ML20080S574
| ML20080S574 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 02/22/1984 |
| From: | Dixon O SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| GL-83-43, NUDOCS 8402290174 | |
| Download: ML20080S574 (27) | |
Text
.
<4 SOUTH CAROLINA ELECTRIC & GAS COMPANY POCT OFPICE 7s4 COLuustA. south CAROUNA 29218
- o. w. oixou an.
February 22, 1984 v.cr m*'aar Nuc0Am OpamATeoN$
Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear. Regulatory Commission Washihgton, D.C.
20555
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 Operating License No. NPF-12 NRC Generic Letter 83-43 Reporting Requirements
Dear Mr. Denton:
On December 29, 1983, South Carolina Electric and Gas Company (SCE&G) received NRC Generic Letter 83-43, " Reporting Require-mencs of 10rVR Part 50, Sections 50.72 and 50.73, and Standard Technical Spacifications."
In this letter, SCE&G was requested to make modifications in the wording of the Virgil C.
Summer j
Naclear Station Technical Specifications to reflect the changes in reporting requirements as outlined in Section 50.73 of Title 10 of the Code of Federal Regulations.
Attached are the marked up pages to the Technical Specifica-tions which incorporate the revisions neaded throughout the document to reflect the revised reporting requirements.
Since the margin of safety for these changes in reporting requirements has previously been approved by the NRC, and because of the administrative nature of this request, SCE&G has determined that a finding of no significant hazards consideration is appropriate.
This change has been reviewed and approved by the Plant Safety Review Committee (PSRC) and the Nuclear Safety Review Committee (NSRC).
As stated in the Generic Letter, a license fee is not required because this change is clarifying in nature and made at the request of the Commission.
6402290174 840222 PDR ADOCK 05000 t
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'Mr.. Harold R..Denton
.NRC Generic-Letter 83-43
. February 22,7 1984 Page #2 I
t Should.there be any questions, please call us at your convenience.
Very truly yours s
O. W.
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AM OWD/fjc Attachments
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C.
Summcr l
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C.
Nichols, Jr./O. W. Dixon, Jr.
E. ' H. Crews, 'Jr. -
E.-C.
Roberts W..A.: Williams, Jr.-
D.
A.
Nauman J.
P. O'Reilly Group Managers O.
S. Bradham C.
A. Price C.E L. Ligon (NSRC)
K. - E. Nodland c-R.
A.
Stough-G.
Percival C. W. Hehl-J.
B.' Knotts,-Jr.
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INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS 1.1 ACTI0N...................................................
1-1
- 1. 2 ACTUATION LOGIC TEST.....................................
1-1
- 1. 3 ANALOG CHANNEL OPERATIONAL TEST..........................
1-1 1.4 AXIAL FLUX DIFFERENCE....................................
1-1 1.5 CHANNEL CALIBRATION......................................
1-1 1.6 CHANNEL CHECK.....................................,......
1-1 1.7 CONTAINMENT INTEGRITY....................................
1-2 1.8 CONTROLLED LEAKAGE.......................................
1-2 1.9 CORE A LTE RAT ION..........................................
1-2 1.11l0SEEQUIVALENTI-131....................................
1.10 1-2
- -AVERAGE DISIN TEGRATION EMERGY..........................
1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME.................
1-3 1.13 FREQUENCY N0TATION.......................................
1-3 1.14 GASEOUS RADWASTE TREATMENT SYSTEM........................
1-3 1.15 IDENTIFIED LEAKAGE.......................................
1-3 1.16 MASTER RELAY TEST........................................
1-3 1.17 0FFSITE DOSE CALCULATION MANVAL (0DCM)...................
1-4 1.18 OPERABLE - OPERABI LITY...................................
1-4 1.19 OPERATIONAL MODE - M0DE..................................
1-4 1.20 PHYSICS TESTS............................................
1-4 1.21 PRESSURE BOUNDARY LEAKAGE................................
1-4
- 1. 22 PROCESS CONTROL PROGRAM (PCP)............................
1-4 1.23 PURGE-PURGING............................................
1-4 1.24 QUADRANT POWER TILT RATI0................................
1-5 1.25 RATED THERMAL P0WER......................................
1-5 EVEbrr 1.26_ REACTOR TtIP SYSTEM RESPONSE TIME........................
1-5
- 1. 2 7 REPORTAB L E'-000tHHtENGE....................................
1-5 1.28 SHUTDOWN MARGIN..........................................
1-5 1.29 SLAVE RELAY TEST.........................................
1-5 1.30 SOLIDIFICATION...........................................
1-5 1.31 SOURCE CHECK.............................................
1-5 1.32 STAGGERED TEST BASIS.....................................
1-6 1.33 THERMAL P0WER............................................
1-6 1.34 TRIP ACTUATING DEVICE OPERATIONAL TEST................... 6 1.35 UNIDENTIFIED LEAKAGE.....................................
