particulate form with half lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit and defines the corrective actions to be taken to reduce these releases in gaseous effluents so that the cumulative dose during the the calendar year is within 15 mrem to any organ.
b.
The provisions of Specifications.3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE R QUIREMEy S 5.3.2.4 Dose Calculations: Cumulative dose contributions for the current calendar quarter and current calendar year _shall be determined in accordance with the ODCM at least once every 31' days.
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l LACBWR I
WP-4.12 48
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A RADIOACTIVE GASEOUS EFFLUENTS GASEOUS RADWASTE SYSTEM LIMITING CONDITION FOR OPERATION 4.3.2.5.A GASEOUS RADWASTE SYSTEM shall be in operation.
APPLICABILITY: Whenever the main condenser air ejector system is in-operation.
ACTION:
i l
With the GASEOUS FADWASTE SYSTEM. inoperable for more than 7 days" a.
prepare and submit to the Commission within 30 days, pursuant to 4
Specification 6.9.2, a Special Report which includes the following
}
information:
.1.
Identification of the inoperable equipment or subsystems and the reason for nonoperability.
l 2.
Action (s) taken to restore the inoperable equipment to OPERABLE status.
3.
Summary description of action (s) taken to prevent a recurrence.
b.
The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE REQUI_ REM _EN_T_S R
5.3.2.5.
Cumulative doses due to gaseous releases to areas beyond EFFLUENT RELEASE BOUNDARY will be calculated at least once per 31 days in accordance with methodology and parameters in the ODCM.
i
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n LACBWR WP-4.12 49
RADIOACTIVE GASEOUS EFFLUENTS VENTILATION EXHAUST TREATMENT SYSTEM (CONTAINMENT BUILDING)
LIMITING CONDITION FOR OPERATION 4.3.2.6 The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM 1
'^
shall be-in-operation and be used to reduce particulate radioactive materials t
in gaseous waste prior _to their discharge.
~ APPLICABILITY: Whenever the CONTAINMENT BUILDING VENTILATION EXHAUST TREATMENT SYSTEM is in operation.
ACTION: -With the Containment Building Ventilation Exhaust being discharged without treatment, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:
l
-1.
Identification of the inoperable equipment or subsystems and the reason for nonoperability.
2.
Action (s) taken to restore the inoperable equipment to OPERABLE status.
l 3.
Summary description of action (s) taken to prevent a recurrence.
The provisions of Specification 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE RE_QU_IREMENTS i'
5.3.2.6 Cumulative doses due to gaseous releases to areas beyond EFFLUENT RELEASE BOUNDARY will be calculated at least once per 31 days in accordance with methodology and parameters in the ODCM.
f RADIOACTIVE EFFLUENTS SOLID RADI0 ACTIVE WASTE I
LIMITING CONDITION FOR OPERATION i'.
4.3.3 Solid radioactive wastes shall be handled in accordance with a PROCESS CONTROL PROGRAM in order to meet' shipping and burial ground requirements.
APPLICABILITY: At all times when processing solid radioactive wastes for shipment and disposal.
ACTION: With the provisions of the PROCESS CONTROL PROGRAM not satisfied, suspend shipments of defectively processed or defectively packaged solid radioactive wastes from the site.
The provisions of Specifications 3.0.3, 3.0.4, and 6.9.1 are not applicable.
SURVEILLA_N_C_E RE_QUIREMENTS l
5.3.3 The PROCESS CONTROL PROGRAM shall be used to assure the appropriate form for packaging each type of radioactive waste (e.g., filter sludges, spent resins, tank bottoms, dry active wastes.)
I LACBWR WP-4.12 50 i
- RADIOACTIVE EFFLUENTS TOTAL DOSE i
LIMITINC CONDITION FOR OPERATION 4.3.4 The dose equivalent'to any MEMBER OF THE PUBLIC due to. releases of radioactivity and radiation, shall be limited to < 25 mrems to the total body 1
or any organ (except_the thyroid, which'is limited to < 75 mrems) over a period of one calendar. year.
APPLICABILITY: At all times.
ACTION:
a.
With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limit of Specifications 4.3.1.3.b, 4.3.2.3.b or 4.3.2.4.b, a determination should be made, including direct radiation from reactor containment and radioactive waste storage tanks to determine if: limits of Specification 4.3.4 have been exceeded.
If the limits of Specifi-cation 4.3.4 have been exceeded, prepare and submit a Special Report (including an analysis which estimates the radiation exposure
~to a MEMBER OF THE PUBLIC for the calendar year,) to the Director, Nuclear Reactor Regulation, U. S. Nuclear Regulatory Commission, Washington, D. C.
20555, within 30 days, which defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding.the limits of Specification 4.3.4. If the release condition resulting in violation of Specification 4.3.4 i-has not already been corrected, the Special Report shall include a request ' for a variance in accordance with the provisions of 40 CFR 190.-
Submittal of the Special Report is considered a timely request, j
and a variance is granted until staff action on the request is complete.
4 b.
The provisions of Specification 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE RE_QQIREMENTS 5.3.4.1 Dose Calculations: Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Specifications 5.3.1.3, 5.3.2.3, 5.3.2.4, and in accordance with the Offsite Dose f
Calculation Manual (ODCM) once per year.
