ML20112C732
| ML20112C732 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 03/15/1985 |
| From: | SOUTH CAROLINA ELECTRIC & GAS CO. |
| To: | |
| Shared Package | |
| ML20112C722 | List: |
| References | |
| NUDOCS 8503220105 | |
| Download: ML20112C732 (10) | |
Text
ATTACHMENT 1 ADMINISTRATINECONTROLS Type of container (e.g., LSA, Type A, Type B, Large Quantity), and e.
Solidification agent (e.g., cement, urea formaldehyde).
f.
. 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 f
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Process Control Program (PCP) made during the reporting period.
MONTHLY OPERATING REPORT Routine reports of operating statistics and shutdown experience, 6.9.1.10 ibourc.e [
including documentation of all challenges to the PORV's or safety valves,
..,J Mgra;;; "n:1ym, U.S. Nuclear Regulatory Commission, Was
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shall be submitted on a monthly basis to the Director, Office oFManagement witti a copy to the Regional Office of Inspection and Enforcement, no later 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 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.
RADIAL PEAKING FACTOR LIMIT REPORT The F limit for Rated Thermal Power (FRTP) shall be provided to 6.9.1.11 x
the Regional Administrator of the Regional Office of Inspection and Enforcement xy with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Core Performance Branch, U. S. Nuclear Regulatory Commission, Washington, D.C.
20555 for all core planes containing bank "D" control rods and all unrodded core In the event that planes at least 60 days prior to cycle initial criticality.
the limit would be submitted at some other time during core life, it shall be submitted 60 days prior to the date the limit would become effective unless i
otherwise exempted by the Commission.
i RTP Any information needed to support F will be by request from the NRC and x
need not be included in this report.
SPECIAL REPORTS
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Special reports shall be submitted to the Regional Administrator of the 6.9.2 Office of Inspection and Enforcement Regional Office within the time period specified for each report.
6.10 RECORD RETENTION
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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.
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8503220105 850315 PDR ADOCK 05000395 PDR P
"n-i.:nt M:. 35 SUWiER - UNIT 1 6-18
ATTACHMENT 2 1
ADMINISTRATIVE CONTROLS
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RECORDS Records of NSRC activities shall be prepared, approved and distributed 6.5.2.10 l
as indicated below:
Minutes of each NSRC meeting shall be prepared, approved and forwarded a.
to the Vice President, Nuclear Operations within 14 days following l
each meeting.
Reports of reviews encompassed by Section 6.5.2.7 above, shall be b.
prepared, approved and forwarded to the Vice President, Nuclear l
Operations within 14 days following completion of the review.
l Audit summary reports encompassed by Section 6.5.2.8 above, shall be I
c.
forwarded to the NSRC and to the Vice Presidea.t, Nuclear Operations.
Full audits shall be forwarded to the management positions responsibli for the areas audited within 30 days after completion of the audit by the auditing organization, j
6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES i
Activities which affect nuclear safety shall be conducted as follows:
6.5.3.1 Procedures required by Technical Specification 6.8 and other procedures
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which affect plant nuclear safety, and changes thereto, shall be a.
prepared, reviewed and approved.
Each such procedure or procedure change shall be reviewed by an individual / group other.than the individual / group which prepared the procedure or procedure change, but who may be from the same organization as the individual / group Procedures other j
which prepared the procedure or procedure :hange.
than Administrative Procedures will be approved as delineated in writing by the Director, Nuclear Plant Operations.
The Director, Nuclear Plant Operations, will approve administrative procedures, security implementing procedures and emergency plan implementing Temporary approval to procedures which clearly do not procedures.
change the intent of the approved procedures can be made by two members of the plant management staff, at least one of whom holds a k
Senior Reactor Operator's License.
For changes to procedures which i
may involve a change in intent of the approved procedures, the person authorized above to approve the procedure shall approve the change.
Proposed changes or modifications to plant nuclear safety-related b.
structures, systems and components shall be reviewd as designated by
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the Director. Nuclear Plant Operations.
Each such modification
"'d shall be designed as authorized by M$clect Cn;;in;; ring and shall be
'N" reviewed by an individual / group other than the individual / group which designed the modification, but who may be from the same organi-zation as the individual / group which designed the modifications.
l Implementation of modifications to plant nuclear safety-related
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structures, systems and components shall be concurred in by the Director, Nuclear Plant Operations.