1-6 1.36 VENTILATION EXHAUST TREATMENT SYSTEM.....................
1-6 1.37 VENTING..................................................
1-6 TABLE 1.1 OPERATIONAL M0 DES...................................
1-7 TABLE 1.2 FREQUENCY N0TATION..................................
1-8 SUMER-UNIT 1 I
4 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE Review...............................................
6-9 Audits......................................................
6-10 Authority...................................................
6-10 Records.....................................................
6-11 6.5.3 TECHNICAL REVIEW AND CONTROL Activities..................................................
6-11 EVF_ N T-6.6 REPORTAB LE SEetHHtENGE ACTI0N..................................
6-12 6.7 SAFETY LIMIT VIOLATION........................................
6-12 6.8 PROCEDURES AND PR0 GRAMS.......................................
6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS n.m-m,..--
m.,P ov nuun.n-mu vuuvnnunuus Startup Report..............................................
6-14a Annual Report...............................................
6-15 Annual Radiological Environmental Operating Report..........
6-16 Semiannual Radioactive Effluent Release Report..............
6-16 Monthly Operating Report....................................
6-18
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P in m.k2 or...................
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m Radial Peaking Factor Limit Report..........................
6-29IB 6.9.2 SPECIAL REP 0RTS.............................................
6-t&ts 6.10 RECORD RETENTION.............................................
6-31 19 6.11 RADIATION PROTECTION PR0 GRAM.................................
6-22 to 6.12 HIGH RADIATION AREA..........................................
6-BE 10 SUMMER-UNIT 1 XIX
INDEX ADMINISTRATIVE CONTROLS SECTION PAGE
- 6.13 PROCESS CONTROL PROGRAM................................,.....
6-24 11 6.14 -0FFSITE DOSE CALCULATION MANUAL..............................
6-23 0 6.15 MAJOR CHANGES TO RADI0 ACTIVE WASTE TREATMENT SYSTEMS.........
6-24 O t
1 4
f SUPMER-UNIT 1 XX
3-DEFINITIONS QUADRANT POWER TILT RATIO E
1.24 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector cali-brated outpdts, or.the ratio of the maximum lower excore detector calibrated j
output to the average'of the lower excore detector calibrated outputs, whichever l
is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.
{
RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2775 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.26 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.
REPORTABLE 96C4HHENGE EVENT E v E t4 7 1.27 A REPORTABLE SeeWRRENeE shall be any of those conditions specified in I
C,eeificatiea; 0.0.1.12 and 0.0.1.13. s ec fion 50.73 To 10 CFR Pact 50-i SHUTDOWN MARGIN 1.28 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 full length 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.
' SLAVE RELAY TEST
.1.29-A SLAVE RELAY TEST shall be the energization of each slave relay and verification of OPERABILITY of each relay.
The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.
SOLIDIFICATION 1.30 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid systems to a uniformly distributed, monolithic, immobilized solid with definite volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).
SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
SUMER - UNIT 1 1-5
= _ _ -. -.
e 2.2 LIMITING SAFETY SYSTEM SETTINGS i
BASES 2.2.1 R'iACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.?-1 are the nominal values at which the Reactor Trips are set for each functional unit.
The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceedirig their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences 1
of accidents.
The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.
t i
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allowable Values for the reactor trip setpoints have been specified in Table 2.2-1.
Operation with setpoints less conservative than the Trip Setpoint J
but,within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
An optional provision has been included for determining the OPERABILITY of a chae.iel when its trip setpoint.is found to exceed the Allowable Value.
The methodology of this option utilizes the "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combina-tion of the other uncertainties of the instrumentation to measure the process variable and the uncertainties in calibrating the instrumentation.
In r
- Equation 2.2-1,'Z + R + S < TA, the interactive effects of the errors in the rack and the sensor, and tee "as measured" values of the errors are considered.
Z, as specified.in Table 2.2-1, in percent span, is the statistical summation of errors assumed in the analysis excluding those associated with the sensor and rack drift-and the accuracy of their measurement.
TA or Total Allowance is the difference, in percent span, between the trip setpoint and the value used in-the analysis for reactor trip.
R or Rack Error is the "as measured"
- deviation, in percent span, for the affected channel from the specified trip setpoint.
S or Sensor Error is either the "as measured" deviation of the sensor-from its calibration point or the value specified in Table 2.2-1, in percent span, from the analysis assumptions.
Use of Equation 2.2-1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE 66C4RRENGE6.
EVENTS.
The methodology to. derive the trip setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the i
trip setpoints are the magnitudes of these channel uncertainties.
Sensors and other instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not l'~
will happen, an infrequent excessive drift is expected.
Rack or sensor drift, met its allowance.
Being that there is a small statistical chance that this
.in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
SUDMER - UNIT 1 B 2-3
POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)
R 2.
When the F is less than or equal to the F limit for the x
appropriate measured core plane, additional power distribution RTP maps shall be taken and F compared to F and F at least x
once per 31 EFPD.
e.