5.3.4.2 Dose Determination: Cumulative dose contributions from direct radiation from the reactor containment or radioactive waste storage tanks shall be determined in accordance with the methodology and parameters of the ODCM once per year.
i LACBWR
.WP-4.12 51
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- 4/5.4 RADIOLOGICAL ENVIRONMENTAL MONITORING 4
I' PROGRAM, LAND USE CENSUS AND INTERLABORATORY COMPARISON LIMITING COfLDITI_QN FOR M RA_T_I,0_N
~
4.4.1 The radiological environmental monitoring program, land use census and interlaboratory comparison shall be conducted as specified in the plant procedures.
The radiological'er.vironmental monitoring program shall include exposure pathway sampling, frequencies of sampling and analyses, reporting levels and LLD's.
APPLICABILITY: At all times.
ACTION:
With the radiological environmental monitoring program not being a.
conducted as specified in plant procedures, or if sample analyses exhibit unexpected results, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.3.b, a description of the reasons for not conducting the program as required, analysis of the causes of unexpected results, and the plans for preventing a recurrence.
b.
The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1 are not applicable.
i RADIOLOGICAL ENVIRONMENTAL MONITORING SURVEILLANCE REQUIREMENTS 5.4.1 The radiological environmental monitoring samples shall be collected in accordance with plant procedures from locations specified in the procedures, and shall be analyzed in accordance with requirements listed in plant procedures.
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i 4/5.3' RADI0 ACTIVE EFFLUENTS BASES 4/5.3.1 RADIOACTIVE LIQUID EFFLUENTS 4/5.3.1.1 INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor the releases of radioactive materials in liquid effluents during actual or potential releases of liquid effluents with the alarm setpoints set
_to ensure that the alarm will occur prior to exceeding the limits of 10 CFR Part 20.
4/5.3.1.2 CONCENTRATION This specification is-provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site will be less than the concentration levels specified in 10 CFR Part 20, Appendix B, Table II,.
Column 2. -The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection l
(ICRP) Publication 2.
4/5.3.1.3 DOSE This specification is provided to implement the requirements of Sections II.A, III.A, IV.A and Annex of Appendix I, 10 CFR Part 50.
The dose calculations in the ODCM implement the requirement in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures
-based on models and data, such that the actual exposure of an individual 4
through appropriate pathways is unlikely to be substantially underestimated.
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LACBWR WP-4.12 53 l
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RADIOACTIVE EFFLUENTS BASES 4/5.3.2 RADIOACTIVE GASEOUS EFFLUENTS 1
.4/5.3.2.1 INSTRUMENTATION The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases'of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm setpoints for these instruments shall be set to ensure that the alarm will
-occur prior to exceeding the limits of 10 CFR Part 20.
4/5.3.2.2' INSTANTANEOUS DOSE RATE This specification is-provided to ensure that the dose rate at any time at the EFFLUENT RELEASE BOUNDARY from gaseous effluents from LACBWR will be within the annual dose limits of 10 CFR Part 20 for unrestricted areas. The annual dose limits are the doses associated with the concentrations of 10 CFR Part 20, Appendix B, Table II, Column 1.
These limits provide reasonable assurance that radioactive material discharged in gaseous effluents will not result in the exposure of an individual in an unrestricted area, outside the EFFLUENT RELEASE BOUNDARY to annual average concentrations exceeding the limits J
.specified in Appendix B, Table II of 10 CFR Part 20 (10 CFR Part 20.106(b)).
For individuals who may at times be within the EFFLUENT RELEASE BOUNDARY, the occupancy of the individual will be sufficiently low to compensate for any increase in the atmospheric diffusion factor above that for the EFFLUENT RELEASE BOUNDARY. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to an individual at or beyond the EFFLUENT RELEASE BOUNDARY to f 500 mrem / year to the total body or to f 3000 mrem / year to the skin. These release rate limits also i
restrict, at all times, the corresponding thyroid dose rate above background to an infant via the cow-milk-infant pathway to f 1500 mrem / year for the i
nearest cow to the plant.
4/5.3.2.3 DOSE, NOBLE GASES l
This specification is provided to implement the requirements of Sections II.B.
III. A, and IV. A' of Appendix I, 10 CFR Part. 50.
The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I that l.
l conformance with the guides of Appendix I is to be shown by calculational procedures. based on models and data such that the actual exposure of an individual through the appropriate pathways is unlikely to be substantially underestimated.
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LACBWR WP-4.12 54 l
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RADIOACTIVE EFFLUENTS BASES L
4/5.3.2'.4 DOSE, RADIONUCLIDES OTHER THAN NOBLE GASES i
This specification is provided to implement the requirements of Sections II.C,
_III.A, IV.A and Annex of Appendix I, 10 CFR Part 50.. The ODCM calculational methods specified in the surveillance requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by' calculational procedures' based on models and data such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.
4/5.3.2.5 GASEOUS RADWASTE SYSTEM The OPERABILITY of the' gaseous radwaste system ensures that the system will be available for use whenever gaseous effluents require treatment prior to release to the environment. This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A.to 10 CFR Part 50.