6-11
- f. ;nt,;nt Mc.10 SUMMER - UNIT 1
PAGE 1 OF 3 ATTACHMENT 3 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION..........................................
3/4 9-2 3/4.9.3 DECAY TIME...............................................
3/4 9-3 3/4.9.4 REACTOR BUILDING PENETRATIONS............................
3/4 9-4 3/4.9.5 COMMUNICATIONS...........................................
3/4 9-5 3/4.9.6 MANIPULATOR CRANE........................................
3/4 9-6 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Level.........................................
3/4 9-7 Low Water Level..........................................
3/4 9-8 3/4.9.8 REACTOR BUILDING PURGE AND EXHAUST ISOLATION SYSTEM......
3/4 9-9 3/4.9.9 WATER LEVEL - REFUELING CAVITY AND FUEL TRANSFER CANAL...
3/4 9-10 3/4.9.10 WATER LEVEL - SPENT FUEL P00L............................
3/4 9-11 3/4.9.11 SPENT FUEL PDOL VENTILATION SYSTEM.......................
3/4 9-12 3/4.9. n S? ENT FUEL ASS EMBtN S TO R A G E..........
3/4.10 SPECIAL TEST EXCEPTIONS 3/4 9-19 j
3/4.10.1 SHUTDOWN MARGIN..........................................
3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS....
3/4 10-2 3/4.10.3 PHYSICS TESTS............................................
3/4 10-3 3/4.10.4
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REACTOR COOLANT L00PS..................'..............._...
3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN....................
3/4 10-5 SUMMER-UNIT 1 X
e m
PAGE 2 OF 3 INDEX BASES SECTION PAGE 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION......................................
B 3/4 9-1 3/4.9.2 INSTRUMENTATION..........................................
B 3/4 9-1 3/4.9.3 DECAY TIME...............................................
B 3/4 9-1 3/4.9.4 REACTOR BUILDING PENETRATIONS............................
B 3/4 9-1 3/4.9.5 C0 MUNICATIONS...........................................
B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE........................................
B 3/4 9-1 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION............
B 3/4 9-2 3/4.9.8 REACTOR BUILDING PURGE SUPPLY AND EXHAUST ISOLATION SYSTEM.................................................
B 3/4 9-2 3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and
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S P ENT F U E L P00 L.......................................... B 3/4 9-2 3/4.9.11 SPENT FUEL POOL VENTILATION SYSTEM....................... B 3/4 9-2 B 3/4 9 -3 3/9.9.n SPENT F0EL ASS EMBLY.5TO R A G E....
3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN..........................................
B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS....
B 3/4 10-1 3/4.10.3 PHYSICS TESTS............................................
B 3/4 10-1 3/4.10.4 REACTOR COO LANT L00PS.................................... B 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN....................
B 3/4 10-1 g
SUMER-UNIT 1 XV p
PAGE 3 OF 3 INDEX OESIGN FEATURES f
PAGE SECTION f
5.1 SITE 5-1 5.1.1 EXCLUSION AREA..............................................
5-1 5.1.2 LOW POPULATION 10NE.........................................
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5.1.3 SITE BOUNDARY FOR GASEOUS EFFLUENTS.........................
5-1 5.1.4 SITE BOUNDARY FOR LIQUID EFFUENTS...........................
5-1 5.2 REACTOR BUILDING 5-1 5.2.1 CONFIGURATION...............................................
5.2.2 DESIGN PRESSURE AND TEMPERATURE.............................
5-1 5.3 REACTOR CORE 5-6 5.3.1 FUEL ASSEM8 LIES.............................................
- y..
A 5-6 5.3.2 CONTROL ROD ASSEMBLIES......................................
5.4 REACTOR COOLANT SYSTEM 5-6 5.4.1 DESIGN PRESSURE AND TEMPERATURE.............................
5-6 5.4.2 V0LUME......................................................
5-6 5.5 METEOROLOGICAL TOWER LOCATION.................................
- 5. 6 FUEL STORAGE 5-7 5.6.1 CRITICALITY.................................................