The F limits for RATED THERMAL PCWER (FRTP) shall be provided for xy x
all core planes containing bank "D" control rods and all unrodded core planes in a Radial Peaking Factor Limit Report per Specification 6.9.1.M.
11 f.
The F limits of e., above, are not applicable in the following core planeIYregions as measured in percent of core height from the bottom of the fuel:
1.
Lower core region from 0 to 15%, inclusive.
2.
Upper core region from 85 to 100%, inclusive.
3.
Grid plane regions at 17.8 1 2%, 32.1 1 2%, 46.4 1 2%,
60.6 1 2% and 74.9 1 2%, inclusive.
(17 x 17 fuel elements).
4.
Core plane regions within i 2% of core height (t 2.88 inches) about the bank demand position of the bank "D" control rods.
g.
With F exceeding F the effects of F n F (Z) shall be evaluated xy q
to determine if F (Z) is within its limits.
q 4.2.2.3 When F (Z) is measured for other than F determinations, an overall 9
xy measured F (Z) shall be obtained from a power distribution map and increased 9
by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.
SUP9tER - UNIT 1 3/4 2-6
v.
~
INTERENTATION RADI0ACTtw tma;g EFFLUENT NONITORING INSTRUMENTATION
..w+
b LIMITI4 CO MITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels shown in Table 3.3-12 shall be OPERA 8LE with their alare/ trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded.
The alarm /
trip setpoints of these channels shall he determined in accordance with the 0FFSITE DOSE CALCULATION MANUAL (ODCN).
APPLICA8ILITY:
At all times.
ACTION:
s.
With a radioactive liquid effluent monitoring instrumentation channel alars/ trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid affluents monitored by the affected channel or declare the channel inoperable.
b.
With less than t.he minimum number of radioactive liquid affluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-12.
Additionally if this condition prevails for more than 30 days, in the next semiannual effluent report, explain why this condit. ion was not corrected in a timely manner.
c.
The provisions of Specifications-&.0.1.12.L, 3.0.3 and 3.0.4 are not applicable.
1 SURVE[LLANCEREQUIREMENTS 4.3.3.8.1 Each radioactive liquid e'ffluent monitoring instrumentation channel shall be demonstrated OPERA 8LE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-8.
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4 SUteER - UNIT 1 3/4 3-67 I
INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION
~
3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded.
The alarm / trip setpoints of these channels shall be determined in accordance with the ODCM.
APPLICABILITY:
As shown in Tabla 3.3-13 ACTION:
a.
With a radioactive gaseous effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.
b.
With less than the minimum r. umber of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13.
Additionally if this condition prevails for more than 30 days, in the next semiannual effluent report, explain why this condition was not corrected in a timely manner.
c.
The provisions of Specifications C.^.1.12.b, 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS f
4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel l
shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE l
CHECK, CHANNEL CALIBRATION and ANA',0G CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-9.
SUMMER - UNIT 1 3/4 3-73 l
l
7.
REACTOR COOLANT SYSTFM k
SURVEILLANCE REQUIREMENTS (Continued) 9.
Preservice Inspection means an inspection of the full length of
.f)/)
each tube in each steam generator performed by eddy current
\\.
techniques prior to service to establish a baseline condition
-i -
of the tubing. This inspection shall be performed after the i
field hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be useo during subsequent inservice inspections.
g,.
- Q. b b.
The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.
4.4.5.5- -Reports a.
Within 15 days following the completion of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam
. generator shall be reported to the Commission in a Special Report pursuant to Specification 6.9.2.
b._
The complete results of the steam generator tube inservice inspection 4
shall be submitted to the Commission in a Special Report pursuant to
- Specification 6.9.2 within 12 months following the completion of the
- inspection. This Special Report shall include:
1.
Number and extent of tubes inspected.
2.~
Location and percent of wall-thickness penetration for each indication of an imperfection.
3.
Identification of tubes plugged.
c.
- Results of steam generator tube inspections which fall into
. Category C-3 and require prompt notification of the Commission shall be reported oursuant toJ pecificati;r. S.0.1 prior to resumption of c-plant operation.
T 6 w it;.e., feilam.p sf this rep;rt :h:11 p rovide
' o M50, n(Q 24) t a description of investigations conducted to determine cause of the
' tube degradation and corrective measures taken to prevent recurrence.
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SUMER - UNIT 1 3/4 4-15
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' STEAM GENERATOR TUBE INSPECTION
!ii 8
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-4 IST SAMPLE INSPECTION 2ND SAMPLE IN3PECTION 3RD SAMPLE INSPECTION g
Sample Site Result '
Action Required Result Action Required Result i Action Required A minimum of C-1 None N/A N/A N/A
'N/A S Tubes per g
S. G.
C-2 Plug defective tubes C-1 None N/A N/A and inspat additional Plug defective tubes C-1 None 2S tubes in this S. G.