4 4/5.3.2.6 VENTILATION EXHAUST TREATMENT SYSTEM (CONTAINMENT BUILDING)
The OPERABILITY of the ventilation exhaust treatment system (Containment Building) ensures that the systems will be available for use whenever gaseous 4
f' effluents require treatment prior to release to the environment. This
~
specification implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50.
4 i
j 4/5.3.3 SOLID RADIOACTIVE WASTE The OPERABILITY of the solid radwaste system ensures that the system will be i
l available.for use whenever solid radwastes require processing and packaging l
prior to being shipped offsite. This specification implements the require-F ments of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to
4/5.3.4 TOTAL DOSE This specification is provided to meet the dose limitations of 40 CFR 190.
The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant radioactive effluents exceed twice the design objective doses of Appendix I.
The Special Report will describe a 1
course of action which should result in the limitation of dose to a real individual for 12 consecutive months to within' the 40 CFR 190 limits.
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LACBWR WP-4.12 55
Y 4/5.4 RADIOLOGICAL ENVIRONMENTAL MONITORING
_B_A_S_ES A
4/5.4.1 -MONITORING PROGRAM, LAND USE CENSUS, AND INTERLABORATORY COMPARISON PROGRAM The radiological monitoring program required by this specification provides measurements'of radiation and of radioactive materials in those exposure pathways and for those radionuclides, which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program theory supplements'the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental exposure pathways.
'This specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census.
The requirement for participation in an Interlaboratory Comparison Program is provided.to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental samples are performed to demonstrate that the results are reasonably valid.
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LACBWR WP-4.12 56
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4/5.5 POST-ACCIDENT RADIATION MONITORING INSTRUMENTATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 4.5.1 The post-accident radiation monitoring instrumentation channels shown in Table 4.5.1 shall be OPERABLE with their alarm setpoints within the specified limits.
APPLICABILITY: As shown in Table 4.5.1.
ACTION:
With a post-accident radiation monitoring channel alarm setpoint a.
exceeding the value shown in Table 4.5.1, adjust the setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or-declare the channel inoperable.
b.
With less than the required number of post-accident radiation monitoring channels OPERABLE, take the ACTION shown in Table 4.5.1.
c.
The provisions of Specifications 3.0.3, 3.0.4 and 6.9.1 are not applicable.
SURVEILLANCE REQUIREMENTS 5.5.1 Each post-accident radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK daily, CHANNEL FUNCTIONAL TEST monthly, and CHANNEL CALIBRATION at least once per 18 months during applicable conditions shown in Table 4.5.1.
BASES FOR POST-ACCIDENT RADIA_TI_0_N MONITORING INSTRUMENTATION 4/5.5.1 POST-ACCIDENT-RADIATION MONITORING INSTRUMENTATION The operability of the post-accident radiation monitoring is required following accidents to indicate core damage as prescribed in NUREG 0578, i
Section 2.1.8.b and NUREG 0737,Section II.F.1-3 and post-accident indication of noble gas effluent as prescribed in NUREG 0578, Section 2.1.8.b and NUREG 0737,Section II.F.1-1.
The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Calibrations of Containment Building Area High Range Radiation Monitors may be performed under the requirements of NUREG 0578.
LACBWR i
WP-4.12 58
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TABLE 4.5.1 POST-ACCIDENT RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE INSTRUMENT OPERABLE CONDITIONS ALARM SETPOINT ACTION 1.
Containment Building High Range Area Radiation Hi Alarm 10 _104 3
Monitors 1
1,2,3 Rad /hr (1)-
2.
Stack High Range Noble Gas Effluent Monitor 1
1,2,3 Hi Alarm <1.25 ac (1) 102 pCi/cc 3.
Stack Radioiodine Effluent Monitor 1
1,2,3 Hi Alarm < 10 pCi (1)
(1) With the number of OPERABLE Channels less than minimum required, either restore the inoperable Channel (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:
a) Initiate an alternate method of monitoring containment building high range radiation levels $ or stack releases using alternate channel of stack monitoring system (range 0
10 -10 pCi/cc for noble gas) or sample every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, as applicable.
- and -
b) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
1 LACBWR WP-4.12 59
5.
MAINTENANCE 5.1 GENERAL
- 5.1.1 Maintenance operations, including refueling operations and routine tests, shall be performed in conformance with these specifications.
5.1.2 Maintenance operations on the reactor coolant system, reactor components, or emergency systems shall be performed as authorized by the Shift Supervisor whenever fuel is in the reactor. -Maintenance involving the opening of systems containing radioactive materials shall be conducted under the surveillance of a Health Physics representative.
5.1.3 Maintenance operations which involve the opening of the reactor primary coolant system shall be performed with the reactor-shut down and depressurized to atmospheric. pressure. The reactor primary coolant auxiliary systems may be serviced during reactor operation if the system is first isolated, depressurized to atmospheric pressure, and cooled to 150*F.
The limitations of this paragraph do not apply to che supervised collection of steam or water samples, and super -' sed venting of process instruments.
