Text

Amend 49 to License NPF-12,modifying Tech Specs to Reflect Administrative Changes
	ML20211A654
Person / Time
Site:	Summer
Issue date:	05/16/1986
From:	Rubenstein L; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20211A660	List: ML20211A654; ML20211A661;
References
	NUDOCS 8606110272
	Download: ML20211A654 (20)
	v • d • e

o UNITED STATES

[

]i NUCLEAR REGULATORY COMMISSION g

j WASHINGTON, D. C. 20555

\\...../

SOUTH CAROLINA ELECTRIC & GAS COMPANY SOUTH CAROLINA PUBLIC SERVICE AUTHORITY DOCKET NO. 50-395 VIRGIL C. SUMMER NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 49 License No. NPF-12 I.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by the South Carolina Electric & Gas Company acting)for itself and South Carolina Public Service Authority (the licensees, dated March 15, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; I

l B.

The facility will operate in conformity with the application, as amended, i

the provisions of the Act, and the regulations of the Comission; l

l C.

There is reasonable assurance:

(i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; l

D.

The issuance of this license amendment will not be inimical to the I

comon defense and security or to the health and safety of the pubifc; E.

The issuance of this license amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is hereby amended by changes to the Technical l

Specifications as indicated in the attachments to this license amendment and paragraph 2.C(2) of Facility Operating License No. NPF-12 is hereby l

amended to read as follows:

(2) Technical Specifications I-l The Technical Specifications contained in Appendix A, as revised through Amendment No. 49, are hereby incorporated into this license.

South Carolina Electric & Gas Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

8606110272 860519 l

hDR ADOCK 05000395 l

PDR l

. 3.

This license amendment is effective seven days after its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

. _ - '. kh k@ PWR Project Directorate #2 lester S. Rubenstein, Director Division of PWR Licensing-A Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: May 16,1986 I

t l

i

ATTACHMENT TO LICENSE AMENDMENT NO. 49 FACILITY OPERATING LICENSE NO. NPF-12 DOCKET NO. 50-395 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding over-leaf page is also provided to maintain document completeness.

Remove Pages Insert Pages X

X XV XV XVII XVII 3/4 3-16 3/4 3-16 3/4 3-35 3/4 3-35 3/4 3-42 3/4 3-42 3/4 3-45 3/4 3-45 8 3/4 6-2 8 3/4 6-2 6-11 6-11 6-18 6-18 l

l

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

+

PAGE SECTION i

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 Leve1..........................................

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 WAT E R LEV E L - S P ENT FU E L P00 L............................

3/4 9-11 3/4.9.11 SPENT FUEL POOL VENTILATION SYSTEM.......................

3/4 9-12 3/4.9.12 SPENT FUEL ASSEMBLY ST0 RAGE..............................

3/4 9-14 l

3/4.10 SPECIAL TEST EXCEPTIONS 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 REACTOR COOLANT L00PS....................................

3/4 10-4 3/4.10.5 POSITION INDICATION SYSTEM - SHUTD0WN....................

3/4 10-5 SUMMER - UNIT 1 X

AMENDMENT N0. 49

INDEX BASES

[-

PAGE SECTION 3/4.9 REFUELING OPERATIONS B 3/4 9-1 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 COMMUNICATIONS.........................................

B 3/4 9-1 3/4.9.6 MANIPULATOR CRANE......................................

B 3/4 9-2 3/4.9.7 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION..........

REACTOR BUILDING PURGE SUPPLY AND EXHAUST ISOLATION i

3/4.9.8 B 3/4 9-2 SYSTEM...............................................

3/4.9.9 and 3/4.9.10 WATER LEVEL - REACTOR VESSEL and B 3/4 9-2 SPENT FUEL P00L........................................

B 3/4 9-2 3/4.9.11 SPENT FUEL POOL VENTILATION SYSTEM.....................

B 3/4 9-3 l

3/4.9.12 SPENT FUEL ASSEMBLY ST0 RAGE............................

3/4.10 SPECIAL TEST EXCEPTIONS B 3/4 10-1 3/4.10.1 SHUTDOWN MARGIN........................................

B 3/4 10-1 l

3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS..

B 3/4 10-1 3/4.10.3 P HY S I C S T EST S..........................................

8 3/4 10-1 3/4.10.4 REACTOR COOLANT L00PS..................................

8 3/4 10-1 3/4.10.5 POSITION INDICATION SYSTEM - SHUT 00WN..................