C-2 and inspect additional C-2 Plug defective tubes M tubes in this S. G.
Perform cetion for C-3 C-3 result of first sample Perform action for w
C-3 C-3 result of first N/A N/A g
sample i
C-3 Inspect all tubes in All other d
this S. G., plug de-S. G.s are None N/A N/A fective tubes and C-1 SP'CI I"
I" Some S. G.s p,,. form action for N/A N/A C-2 but no C-2 result of second additional sample Prompt notification S. G. are to NRC pursuant C-3 to :. ::$ :f:-
Additional Inspect all tubes in N'CNON S. G. is C-3 each S. G. and plug M M*d o^d '0GR defective tubes.
So.7.3(a) L(i~g)
Prompt notification N/A N/A to NRC pursuant to :;--T::M:n
&9:410CFR50O1%
4 ) na soc,F K bo. 7.:La) I pi)
Where N is the number of steam generators in the unit, and n is the number of steam generators inspected S=3 during an inspection n
,f'
REACTOR COOLANT SYSTEM ACTION:
(Continued)
MODES 1, 2, 3, 4 and 5:
With the specific activity of the primary coolant greater than_1.0
-a.
microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E
(
microcuries per gram, perform the sampling and analysis requirements s
of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.
A REPORTAiH:E-GGGURAEE45pec.al
~
Report shall be prepared and submitted-to the Commission $ purr, sert te Spectri w,S L So dap n.;.;,,G.0.1.
This report shall contain the results of the specific activity analyses together with the following information:
(
1.
Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded, 2.
Fuel burnup by core region, 3.
Clean-up' flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded,
-4.
History of de gassing operations, if any, starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded, and 5.
The time duration when the ' specific activity of the primary
[
coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENT I-131.
(
j SURVEILLANCE REQUIREMENTS l
4.4.8.The s'pecific activity of the primary ;oolcnt shall be determined to be within the limits by performance of the sampling and analysis program of l
Table 4.4-4.
I
(,
SUN 4ER - UNIT 1 3/4 4-26 i
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RADI0 ACTIVE EFFLUENTS LIQUID I!OLDUP TANXS LIMITING CONDITION FOR OPERATION 3.11.1.4 The quantity of radioactive material contained in each of the following tanks shall be limited to less than or equal to 10 curies, excluding tritium and dissolved or entrained noble gases, a.
Candensate Storage Tank b.
Outside Temporary Storage Tank APPLICABILITY:
At all times.
ACTION:
a.
With the quantity of radioactive material in any of the above listed tanks exceeding the above limit, immediately suspend all additions or ra6ioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.
b.
The provisions of Specifications 0.0.1.12.b, 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.1.4 The quantity of radioactive material contained in each of the above listed tanks shall be determined to be within the above limit by analyzing a represantative sample of the tank's contents at least once per 7 days when radioactive materials are being addad to the tank.
SUMMER - UNIT 1 3/4 11-7 l
RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION
-3.11.2.5 The concentration of oxygen in the waste gas holdup system shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration excee.ds 4% by volume.
APPLICABILITY:
At all times.
ACTION:
a.
With the concentration of oxygen in the waste gas holdup system greater than 2% by volume but less than or equal to 4% by volume, restore the concentration of oxygen to within the limit within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, b.
With the concentration of oxygen in the waste gas holdup system greater than 4% by volume, itenediately suspend all additioat of waste gases to the system and reduce the concentration of oxygen tn less than 4% by volume within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and less than or equal to 2%
by volume within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
c.
The provisions of Specifications 0.0.1.12.b, 3.0.3 and ?.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.11.2.5 The concentration of hydrogen and oxygen in the waste gas holdup system shall be determined to be within the above limits by continuously monitoring the waste gases in the waste gas holdup system with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of Specification 3.3.3.9.
I SUMMER - UNIT 1 3/4 11-17 l
~
3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 7
3/4.12.1 MONITORING PROGRAM b
LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted
_l as specified in Table 3.12-1.
APPLICABILITY:
At all times.
ACTION:
a.
With the radiological environmental hcnitoring program not being conducted as specified in Table 3.12-1, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission, in the Annual Radiological Operating Report, a dcscription of the reasons for not conducting the program as required and the plans for preventing a recurrence.
b.
With the level of radioactivity in an environmental sampling medium exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, in lieu of any other report required by Specification 6.9.1, prepare and submit to the Commission within spedo) 30 days from the end of the affected calendar quarter a* Report.;;unu=t te Spcificetieri 0.0.1.13.
When more than one of the radionuclides in Table 3.12-2 are detected in the sampling medium, this report shall be submitted if:
concentration (1), concentration (21 + ***> 1.0 limit level (1) limit level (2)_
When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose to an individual is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3.
This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the
{-
Annual Radiological Environmental Operating Report.
l c.
With milk or fresh leafy vegetable samples unavailable from one or l
more of the sample locations required by Table 3.12-1, in lieu of any other report required by Specification 6.9.1, prepare and submit I
i to the Commission within 30 days, pursuant to Specification 6.9.2, a
(...