5.1.4
- Deleted -
+
1 LACBWR 5-1 l
WP4.12
5.1.5-Components which have been repaired, replaced, or otherwise subjected to temporary or permanent modification shall be tested in accordance with procedures which are appropriate in view of the nature of the repair, replacement or modification, and in view of the condition of the system.
5.1.6
- Deleted -
5.1.7 D'efinitions applicable'to Sec. 5.2.15 are as follows:
Channel Check: A qualitative determination of acceptable operability by observation of channel behavior during operation. This determination shall include, where possible, comparison of the-channel with other-independent channels measuring the same variable.
Channel Test:
Injection of a' simulated signal into the channel to verify its proper response including, where applicable, alarm and/or trip initiating action.
Channel Calibration: Adjustment of channel output such that it responds, with acceptable range and accuracy, to known_vnlues of the parameter which the channel measures. Calibration shall encompass the entire channel, including equipment actuation, alarm, or trip.
5.2 TESTING 5.2.1 Containment Testing 5.2.1.1 Containment Integrated Leakage Rate Test (Type A Tests):
(a) The integrated leakage rate test shall be performed at a pressure of at least 52 psig without any preliminary leak detection surveys and-repairs except as necessary-to correct any evidence of structural deterioration which
.may affect either the containment's structural integrity or leak tightness.
Such structural deterioration and corrective actions taken shall be reported as part of the-Type A test report.
Closure of containment isolation valves shall be accomplished by normal mode of actuation and without any preliminary exercise or adjustments. If valve closure malfunction is detected which requires corrective action before the test,.this information shall be included in the report submitted to the Commission as required under Section 5.2.1.5.
.The test duration shall be for a sufficient period of time to obtain meaningful leakage rate results. In addition, a controlled leakage rate test shall be included to verify the_ test accuracy.
i i
i LACBWR 5-2 WP4.12
(b) Acceptance Criteria:
Tne maximum allowable test leakage rate Lpm shall not exceed 0.1 percent per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the test pressure of 52 psig.
If local leakage measurements are taken to effect repairs in order to meet the acceptance criterion, these measurements shall be taken at a test pressure of 52 psig.
To provide a margin for possible deterioration of the containment leakage integrity during the service interval extending to the subsequently scheduled Type A test, the measured leakage rate Lpm shall be reduced, if necessary, to a value Lpo not in excess of 75% of the leakage rate limit (0.075% per 24 hrs.
at the test pressure of 52 psig). This leakage reduction shall be accomplished prior to resumption of plant operation.
(c) Corrective Actions: Where excessive leakage is expertenet'd during a Type A test, leaks may be found and isolated from the test. Penntrations so isolated must be capable of local leakage testing. Once these leaks have been isolated, the Type A test may continue. Following the Type A test, local leakage rates must be measured before and after repairs to each isolated leakage path. The results of the Type A test are then back-corrected utilizing the conservative assumption that all measured local leakage is in a direction out of the containment unless it can be demonstrated otherwise. The local leakage measurements before the repair are added to the Type A results to determine the "as is" condition (except that leakage which is determined to be into the containment), while the af ter-repair measurements determine the "as left" condition. For a satisfactory Type A test, the sum of the Type A test leakage and local leakage measurements must be less than the maximum allowable test leakage rate.
(d) Test Frequency:
(1) A set of three Type A tests shall be performed, at approximately equal intervals during each 10 year service period, with the third test of each set coinciding with the end of each 10 year service period. Type A test periods may coincide with the plant in-service inspection shutdown periods.
(2) If any Type A test fails to meet the acceptance criteria of (b) above, the test schedule applicable to subsequent Type A tests shall be subject to review and approval by the Commission.
(3) If two consecutive periodic Type A tests fail to meet the acceptance criterion of (b) above, notwithstanding the periodic retest schedule, a Type A test shall be performed at intervals not greater than 18 months until two consecutive Type A tests meet the acceptance criteria, at which time, the retest schedule specified in (d.1) above, may be resumed.
LACBWR 5-3 WP4.12
4-3
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s 5.2.1.2 Individual Leak-Detection Tests (Type B and C Tests):
Type B and C tests shall be performed as follows:
(a) Type B Tests: Leak-detection tests shall be performed at a
.. pressure of at least 52 psig-by using-the soap-bubble technique (or other -
' methods of equivalent sensitivity)-or by determining the rate of pressure loss of pneumatically pressurized test chambers on the following containment
- components:
the_ electrical-penetrations, the reactor building spray valve shaft penetration, theifreight door,'and the containment building airlocks i
~
Containment _ components other than mentioned above, which develop leaks
- requiring repairs during the performance of Type A test, shall be included in a subsequent Type B test.
j.
Component Leak Surveillance System:
A. leak surveillance system (i.e.,
continuous pressurization of individual centainment components) that maintains a pressure not less than 52 psig at individual test chambers of containment penetrations and seals during normal reactor operation.shall be acceptable in lieu of Type B tests of the components under such leak surveillance.
(b) Type C Tests: Containment Isolation Valve leak detection testing l
shall be conducted at a pressure of 52_psig.
_ (c) ' Acceptance. Criteria: The combined leakage rate for all penetrations j
.and valves subject to Type B and C tests shall not exceed 60% of the maximum j-
_ allowable Type A test leakage rate.
r (d) Corrective Actions: Leaks which cause the acceptance criteria of (c) to be. exceeded shall be repaired and retested until the criteria'is met.
l
. Repairs,of lesser leaks are optional.