~

l l

i XV AMENDMENT NO. 49 SUMMER - UNIT 1

INDEX BASES SECTION PAGE 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS.........................................

B 3/4 11-1 3/4.11.2 GASEOUS EFFLUENTS........................................

B 3/4 11-3 3/4.11.3 SOLID RADI0 ACTIVE WASTE..................................

B 3/4 11-6 3/4.11.4 TOTAL 00SE...............................................

B 3/4 11-6 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM.......................................

B 3/4 12-1 3/4.12.2 LANo uSE CENSUS..........................................

B 3/4 12-1 3/4.12.3 INTERLABORATORY COMPARISON P.90 GRAM.......................

B 3/4 12-2 SUW1ER-UNIT 1 XVI

INDEX DESIGN FEATURES SECTION PAGE 8

5.1 SITE 5-1 5.1.1 EXCLUSION AREA.................................................

5-1 5.1.2 LOW POPULATION Z0NE............................................

5.1.3 SITE BOUNDARY FOR GASEOUS EFFLUENTS............................

5-1 5.1.4 SITE BOUNDARY FOR LIQUID EFFLUENTS.............................

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 ASSEMBLIES................................................

5-6 5.3.2 CONTROL R00 ASSEMBLIES.........................................

5.4 REACTOR COOLANT SYSTEM 5-6 5.4.1 DESIGN PRESSURE AND TEMPERATURE................................

5-6 5.4.2 V0LUME.........................................................

5-6 1

5.5 METEOROLOGICAL TOWER L0 CATION....................................

l l

5.6 FUEL STORAGE 5-7 5.6.1 CRITICALITY....................................................

5-7 5.6.2 DRAINAGE.......................................................

l 5-7(a) l 5.6.3 CAPACITY.......................................................

t 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT..............................

5-7(a) l AMENDMENT NO. 49 SUMMER - UNIT 1 XVII

INDEX

.s ADMINISTRATIVE CONTROLS SECTION PAGE s

i.

-k 6.1 RESPONSIBILITY....................................

4 6-1

s 6.2 ORGANIZATION

^

6.2.1 0FFSITE.....................................................

6-1

~

6.2.2 UNIT STAFF.................................................

15 - 1

~

6.2.3 INDEPENDENT SAFETY ENGINEERING GR00P..'......................

6-5 s.

6.2.4 SHIFT TECHNICAL ADVIS0R............'.........................

6-5 6.3 UNIT STAFF QUALIFICATIONS.....................................

'6-5

' ' ~ 6-5 6.4 TRAINING..................................,

6.5 REVIEW AND AUDIT s

6.5.1 PLANT SAFETY REVIEW COMMITTEE Function.................................................

6-6 Composition.................................................

6-6 Alternates..................................................

6-6 Meeting Frequency...........................................

6-6 Quorum......................................................

6-6 Responsibilities............................................

6-6 Authority...................................................

6-7 Records..........................

6-7 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE Function....................................................

6-8 Composition.................................................

6-8 Alternates..................................................

6-8 Consultants.................................................

6-8 Meeting Frequency...........................................

6-9 Quorum......................................................

6-9 SUMMER-UNIT 1 XVIII a

1

~

3/4.3 INSTRUMENTATION j

SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic P

actuation logic and relays shall be demonstrated OPERABLE by performance of the engineered safety feature actuation system 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 demons'trated.to be within the limit at least once per 18 months.*

I Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 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.

A one time extension of the frequency of response time tests is granted until

)

June 30,1983 for all tests due to be completed before this date.

Surveillance tests for response time will be conducted on or before June 30, 1983.

5 l

l SUMMER - UNIT 1 3/4 3-15a Amendment No. 13

l

.g TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 9

MINIMUM E

TOTAL N0.

CHANNELS CHANNELS APPLICABLE Z

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w

1.

SAFETY INJECTION, REACTOR TRIP, FEEDWATER ISOLATION, CONTROL ROOM ISOLATION, START DIESEL GENERATORS, CONTAINMENT l

COOLING FANS AND ESSENTIAL l

SERVICE WATER.

a.

Manual Initiation 2

1 2

1,2,3,4 18 b.

Automatic Actuation 2

1 2

1,2,3,4 14 Logic and Actuation y

Relays s

y c.

Reactor Building 3

2 2

1,2,3 15*

l g

Pressure - High-1 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 g

Steam Lines - High steam lines E

a m

g 8

f TABLE 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TRIP E

ANALOG ACTUATING MODES FOR 2

CHANNEL DEVICE MASTER SLAVE WHICH 9

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE e FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED E 1.