Special Report which identifies the cause of the unavailability of samples and identifies locations for obtaining replacement samples.
The locations from which samples were unavailable may then be deleted from those required by Table 3.12-1, provided the locations from which the replacement samples were obtained are added to the environmental monitoring program as replacement locations.
d.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SUMER UNIT 1 3/4 12-1
o POWER DISTRIBUTION LIMIT BASES HEAT FLUX HOT CHANNEL FACTOR and RCS FLOWRATE and NUCLEAR ENTHALPY RISE MT CHANNEL FACTOR (Continued)
The control rod insertion limits of Specifications 3.1.3.5 and c.
3.1.3.6 are maintained.
d.
The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.
F will be maintained within its limits provided conditions a. through H
- d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow rate and F" may be " traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the measured F is also low) to ensure that the g
calculated DNBR will not be below the design DNBR value.
The relaxation of F
as a function of THERMAL POWER allows changes in the radial power shape H
for all permissible rod insertion limits.
R, as calculated in 3.2.3 and used in Figure 3.2.3, accounts for F t
g less than or equal to 1.49.
This value is used in the various accident analyses where F influences parameters other than DNBR, e.g., peak clad H
temperature and thus is the maximum "as measured" value allowed.
R, as 2
-defined, allows for the inclusion of a penalty for rod bow on DNBR only.
Thus N
knowing this "as measured" values of F and RCS flow allows for " tradeoffs" 3H in excess of R equal to 1.0 for the purpose of offsetting the rod bow DNBR penalty.
When an F measurement is taken, an allowance for both experimental error 9
and manufacturing tolerance must be made.
An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.
The radial peaking factor Fxy(Z) is measured periodically to provide assurance that the hot channel factor, F (Z), remains within its limit.
The 0
F limit for Rated Thermal Power (F
) as rovided in the Radial Peaking x
Factor Limit Report per. specification 6.9.1.1+was determined from expected power control maneuvers over the full range of buraup conditions in the core.
When RCS flow rate and F are measured, no additional allowances are H
necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4.
Measurement errres of 3.5% for RCS total flow rate and 4% for F have been g
allowed for in determination of the design DNBR value.
SUMMER - UNIT 1 B 3/4 2-4 e
INSTRUMENTATION BASES w
REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM
. INSTRUMENTATION (continued) setpoint.
S or Sensor Error is either the "as measured" deviation of the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions.
Use of Equation 3.3-1 allows for
-a sensor drift factor, an increased rack drift factor, and provides a threshold i
~value for REPORTABLE 466tfRRENGE6. EVENTS.
The methodology to derive the trip setpoints is based upon combining all
-of the uncertainties ir, the channels.
Inherent to the determination of the trip setpoints'are the magnitudes of these channel uncertainties.
Sensor and
~ rack instrumentation utilized in these channels are expected to be capable of 3
-operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
t The measurement of response time at the specified frequencies provides assurance-that the reactor trip and the engineered safety feature actuation associated with each channel is completed within the time limit assumed in the accident analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be demon-t strated by any series of sequential,. overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defi'ned.
Sensor response time verification may be demonstrated by either 1) in place, onsite, or offsite test measurements or 2) utilizing replacement sensors with certified response times.
The Engineered Safety Features-Actuation System senses selected plant parameters ~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) safety injection pumps start and automatic valves pos*+ ion, 2) reactor trip, 3) feed-water isolation, 4) startup of the emergency diesel generators, 5) containment spray pumps start and automatic valves position, 6) containment isolation,
[
L7) steam.line isolation, 8) turbine trip, 9) auxiliary feedwater pumps start l'
and automatic va ves position, 10) containment cooling fans start and auto-matic talves position, 11) essential service water pumps start and automatic
- valves position, and 12) control room isolation and ventilation systems start.
l-i i
- SUMMER - UNIT 1 B 3/4 3-la
BASES 3/4.4.5~ STEAM GENERATOPS The Surveillance Requirements for inspection of the steam generator tubes 3)-
ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of.the tubes in the event that there is i t. -
evidence of mechanical damage or progressive degradation due to design, Y )3-manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and.cause of any tube degradation so that ccrrective measures can be taken.
. The p.lant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligib1c corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained within -these limits, localized corrosion may likely result in stress corrosion cracking. The exter.t of cracking during plant operation would be limited by the listitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 500 gallons per day per steam generator).
Cracks' having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the. loads imposed during normal operation and by postulated accidents.
Operating plants have demon'strated that-primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.
Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.
s Wastage-type defects are unlikely with proper chemistry treatment 'of the'
-secondary coolant.
However,. even if a defect should develop in service, it-l will be found during scheduled inservice steam generator tube examinations.
Plugging will be required for all tubes with imperfections exceeding the 4
plugging limit of 40% of the tube nominal wall thickness.. Steam generator
- (.e M'
tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.