-(e)~ Test Frequency: Type B and C tests (except for air locks and i-electrical penetrations) shall be performed at intervals no greater than 2-years. Air locks shall be-tested at 4-month intervals. The freight door shall be tested following each closure prior to plant startup. Electrical penetrations shall be tested at intervals no greater than one year.
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I LACBWR 5-4 WP4.12 I
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5.2.1.3 ' Permissible Periods for Testing: The performance of Type A tests l
shall be limited to periods when the plant facility is nonoperational and secured in the shutdown condition under administrative control and safety procedures.
5.2.1.4 Report of Test Results: The leakage rate results of Type A, B, and C l
tests that meet the acceptance criteria shall be reported in the applicable LACBWR operating report. Leakage test results of Type A, B, and C tests that fail to meet the acceptance criteria shall be reported in a separate summary -
that includes an analysis and interpretation of the test data, the least-squares fit analysis of the test data, the instrumentation error analysis, and the structural conditions of the containment or components, if any, which contributed to the failure in meeting the acceptance criteria. Results and analyses of the supplemental verification test employed to demonstrate the validity of the leakge rate test measurements also shall be included.
5.2.1.5 Containment Ventilation Isolation Valve Leakage Tests:
(a) Tests: The Containment ventilation system dampers shall be subjected to leakage tests, in addition to the tests-required by Section 5.2.1.2.
The tests shall be conducted at an intial pressure of at least 52 psig.
(b) Acceptance Criteria: Excessive degradation is determined not to exist and the isolation valve (s) is considered operable if pressure decreases by no greater than 10 psi in a ten-minute test period.
(c) _ Corrective Action: If excessive degradation exists, the leakage path must be repaired or. isolated,'and retested until the criteria is met.
A passing leakage test must be achieved within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or the reactor must be placed in hot shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless a passing
+
test is achieved.
'(d) Test Frequency:
The leakage tests of the containment ventilation l
isolation dampers shall be conducted at least quarterly.
i 5.2.2 The reactor building isolation system will be tested for proper operation. prior to every cold startup, but this test will not be required more often than at 30-day intervals.
i
[
5.2.3 The exterior surfaces of the LACBWR ventilation stack and the smoke i_
stack of the conventional steam power generating station, Genoa 3, adjacent to the LACBWR plant shall be inspected for structural integrity at an interval-no l
longer than'S years following the initial construction inspection, and at subsequent intervals not longer than 5 years apart.
i.
l' 1
l l
.LACBWR 5-5 WP4.12 i
5.2.4 The reactor vessel shall be hydrostatically tested at 1400 psig after
,any of its gasketed joints have been opened and resealed. All hydrostatic
" tests shall be performed with the vessel at a temperature no lower than that specified in Section 4.2.2.4.
5.2.5 The forced circulation system controls and automatically-operated valves shall be tested for proper operation at each refueling shutdown with test intervals not to exceed 18 months.
5.2.6 The shutdown condenser system control valves shall be tested at least quarterly to demonstrate their operability. The integrated system shall be tested for proper operation at each refueling shutdown with test intervals not to exceed 18 months. In addition, the condenser tube bundle shall be pressurized to greater than 1250 psig and tested for leakage at each refueling shutdown.
(Next number is 5.2.9) l 5.2.9 The boron-injection system controls and the remotely-operated valves shall be tested for proper operation during cold shutdowns, but not more often than every 92 days.
(Next number is 5.2.12) 5.2.12 control rod scram time tests shall be performed and the total scram insertion time shall be demonstrated to be within the limit specified in Section 4.2.5.1:
(a) for all rods prior to each cold startup after the reactor vessel head has been removed, but at least annually unless 50% of the rods had been tested during the previous 365 days as describe in (c)
- below, (b) for specifically affected individual control rods following main-tenance on or modifications to the control rod or control rod drive mechanism which could affect the scram insertion time of those specific control rods, and (c) for 10% of the individual control rods (3 rods) on a rotating basis prior to resuming power operation following each reactor shutdown, hot or cold, but not required more often than every 30 days.
5.2.13 Each control rod drive mechanism shall be exercised by moving each partially or fully withdrawn control rod at least one-half inch in any one direction at least once per 31 days.
5.2.14 Proper operation of both control solenoids for each of the hydraulic scram valves will be determined in conjunction with each scram time test required by Section 5.2.12.a.
5.2.15 Instrumentation shall be checked, tested and calibrated as indicated in the following chart.
I LACBWR 5-6 WP4.12
(
MINIMUM FREQUENCIFS
'FOR TESTING, CALIBRATING, AND/OR CHEKING OF INSTRUMENTATION 1:
Channels Action Minimum Frequency-1.-
Reactor Water Level Calibration At each refueling shutdown.
Test *
. Monthly when in service and prior to each reactor startup if test has not been performed within 30 days.
Check Daily.
2.
Reactor Pressure Calibration At each refueling shutdown.
Test
- Monthly when in service i
and prior to each reactor startup if test has not been performed within 30 days.