5-7 5.6.2 0RAINAGE....................................................
5.6.3 CAPACIT)................................................:...
5-7 (a) l 5-7 ba) h 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT...........................
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SUMMER-UNIT 1 XVII L
TABLE 3.3-3 52 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION C5 MINIMUM TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.
SAFETY INJECTION, REACTOR TRIP, FEEDWATER ISOLATION, a
CONTROL ROOM ISOLATION, START DIESEL GENERATORS, CONTAINMENT Q
COOLING FANS AND ESSENTIAL 2:
Ee a.
Manual Initiation 2
1 2
1,2,3,4 18 R
b.
Automatic Actuation 2
1 2
1,2,3,4 14 Logic and Actuatien y
Relays g
a, c.
Reactor Building 3
2 2
1,2,3 15*
~-
Pressure-High-l o
d.
Pressurizer 3
2 2
1, 2, 3#
15*
Pressure - Low e.
Differential 3/ steam line 2/ steam line 2/ steam line' 1, 2, 3 15*
Pressure Between twice and 1/3 Steam Lines - High steam lines O
s s
s
^
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.^
E TABLE 4.3-2 E
E ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
- l Z
TRIP e
ANALOG ACTUATING MODES FOR CHANNEL DEVICE MASTER SLAVE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS RFOUIRED-1.
SAFETY INJECTION, REACTOR TRIP FEEDWATER ISOLATION, CONTROL ROOM ISOLATION START DIESEL GENERATORS, CONTAINMENT COOLING FANS AND ESSENTIAL SERVICE WATER a.
Manual Initiation N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
1, 2, 3, 4
- pe M b.
Automatic Actuation N.A.
N.A.
N.A.
N.A.
M(1)
M(1)
Q 1, 2, 3, 4 8
Logic and Actuation Relays w
- x:
O c.
Reactor Build Sg S
R M
N.A.
N.A N.A.
N.A.
1, 2, 3 x
Pressure-High-l E
1 d.
Pressurizer Pressure--Low 5 R
M N.A N.A.
N.A.
N.A.
1, 2, 3 e.
Differential Pressure S
R M
N.A.
N.A.
N.A.
N.A.
1, 2, 3 Between Steam Lines--High f.
Steam Line Pressure Low 5
R H
N.A.
N.A.
N.A.
N.A.
1, 2, 3 2.
REACTOR BUILDING SPRAY a.
Manual Initiation N.A.
N.A.
N.A.
R N.A.
N.A.
N.A.
1, 2, 3, 4 f
b.
Automatic Actuation N.A.
N.A.
N.A.
N.A.
M(1)
M(1)
Q 1,2,3,4 i
Logic and Actuation N
{
Relays o
c.
Reactor Building S
R M
N.A.
N.A.
N.A.
N.A.
1,2,3 Pressure "?;i "!;.-
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ATTACHMENT 5 CONTAINMENT SYSTEMS
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BASES 3/4.6.1.4 INTERNAL PRESSURE The limitations on reactor building internal pressure ensure that 1) the reactor building structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of S59 psig and 3,5
- 2) the reactor building peak pressure does not exceed the design pressure of 57 psig during steam line break conditions.
The maximum peak pressure expected to be obtained from a steam line break event is 47.1 psig. The limit of 1.5 psig for initial positiva containment pressure will limit the total pressure to 47.1 psig which is,less than design pressure and is consistent with the accident analyses.
3/4.6.1.5 AIR TEMPERATURE The limitations on reactor building average air temperature ensure that the overall containment average air temperature does not exceed the initial temperature condition assumed in the accident analysis for a steam line break accident.
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3/4.6.1.6 REACTOR BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility.
Structural integrity is required to ensure that the containment will withstand the maximum pressure of 47.1 psig in the event of a steam line break accident. The measurement of containment tendon lift off force, the tensile tests of the tendon wires, the visual ~ examination of tendons, anchorages and exposed interior and exterior surfaces of the containment, and the Type A leakage test are sufficient to demonstrate this capability.
The tendon lift off forces are evaluated to ensure that 1) the rate of tendon force loss is within predicted limits, and 2) a minimum required prestress level exists in the containment.