Whenever the results of any steam generator tubing inservice inspection
[.
fall into Category C-3,-these results will be promptly reported to the Commission pursuant to @ ::ificatic: 5.9.1 prior to resumption of plant opcration. Such w
d cases will be considered by the Commission on a case-by-case basis and may result in 'a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if
(.\\
necessary.
10 CFR 50.~11(Q 10)
SUPMER - UNIT 1 B 3/4 4-3
..-,_.-.--._.___,,.,m.
.-.___.,.m,
.m.
--,. - -, -,. -.. -. - _ ~ -, -.
ADMINISTRATIVE CONTROLS f.
Reports of violations of codes, regulations, orders, Technical Specificati.ons, or Operating License requirements having nuclear safety significance or reports of abnormal degradation of systems designed to contain radioactive material.
g.
Reports of significant operating abnormalities or deviations from normal and expected performance of plant equipment that affect nuclear safety, y"
s..
,,u....
- ss......
Etbd5$SN$55b. kNieS Of"E53INEYN$ $."NY$NI5 i.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures, systems, or components.
j.
The plant Security Plan and changes thereto.
k.
The Emergency Plan and changes thereto.
1.
Items which may constitute a potential nuclear safety hazard as identified during review of facility operations.
m.
Investigations or analyses of special subjects as requested by the Chairman of the Nuclear Safety Review Committee.
n.
The unexpected offsite release of radioactive material and the report as described in 5. 0.1.12(-).10 CF R 50.73 o.
Changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL.
AUTHORITY 6.5.1.7 The-Plant Safety Review Committee shall:
a.
Recommend in writing to the Director, Nuclear Plant Operations approval or disapproval of items considered under 6.5.1.6a, c, d, e, j,,and k above.
b.
Render determinations in writing to the Director, Nuclear Plant Operations with regard to whether or not each item considered under 6.5.1.6a,.c, and d above constitutes an unreviewed safety question.
c.
Make recommendations in writing to the Director, Nuclear Plant Operations that actions reviewed under 6.5.1.6(b) above did not constitute an unreviewed safety question.
I8 d.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President, Nuclear Operations and the Nuclear Safety Review Committee of dis-agreement between the PSRC and the Director, Nuclear Plant Operations however, the Director, Nuclear Plant Operations shall have responsi-bility for resolution of such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The Plant Safety Review Committee shall maintain written minutes of tach PSRC neeting that, at a minimum, cocument the results of all PSRC activ-ities performed under the responsibility and authority provisions of these technical specifications.
C) pies shall be provided to the Vice President Nuclear Operations and the Chairman of the Nuclear Safety Review Committee.
SUMMER - UNIT 1 6-7 Amendment No.18
i ADMINISTRATIVE CONTROLS l
MEETING FREQUENCY 6.5.2.5: The NSRC shall meet at least nnce per calendar quarter during the initial year of unit operation following fuel loading and at least once per six months thereafter.
000 RUM 6.5.2.6 A quorum of the NSRC necessary for the performance of the NSRC review and audit functions of these Technical Specifications shall consist of the Chairman or his designated alternate and at least 3 NSRC members including alternates.
No more than a minority of the quorum shall have line responsibility.
for operation of the unit.
REVIEW 6.5.2.7 The NSRC shall review:
c.
The safety evaluations for 1) changes to procedures, equipment or systems, and 2) tens or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
.4 Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
c.
Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59., 10 CFR.
d.
Proposed changes to Technical Specifications or this Operating
".[
License.
e.
Violations of codes, regulations, orders, Technical Specifications, license requirements, or internal proc 6dures or instructions having' nuclear safety significance.
f.
Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect ne: lear safety.
Dar.t; r;;;;r@:ng 2(, h:ur writt:n notific:tien to th:
Att REPORT L
E VE NT.S.
C=..i: icn.
g.
h.
All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.
i.
Reports and meetings minutes of the Plant Safety P.eview Committee.
1 SUMMER - UNIT 1 6-9
ADMINISTRATIVE CONTROLS
,I c.
Proposed tests and experiments which affect pisnt nuclear safety and are not addressed in the Final Safety Analysis Report shall be reviewed by an individual / group other than the individual / group which prepared the proposed test or experiment.
Eveds
(' )
d.
00: = cacu reportable pursuant to the Technical Specification 6.9
\\
and violations of Technical ~ Specifications ghall be investigated and a *eport prepared which evaluates the c E h ence and which provides recommendations to prevent recurrence.
Such report shall be approved by the Director, Nuclear Plant Operations and forwarded to the L
Chairm e of the Nuclear Safety Review Committee.
{, "'
e.
Individuals responsible for reviews performed in accordance with 6.5.3.1.a. 6.5.3.1.b, 6.5.3.1.c and 6.5.3.1.d shall be members of the plant staff tha': meet or exceed the qualification requirements of Section 4.4 of ANSI 18.1, 1971, as previously designated by the
_ jg Director, Nuclear Plant Operations.