' Check Daily.
3.
Reactor Power - Flow Calibration At each refueling shutdown.
}
Test
- Monthly when in service and prior to each reactor startup if test has not been performed within 30 days.
Check Daily.
4.
Reactor Coolant Flow Calibration At each refueling shutdown.
Rate Low Test * ~
Monthly when in service and prior to each reactor startup if test has not been performed within 30 days.
Check Daily.
5.
Source Range Test Prior to each reactor i.
(Channels 1 and 2)
(60 cycles startup if test has not been per sec.)
performed within 30 days.
Check once per shift when in service.
i i
LACBWR 5-7 WP4.12-1
. _ _.. - _. ~.. _.,
MINIMUM FREQUENCIES FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION - (cont'd)
Channels Action Minimum Frequency 6.
Intermediate Range Test (10-10 Prior to each reactor start (Channels 3 and 4) and 10-5 amps; up if test has not been per-period *)
formed within 30 days.
Check Once per shift when in service.~
7.
Wide Range and Power Check by heat Monthly when in service.
Range (Channels 5, balance 6, 7, and 8)
A.
Nuclear Instru-Test
- Monthly when in service mentation & Auto-and prior to each reactor matic Gain Control startup if test has not been Sub-System.
performed within 30 days.
B.
Nuclear Instru-Check Once per shift when in mentation & Auto-service.
matic Gain Control Sub-System.
C.
Automatic Gain Control Sub-System Calibration At each refueling shutdown.
NOTE: Testing of the Nuclear Instruments and Automatic Gain Control Sub-System shall be done concurrently.
(
- 8. ' Full Scram Circuits Test for hot Once a month, short by means of built-in test switch 9.
Area Radiation Calibration At each refueling shutdown.
Monitors Test Every two weeks.
Check Daily.
i l
I LACBWR 5-8 WP4.12 P
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n MINIMUM FREQUENCIES
-FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION - (cont'd) i Channels Action Minimum Frequency 10., 11., 12.
- deleted -
13.
Portable Radiation Calibration Semi-annually.
Detectors Check Every two weeks.
i 14.
Main Condenser Vacuum Calibration At each refueling shutdown.
j; Test
- Prior to each plant.startup j
if test has not been per-formed within the last 30 days.
4 15.
Reactor Building Calibration J At each refueling shutdown.
Pressure.
- 16.. Low Main Steam Calibration At each refueling shutdown.
-Pressure 17.
Reactor Building Main Test
- Prior to each plant startup Steam Isolation Valve if test has not been per-formed within 30 days.
18.
Turbine Building Main Test
- Prior to each plant startup Steam Isolation Valve if test has not been per-formed within 30 days.
I 19.
Reactor Building MCC Test
- At each refueling shutdown.
1A Under-Voltage Relays 20.
2400 v Busses 1A and' Test
- At each refueling shutdown.
1B Under-Voltage Relays LACBWR 5-9 s,,
WP4.12-
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i MINIMUM FREQUENCIES FOR TESTING, CALIBRATING, AND/OR CHECKING OF INSTRUMENTATION - (cont'd)
Channels Action Minimum Frequency 21.
CRD Accumul.$ tors Low Test Prior to each plant startup 011 Level.Jeram Relay if test-has not been per-formed within 30 days.
t 22.
CRD iccumulators Low Tesk.
Prior to each plant startup Gas Pressurei cram
- if test has not been per-S Relay
> j formed within 30 days.
Check pressure Weekly.
I '.
indication 23.
Turbine Stop Valve Test Prior to each plant startup
-if test has not been per-formed within 30 days.
l 24.
Reactor Pressure Calibration At each refueling shutdown.
(RPTS) 25.
Reactor Water Level Calibration At each refueling shutdown.
4 (RPTS) s 26.
Reactor Safety Valve Check Monthly.
Position Indication Calibration At each refueling shutdown. l
_f i
- Test shall include trippind of-the scram relays K-ll4.
i 5.2.16 Corrosion test coupons.shall be inserted in the forced circulation i
loop to evaluate the corrosian deteriora'cion of chromium-molybdenum piping; and that piping shall be replaced if the. reduction in pipe wall thickness, as indicated by weight loss and metallographic evaluation of the test coupons, is t
-greater than 0.190 inches. The replacement piping shall be stainless steel or shall be clad'iaternally with stainless steel and shall meet the design requirements of Section 2.3.2.
5-
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5 '
I LACBWR WP4.12 L'
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4'.
O a
i TABLE 1 OPERATING LIMITS KEYSWITCH BYPASS CONDITION CHANNEL OR' SENSOR SET POINT ACTION PROVISION'
- 1. Reactor Power Two of four nuclear Table 4.0.2.2.1-1 Full scram None High channels 5, 6, 7 &
8 if power level-is m
4 l
>5% of full power Either nuclear Table 4.0.2.2.1-1 Fall scram One channel may be by-passed for calibration channel 5 or 6 ifJ and testing.
~
power level is <5%
of full power.
4
- 2. Reactor Period Nuclear channel
' Table 4.0.2.2.1-1 Full scram
- 1) Both channels may be Short 3 or 4 bypassed only when reactor power exceeds 3 Mwt.