In order to assess the rate of i
force loss, the lift off force for a tenden is compared with the force predicted for the tendon times a reduction factor of 0.95.
This resulting force is l
referred to as the 955 Base Value.
The predicted tendon force is equal to the i
original stressing force minus losses due to elastic shortening of the tendon, stress relaxation of the tendon wires, and creep and shrinkage of the concrete.
The SX reduction on the predicted force is intended to compensate for both uncertainties in the prediction techniques for the losses and for inaccuracies l
in the lift-off force measurements.
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i SUMMER - UNIT 1 B 3/4 6-2 i
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3" TABLE 3.3-6 E
RADIATION MITORING INSTRLSENTATION MINIftlM e
E CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT ACTION
]
INSTPUMENT OPERABLE MDOES SETPOINT RANGE 1
1.
AREA MDNITORS l
a.
Spent Fuel Pool Area (RM-G8)
,j 4
10 mR/hr 25 1 15 mR/hr 10 1
b.
Reactor Building Manipulator 5
Crane Area (RM-G17A-or 1
6 1 1 R/hr 1 - 10 ar/hr 28 RM-G178) c.
Reactor Building Area j
f.
High Range RM-G7 and 2
1, 2, 3 & 4 N/A 10 - 107 R/hr 30 High Range RM-G18 1 - 107 R/hr 5d 3:
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2.
PROCESS MDNITORS g
8
't' a.
Spent Fuel Pool Exhaust -
C Ventilation System (RM-A6)
-5 6
1.
Gaseous Activity I
t 1 x 10 pCi/cc 10 - 10 cpm 27 TKr-85) 6 N/A 10 - 10 cpm 27 11.
Particulate Activity 1
b.
Containment 1.
Gaseous Activity
- Purge & Exhaust 6
Is lation (RM-A4) 1 6
5 2 x background *** 10 - 10 cpm 28 ond G<oece 11.
Particulate
- Activity (RM-A2) - RCS Leakage 6
Detection 1
1, 2, 3 & 4 N/A 10 - 10 cpm 26 6
c.
Control Room Isolation 1
ALL NODES 5 2 x background 10 - 10 ce 29 (RM-A1)
- With fuel in the storage pool or building
'p O
- With irradiated fuel in the storage pool c) tr2
- Alarm / trip setpoint will be per the Operational Dose Calculation Manual when purge exhaust operations H
are in progress O
J J
w
v
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E TABLE 4.3-3 E
RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIRENENTS ANALOG CHANNEL MODES FOR WHICH E
CHANNEL CHANNEL OPERATIONAL SURVEILLANCE
-i INSTRUMENT CHECK CALIBRATION TEST IS REQUIRED 1.
AREA MONITORS a.
Spent Fuel Pool Area (RM-G8)
S R
M b.
Reactor Building Manipulator S
R H
6 Crane Area (RM-G17A or RM-G178) c.
Reactor Building Area 1.
High Range (RM-G7)
S R***
M 1, 2, 3 & 4 11.
High Range (RM-G18)
S R***
M 1, 2, 3 & 4 2.
PROCESS MONITORS a.
Spent Fuel Pool Exhaust Area -
R Ventilation System (RM-A6) 1.
Gaseous Activity S
R M
Y 11.
Particulate Activity S
R H
g b.
Containment 1.
Gaseous Activity
- Purge & Exhaust an d G xcy Isolation (RM-A4)
S R
M 6
11.
Particulate # Activity
- RCS Leakage Detection S
R H
1, 2, 3 & 4 (RM-A2) c.
Control Room Isolation (RM-A1) S R
H All MODES d.
Noble Gas Effluent Monitors j
(High Range) j
- 1. Main Plant Vent (RM-A13)
S R
M 1, 2, 3 & 4
- 11. Main Steam Lines (RM-G19A, B, C)
S R
H 1, 2, 3 & 4 111. Reactor Building Purge Supply & Exhaust
,a System (RM-A14)
S R
H 1, 2, 3 & 4 g
in
- With fuel in the storage pool or building g
- With irradiated fuel in the storage pool
- Channel Calibration will consist of an electronic calibration of the channel, not including the detector, @
for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.
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