ML030690267
| ML030690267 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 03/07/2003 |
| From: | Chandu Patel NRC/NRR/DLPM/LPD2 |
| To: | Scarola J Carolina Power & Light Co |
| References | |
| TAC MB6230 | |
| Download: ML030690267 (6) | |
Text
DEFINITIONS T - AVERAGE DISINTEGRATION ENERGY 1.12 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 (MeV d) for isotopes. with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
ENGINEERED SAFETY FEATURES RESPONSE TIME 1.13 The ENGINEERED SAFETY FEATURES (ESF) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF Actuation Setpoint its at the channel sensor until the ESF equipment is capable of performing pump safety function (i.e.. the valves travel to their required positions, discharge pressures reach their required values. etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential.
overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.
EXCLUSION AREA BOUNDARY 1.14 The EXCLUSION AREA BOUNDARY shall be that line beyond which the land is not controlled by the licensee to limit access.
FREQUENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
GASEOUS RADWASTE TREATMENT SYSTEM 1.16 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolantforsystem the off-gases from the primary system and providing for delay or holdup purpose of reducing the total radioactivity prior to release to the environment.
IDENTIFIED LEAKAGE 1.17 IDENTIFIED LEAKAGE shall be:
- a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured and conducted o a sump or collecting tank. or
- b. Leakaqe 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 LEAKAGE, or
- c. Reactor Coolant System leakage through a steam generator to the Secondary Coolant System.
SHEARON HARRIS - UNIT 1 1-3 Amendment No. 112
DEFINITIONS PROCESS CONTROL PROGRAM 1.25 The PROCESS CONTROL'PROGRAM (PCP) shall contain the current formulas.
sampling, analyses., test, and determinations to be made to ensure that processing and packaging Of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure.compliance with 10 CFR Parts 20. 61. and 71 and State regulations, burial around requirements. and other requirements governing the disposal of solid radioactive waste.
PURGE - PURGING 1.26 PURGE or PURGING shall be any controlled process of discharging air or gas from a confinement to maintain temperature, pressure. humidity.
concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.
QUADRANT POWER TILT RATIO 1.27 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs. whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
RATED THERMAL POWER 1.28 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2900 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.29 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its Trip Setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential. overlapping, or total steps so that the entire response time is measured. In lieu of measurement. res onse time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.
REPORTABLE EVENT 1.30 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 of 10 CFR Part 50.
SHUTDOWN MARGIN 1.31 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.
SITE BOUNDARY 1.32 For these Specifications, the SITE BOUNDARY shall be identical to the EXCLUSION AREA BOUNDARY defined above.
SHEARON HARRIS - UNIT 1 1-5 Amendment No. 112
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3-1.
ACTION: As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channels and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit, specified in the Technical I Specification Equipment List Program. plant procedure PLP-106, at least once per 18 months. Each verification shall include at least one train such that oth trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once every N times I
18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.
SHEARON HARRIS - UNIT 1 3/4 3-1 Amendment No. 112
INSTRUMENTATION ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within its limit specified in the Technical I Specification Equipment List Program. plant procedure PLP-106. at least once per 18 months. Each verification shall include at least one train such that oth trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once per N times 18 L
months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" column of Table 3.3-3.
SHEARON HARRIS - UNIT 1 3/4 3-17 Amendment No. 112
INSTRUMENTATION BASES REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSILM INSIRUMENIAHIUN Tontinued)
Z + R + S RTA. the interactive effects of the errors in the rack and the sensor, and the "as measured values of the errors are considered. Z, as specified in Table 3.3-4. in percent span .is the statistical summation of errors assumed ip the analysis e ciudilig those associatedawith tMy sensor and rack drift and tne accuracy of tHeir measurement. TA or Total A owagce is the differ nce, in percent span. between the trip setpoint and the value used in the.analysis for tne actuation. R or Rack Nro is ne 'as measured" deviaýion, in the pergent span. Tor the affected channel from twe.specified Trip Setpoint. S r Sensor Error is either the "as me.aued" d viat on of the sensor from its capi ratioy point or the value specified in Table 3.,-4. in ercent span from the analysis assumptions. use of Equation 3*3-1 allows for aseno draft factor*, an increased r Rckli ft factor. and provides a thresho d va1ue for determination of OPERABILITY.