~
- 2) One channel may be bypassed for cali-bration and testing.,
- 3. Reactor Pressure safety j:,1325 psig
- 1) Full scram One channel may be by-Pressure channel 1 or 2
- 2) Shutdown condenser passed for calibration High operates and testing.
- 3) Closure of venti-lation inlet and outlet dampers
- 4) Closure of con-tainment'off gas vent header valve.
LACBWR 5-11 WP4.12
TABLE 1 - OPERATING LIMITS - (cont'd) i KEYSWITCH BYPASS CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION
]
- 4. Reactor Power-Power-flow safety Table 4.0.2.2.1-1 Full scram One channel may be by-Flow Rate channel 1 or 2 passed for calibration Abnormal and testing.
- 5. Reactor Power-flow safety Table 4.0.2.2.1-1 Full scram One channel may be by-Coolant Flow channel 1 or 2 passed for calibration Rate Low and testing.
- 6. Reactor Water Water level safety Table 4.0.2.2.1-1 Full scram One channel may be by-Level High channel 1 or 2 passed for calibration and testing.
(Nominal indicated unvoided saturated water level shall be permitted to vary from 2'9" above the fuel to up 4'6" above the fuel during reactor heatup and operation.
i
- 7. Reactor Water Water level safety
<12" below nominal
- 1) Full scram One channel of Item 7 Level Low channel 7 nr 2 indicated level
- 2) Initiation of high or channel 3 of Item 7A pressure core may be bypassed for 4
spray pumps calibration and testing.
- 3) Closure of reactor building steam isolation valve and its bypass
- 4) Closure of reactor blowdown through decay heat removal valve 4
- 5) Start 1A and IB diesel generators 1
- 6) Closure of shutdown condenser conden-sate drain valve LACBWR 5-12 WP4.12
TABLE 1 - OPERATING LIMITS - (cont'd)
KEYSWITCH BYPASS CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION
- 7. Reactor Water
- 7) Closure of ventila-Level Low tion inlet and outlet dampers (continued)
- 8) Closure of contain-ment offgas vent header valve
- 9) Closure of heating steam condensate return valve
- 10) Closure of retention tank pump discharge l
valve 7A Reactor Water Water level safety
< 12" below
- 1) Full scram One channel of Item 7 Level Low channel 3 nominal indicated
- 2) Initiation of high or channel 3 of Item 7A level pressure core may be bypassed for spray pumps calibration and testing.
- 8. Main Condenser Vacuum switches 2> 19" hg
- 1) Full scram
- 1) One channel may be Vacuum Low 1 or 2
- 2) Closure of reactor bypassed during cali-building steam bration and testing.
isolation valve
- 2) May be bypassed during plant startup and shutdown.
- 9. Reactor Build-Reactor building
]>,90% full open
- 1) Full scram
- 1) May be bypassed ing Steam steam isolation travel
- 2) Shutdown condenser during testing.
Isolation valve closure operates
- 2) May be bypassed Valve Not relays 1 or 2 during plant startup or shutdown.
Fully Open LACBWR 5-13 WP4.12
TABLE 1 - OPERATING LIMITS - (cont'd)~
KEYSWITCH BYPASS CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION
- 10. Turbine Turbine building
> 90% full open
- 1) Full scram
- 1) May be. bypassed i
Building Steam steam isolation travel
- 2) Shutdown condenser during testing.
Isolation valve closure operates
- 2) May be bypassed Valve Not relays.1 or 2 during plant startup-Fully Open or. shutdown.
- 11. Turbine Stop Limit switch Table 4.0.2.2.1-1 Partial scram
- 1) May be bypassed Valve Not during testing.
Fully Open
- 2) May be bypassed when-ever the turbine load i
is less than 10 Mwe.
i I
- 12. Low 011 Level Limit switches Table 4.0.2.2.1-1 Partial scram
- 1) May be bypassed In Any Control during testing.
j Rod Drive
- 2) May be bypassed prior Accumulator to withdrawing con-trol rods in order to charge accumulators.
- 13. Low Cas Pressure switches Table 4.0.2.2.1-1 Partial-scram
- 1) May be bypassed 1
Pressure In during calibration Any Control and testing.
Rod Drive
- 2) May be bypassed prior j
Accumulator to withdrawing con-i trol rods in order to
{
charge accumulators.
i 1
?
LACBWR 5-14
]
WP4.12 2
T "
t-'E
TABLE 1 - OPERATING LIMITS -- (cont'd)
KEYSWITCH BYPASS CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION
- 14. Low Voltage 2400 v bus 1A under-Table 4.0.2.2.1-1 Partial scram None.
(for a time voltage relay 1 or 2 longer than or required for 2400 v bus 1B under-Reserve Feed voltage relay 1 or 2 Breakers to operate auto-2400 v bus 1A under-Table 4.0.2.2.1-1 Full scram
- None, matica11y) voltage relay 1 and-2400 v bus 1B under-voltage relay 1 or 2400 v bus 1A under-voltage relay 2 and 2400 v bus 1B under-voltage relay 2 Reactor building Table 4.0.2.2.1-1 Full scram None.
motor control center 1A relay 1 or 2 Turbine building Table 4.0.2.2.1-1 Full scram None.
motor control center 1A relay 1 or 2
- 15. Low Main Main steam pressure
> 1000 psig Closure of reactor May be bypassed during Steam transmitter building steam plant startup and shut-Pressure isolation valve
.down.