Each such review shall include a determination of whether or not additional, cross-disciplinary, review is necessary.
If deemed necessary, such review shall be performed by the review personnel of the appropriate discipline.
f.
Each review will include a determination of whether or not an unreviewed safety question is involved.
RECORDS 6.5.3.2 Records of the above activities shall be provided'to the Director,
/g Nuclear Plant' Operations PSRC and/or NSRC as necessary for required reviews.
EVENY S.6 - REPORTABLE 066tfRRENSE ACTION EVE NTS 6.6.1 The following actiora shall be taken for REPORTABLE SECURRENCES:
TheCommissionshallbenotifiedanli/erareportsubmittedpursuant a.
f 5.,2;ificati;n C.^.hbon so,73 h tocrn M so, y totherequiremengs 4
b.
.Each REPORTABLE 0CCURRENCE. requiring 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> notification to the Oca.T.i :icn shall be reviewed by the PSRC and submitted to the NSRC f
and the Vice President, Nuclear Operations. k y, g g g,
[
6.7 SAFETY LIMIT VIOLATION rev'iew M be
[
6.7.1 The following actions shall be taken in the event a Safety Limit is h- )
L violated:
a.
The NRC Operations Center shall be notified by telephone as soon as L
possible nnd in all cases within one hour.
The Vice President,
--fg Nuclear Operations and the NSRC shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
A Safety Limit Violation Report shall be prepared.
The report shall h be reviewed by the PSRC.
This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation (\\
upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.
The Safety Limit Violation Report shall be submitted to the Commission,{,
c.
the NSRC and the Vice President, Nuclear Operations within 14 days l
i of the violation.
4 SUMMER - UNIT 1 6-12 Amendment No.18 l
ADMINISTRATIVE CCNTROLS 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS -
6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator Office of Inspection and Enforcement unless otherwise noted.
STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalatien testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and.(4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.
l SUMMER - UNIT 1 6-14a
ADMINISTRATIVE CONTROLS' 9
Type of container.(e.g., LSA, Type A, Type B, Large Quantity), and e.
- i f.
Solidification agent (e.g., cement, urea formaldehyde).
The-radioactive effluent release reports shall include unplanned releases from
-site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.
The radioactive effluent release reports shall include any changes to the Process Control' Program (PCP) made during the reporting period.
z MONTHLY OPERATING REPORT g'
-6.9.1.10 ' Routine reports of operating statistics and shutdown experience, including documentation-of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Director; Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C.
- 20555, with a copy to the Regional Of fice of Inspection and Enforcement, no lv.er than the 15th of each month following the calendar month covered by the report.
Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the i Monthly Operating Report within 90 days in which the change (s) was made effective.
In. addition, a report of any major changes to the radioactive waste treatment
. systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted as set forth in 6.5 above.
' REP 6MARLLOCCURRENCES 6.9.1.11~ The REPO ENCES of Specificati ow, including corrective actions and m
-- ta
~
% ws recurrence, shall be reported to the NRC.
Supplemental r e requ re describe final resolu-tion of occu'r case of corrected or supplemental repo ensee event a
be completed and reference shall be made to the original repor PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP
- 6.9.1.
-The-types of events listed shall be reported within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> b telephone confirmed by. telegraph, mailgram, or facsimile transai on to the Regional Admin rator of the Regional Office, or his designate, later than the first working following the event, with a written foll up report within 14 days.
The written llowup report shall include, as a imum, a completed ccoy of a licensee event-ort form.
Information prov ed'on the licensee event report form'shall be-suppleme d, as needed, by ad ional narrative material to provide complete explanation of-circumstanc surrounding the event.
-a.
Failure'of the reactor prote ystem'or other systems subject to Limiting Safety System Setti initiate the required protective function by the time a mo ored par ter reaches the setpoint specified as the Limit Safety System ting in the Technical Specifications or ure to complete the re red protective function.
~
b.
Operation of unit or affected systems when an rameter or operation jv.t to a Limiting Condition for Operati is less conser ive than the least conservative aspect of the L ing Co ion for Operation established in the Technical Specific ons.
^
c.
bnormal degradation discovered-in fuel cladding, recctor coolant pressure boundary, or primary containment.
SUlWER r UNIT 1 6-18 3
r
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.w,-v w w p v,+--.--g,.-,#--,-7
ADMINISTRATIVE CONTROLS d.
Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1% delta k/k; a calculated reac vity balance indicating a SHUTDOWN MARGIN less conservative than spec ied the Technical Specifications; short-term reactivity increas th t correspond to a reactor period of less than 5 seconds or if sub itical, an unplanned reactivity insertion of more than.5%
delta k; or occurrence of any unplanned criticality.
e.
Failure o malfunction of one or more components whic prevents or could preve t, by itself, the fulfillment of the fu tional require-ments of sys m(s) used to cnpe with accidents an yzed in the SAR.
f.