The methodology to.derive the T ip Setpoints is based upon combininq all of tie uncertainties in the channelis. tnherent to the determination of the Tri Set oints are the magnitudes of these channel uncertainties. Sensor and rack inszrumentation uti lzed in these channels are expected to be capable of operating within t al wances of these uncertaihty magnitudes. Rack drift in excess f the Abiowabie Va ue exhibits the behavior at the rack has not mejei~s allowance. Being that there is a small statistica chance that this Wll happeno ap infequent encessive drift is expected Rack or sensor drift.
in excess OT the allowance tna is more tnan occasional, may be indicative of more serious problems and should warrant further investigation.
The measurement of response time at the specified frequencies rovides assurance that the reactor trip and the ýhglneered Safety ýeatures actuation associated with each channel its complete within the time limit assumei in.the safety analyses. ,No cre it was taken in the analyses for those ch nnels with response times indicated as not applicable. Response time may be demonstrated ny series of sequenti&a. over'npp rg,.or *ot~i channel test measurements provided that such tests demonstrate the total c annel response time as defined. Response time may be verified by actual response time tests in any series of sequyntia1. overlapping or total channel measurements or by the summation of a llcated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel.
Allocations for sensor response times may be obtained flom: (1) historical records based on acceptaboe response time tests (hydraulic. noi e or power interrupttests);o*) in~ace, onste, e or Tfs!*tee.g, .vendor) tes.
PeAsuRemens.,, t.Ulizino ven ngineering specficalions. W AP-13632 P-A. Rev.y. z, ,,* nai ressure Sensor eponse Time es ing Requirements. provides the basis and methodo' oy for using allocated sensor response times inthe oveoral veif'i on of the channel response time for specific sensors identified in t-e0 WeAPs R e e Cation TOr other sensor types must be demonstrated Dy test.
WCAP X4036-P-A. Rev. J. "El imiation of Period'G Respqnse TiLe Tests."
ctuatiopthe rovides log basis and methodology c response times ip thy oc*ted signal using alveriTication Toroverall of UTocessing and protection system channel response time Tine a locations for sensor. sigqa .i conditioning. and actuation logic response times must be *erTiedpror to placing the compoge~t into operational service and re-verified fol'owi ng maintenance or mod]i Cation hat may adversely aTfect response time. In general, eRlectrica repair work does not impact response tMen provided the tsr U for the repair are the same typ e and vaMWe. Speci fc components loenrTfied in the WCA Pmay be replaced without verifCation testing, un ,
example where response time cou ld e affected is replacing the sensing element of a transmitter.
SHEARON HARRIS - UNIT 1 B 3/4 3-2 Amendment No.l12
INSTRUMENTATION BASES REACTOR TRIP SYSTEM INSTRUMENTATION AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSIRUMENTATION (Continued)
The Engineered Safety Features Actuation System senses selected plant param eters and determines whether or not predetermined limits are being exceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents events, and transients. Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss-of-coolant accident: (1)charging/safety injection pumps start and automatic valves position, (2) reactor trip, (3) feedwater isolation. (4) startup of the emergency diesel generators. (5) containment spray pumps start and automatic valves position (6) containment isolation. (7) steam line isolation, (8) turbine trip, (9)auxiliary feedwater pumps start and automatic valves position, (10) containment fan coolers start and automatic valves position, (11) emergency service water pumps start and automatic valves posit ion, and (12) control room isolation and emergency filtration start.
SHEARON HARRIS - UNIT 1 B 3/4 3-2a Amendment No. 112 1