LACBWR 5-15 WP4.12
- ~
TABLE 1 - OPERATING LIMITS - (cont'd)
KEYSWITCH BYPASS CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION
- 16. Reactor Reactor building
< 5 psig
- 1) Initiation of high None.
Building pressure trans-pressure core Pressure mitter 1 or 2 spray pump
- High
- 2) Initiation of alternate core spray pump
- 3) Closure of ventilation inlet & outlet dampers
- 4) Closure of containment off gas vent header valve
- 5) Closure of retention tank pump discharge valve
- 6) Closure of shutdown condenser condensate drain valve
- 7) Closure of reactor blowdown through decay heat removal valve-
- 8) Closure of containment high pressure service water valve l
- 9) Closure of containment demineralized water valve
- 10) Closure of containment heating steam conden-sate valve.
LACBWR 5-16 WP4.12
TABLE 1 - OPERATING LIMITS - (cont'd) 4 KEYSWITCH BYPASS CONDITION CHANNEL OR SENSOR SET POINT ACTION PROVISION
- 17. Off gas Radiation monitor j[ gaseous activity Diversion effluent None.
Holdup Tank levels which gas to the storage Effluent correspond to tanks Activity the limits of High Specification 4.3.2.2
- 18. Reactor Radiation monitors j[ radiation levels
- 1) Closure of ventila-None.
Building which correspond tion inlet and Ventilation to the limits of outlet dampers Exhaust Specification
- 2) Closure of contain-4.3.2.2 ment off-gas vent header valve
- 19. Simultaneous Pressure trans-25-30 psig and Opening of diaphragm None.
Low Reactor mitter and water j[ 12" below valve allowing water Pressure and level safety nominal indicated to flow directly from i
Low Water channel 1 or 2 level overhead storage tank Level to core spray nozzles l
- 20. Simultaneous Reactor _ building j[5 psig and Opening of motor _
None.
High Reactor pressure transmitter j(12" below operated valves and l
t Building 1 or 2 and reactor nominal indicated start of engine-driven Pressure and water-level safety level pumps of alternate Reactor Low channel 1 or 2 core spray system
+
Water Level i
1 4
LACBWR 5-17 WP4.12 4
I
TABLE 1 - OPERATING LIMITS - (cont'd) i KEYSWITCH BYPASS CONDITION CRANNEL OR SENSOR SET POINT ACTION PROVISION
- 21. High Reactor 3 reactor pressure
< 1350 psig or Trip of both recirc-The protective function Pressure or or 3 reactor level
< 30 inches below ulation pump breakers can be bypassed whenever Low Reactor transmitters nominal indicated the reactor is shutdown Water Level level or recirculation pump operation is required for safety reasons.
- 22. Steam Safety Position switches Ope n-close None post accident None.
Valves Not on each of the three indication only Fully Closed inservice safety valves 4
1 i
LACBWR 5-18 WP4.12
- ,. s g INISTRATIVE CONTROLS 6.9.3 Unique Reporting Requirements (continued) b.
Annual Radiological Environmental Operating Report Routine radiological environmental operating reports covering the operation of the plant during the previous calendar year shall be submitted prior to May 1 of each year.
The annual radiological environmental operating reports shall include summarized and tabulated results, using Regulatory Guide 4.8, dated December 1975, as guidance, of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be
-submitted noting and explaining the reasons for the missing results.
The missing data shall be submitted as soon as possible in a subsequent report.
The reports shall also include the following: a summary description of the radiological environmental monitoring program and a map of all sampling locations keyed to a table giving distances and directions from the plant.
The tabulated results and sample analyses of the radiological environmental monitoring program may be submitted in the same report with semi-annual radioactive effluent releases in accordance with Regulatory Guide 4.1 dated January 18, 1973.
LACBWR 6-16a WP4.12
- =
- g NISTRATIVE CONTROLS ~
'6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.1 The following records shall be retained for at least five years:
Records and logs of facility operation covering time interval at a.
each power level.
b.
Records and logs of principal ma'intenance activities, inspections, repair and replacement of principal items of equipment related to
. nuclear safety.
c.
All REPORTABLE OCCURRENCES submitted to the Commission.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
i Records of changr.s made to the procedures required by Specification e.
6.8.1.
f.
Records of radioactive shipments.
g.
Records of sealed source and fission detector leak tests and results.
d h.
Records of annual physical inventory of all sealed source material 4
of record.
6.10.2 The following records shall be retained for.the duration of the LACBWR Operating License:
a.
Records and drawing changes reflecting facility design modifications made to systems and equipment described in the Safeguards Report for Operating Authorization.
b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
Records of radiation exposure.for all individuals entering radiation c.
~
control areas.
d.
Records of gaseous and liquid radioactive material released to the environs, and records of analyses required by the Radiological Environmental Monitoring Program.
+
LACBWR 6-17 WP4.12
. -