Personnel error procedural inadequacy whic revents or could prevent, by itsel, the fulfillment of the f ctional requirements of systems required cope with accidents nalyzed in the SAR.
g.
Conditions arising from atural or man-ade events that, as a direct result of the event reoui unit shu own, operation of safety systems, or other protective mecoure requir d by technical specifications.
h.
Errors discovered in the trans1 t or accident analyses or in the methods used for such analyse a described in the safety analysis report or in the bases for e tec ical specifications that have or could have permitted reac r operati in a manner less conservative than' assumed in the ana ses.
i.
Performance of stru ures, systems, or co onents that requires remedial action or correcti e measures to prevent op ation in a manner less con-servative than sumed in the accident analys in the safety analysis report or tech cal specifications bases; or di overy during unit life of condition not specifically considered in the afety analysis report or technic specifications that require remedial tion or corrective measures o prevent the existence of development of unsafe condition.
j.
Offs e releases of radioactive materials in liquid and seous eff uents which exceed the limits of Specification 3.11.1.
or 3.11.2.1.
k.
xceeding the limits in Specification 3.11.1.4 or 3.11.2.6 fo the i
storage of radioactive materials in the listed tanks.
The writ n l
follow-up report shall include a schedule and a description of ac '-
vities planned and/or taken to reduce the contents to within the specified limits.
N Y WRITTEN REPORTS 6.9.1.13 The type ts listed below shall be the of written reports to the Regional Administrator Regional within 30 days of occurrence of the event.
The written report as a minimum, a completed copy of a licensee event repor nformation provide licensee event report form shall emented, as needed, by additional narrative al to omplete explanation of the circumstances surrounding the event.
SUMER - UNIT 1 6-19
ADMINISTRATIVE CONTROLS a.
Reactor protection system or engineered safety feature instrume settings which are found to be less conservative than those e ab-lished by the Technical Specifications but which do not pr ent the ulfillment of the functional requirements of affected s tems.
b.
Con tions leading to operation in a degraded mode p itted by a Limit Condition for Operation or plant shutdown equired by a Limiting ondition for Operation.
c.
Observed in quacies in the implementation administrative or procedural con ois which threaten to cau reduction of degree of redundancy provi in reactor protecti systems or engineered safety feature syst s.
d.
Abnormal degradation o stems o er than those specified in 6.9.1.12.c above designed o co ain radioactive material resulting from the fission process.
e.
An unplanned offsite rel e of more than 1 curie of radioactive material in liquid eff ents, 2) e than 150 curies of noble gas in gaseous effluents or 3) more tha
.05 curies of radiciodine in gaseous effluents.
The report of an un anned offsite release of radioactive mat ial shall include the fo owing information:
1.
A descr' tion of the event and equipment 'nvolved.
2.
Ca s(s) for the unplanned release.
3.
ctions taken to prevent recurrence.
Consequences of the unplanned release.
/f.
Measured levels of radioactivity in an environmental samplin medium determined to exceed the reporting level values of Table 3.12-hen averaged over any calendar quarter sam 41ng period.
1 R ADI AL P EMIN G FArrnR Lm rr R f'90 R g
- 6. 9.1.M The F limit for Rated Thermal Power (Fxy ) shall be provided to xy the Regional Administrator of the Regional Office of Inspection and Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Core Performence Branch, U. S. Nuclear Regulatory Commission, Washington, D.C.
20555 for all core planes containing bank "0" control rods and all unrodded core planes at least 60 days prior to cycle initial criticality.
In the event that the limit would be submitted at scme other time during core life, it shall be submitted 60 days prior to the date the limit would become effective unless otherwise exempted by the Commission.
R Any information needed to support F will be by request from the NRC and need not be included in this report.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Office of Inspection and Enforcement Regional Office within the time period specified-for each report.
SUMMER - UNIT 1 20 6-h
t o~
ADMINISTRATIVE CONTROLS 6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, ;he 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:
a.
Records and logs of unit operation covering time interval at each power level.
b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. gygg73 c.
All REPORTABLE SCWRRENGE6 submitted to the Commission.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
e.
Records of changes made to the procedures required by Specification 6.8.1.
f.
Records of radioactive shipments.
g.
Records of sealed source and fission detector leak tests and results.
h.
Records of annual physical inventory of all sealed source material of record.
6.10.2 -The following records shall be retained for the duration of the Unit Operating License:
Records and drawing changes reflecting unit design modifications d.
made to systems and equipment described in the Final Safety Analysis Report.
h.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
c.
Records of radiation exposure for all individuals entering radiation centrol areas.
d.
Records of gaseous and liquid radioactive material released to the environs.
I e.
Records of transient or operational cycles for those unit components identified in Table 5.7-1.
f.
Records of reactor tests and experiments.
g.
Records of training and qualification for current members of the unit staff.
l h.
Records of in-service inspections performed pursuant to these Technical Specifications.
SUMER - UNIT 1 6-M M
'