SAFETY INJECTION, REACTOR TRIP G

FEEDWATER ISOLATION, CONTROL ROOM ISOLATION START DIESEL GENERATORS, CONTAINMENT COOLING FANS AND ESSENTIAL SERVICE WATER a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4 b.

Automatic Actuation N.A.

N.A.

H.A.

N.A.

M(1)

Q 1,2,3,4 Logic and Actuation Relays c.

Reactor Building S

R H

N.A.

N.A N.A.

N.A.

1,2,3

[

g Pressure-High-1

[

d.

Pressurizer Pressure--Low S R

M N.A N.A.

N.A.

1,-2, 3 En e.

Differential Pressure 5

R M

N.A.

1,2,3 Between Steam Lines--High f.

Steam Line Pressure Low S

R M

N.A.

1,2,3 2.

REACTOR BUILDING SPRAY a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1,2,3,4 k

b.

Automatic Actuation N.A.

N.A.

M(1)

Q 1,2,3,4 5

Logic and Actuation y

Relays

+

c.

Reactor Building S

R M

N.A.

1, 2, 3

,5 Pressure-3

I

' TABLE 4.3-2 (Continued)

E ENGINEEREDSAFETYFEATUREACTUATIONSYSTkNINSTRtMENTATION SURVEILLANCE REQUIREMENTS e

EQ TRIP w

ANALOG ACTUATING NODES FOR CHANNEL DEVICE NASTER SLAVE

%dHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TESY TEST TEST IS REQUIRED

3. CONTAINNENT ISOLATION a.

Phase "A" Isolation

1) Manual N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4

2) Safety Injection See 1 above for all Safety Injection Surveillance Requirements

3) Automatic Actuation N.A.

M.A.

N.A.

M(1)

Q 1,2,3,4 R

Logic and Actuation

[

Relays g b.

Phase "B" Isolation

1) Automatic Actuation N.A.

M.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation Relays

2) Reactor Building S

R M

N.A.

M.A.

N.A.

1, 2, 3 Pressure--High-High-High c.

Purge and Exhaust Isolation

1) Automatic Actuation M.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation Relays

2) Containment Radio-S R

M N.A.

N.A.

W.'d7 1,2,3,4 activity-High

3) Safety Injection See 1 above for all Safety Injection Surveillance Requirements.

1

_ INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION t

s LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm / trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

With a radiation monitoring channel alarm / trip setpoint exceeding a.

the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

I" b.

With one or more radiation monitoring channels inoperable, take the i --

ACTION shown in Table 3.3-6.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

c.

SURVEILLANCE REQUIREMENTS l

i 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-3.

l i

i SUMMER - UNIT 1 3/4 3-41

~

'~ '.

l

,y

=

s.

~.

I 1

g TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT E

INSTRUMENT OPERABLE MODES SETPOINT RANGE ACTION

[

1.

AREA MONITORS Spent Fuel Pool Area (RM-G8)

_j 4

a.

1 5 15 mR/hr 10

- 10 mR/hr 25 b.

Reactor Building Manipulator 5

Crane Area (RM-G17A or 1

6 1 1 R/hr 1 - 10 mr/hr 28 RM-G178) c.

Reactor Building Area

i. High Range RM-G7 and 2

1, 2, 3 & 4 N/A 10 - 107 R/hr 30 1 - 107 R/hr High Range RM-G18 2.

PROCESS MONITORS a.

Spent Fuel Pool Exhaust -

E Ventilaticn System (RM-A6)

-5 6

< 1 x 10 pCi/cc 10 - 10 cpm 27

[

i. Gaseous Activity 1

{Kr-85) 6 N/A 10 - 10 cpm 27 ii.

Particulate Activity 1

b.

Containment i.

Gaseous Activity

- Purge & Exhaust 6

Isolation (RM-A4) 1 6

5 2 x background *** 10 - 10 cpm 28 g

9 ii.

Particulate and Gaseous E

Activity (RM-A2) - RC3 6

g Leakage Detection 1

1, 2, 3 & 4 N/A 10 - 10 cpm 26 6

29 c.

Control Room Isolation 1

ALL MODES 5 2 x background 10 - 10 cpm

~+

2 (RM-A1)

With fuel in the storage pool or building

- With irradiated fuel in the storage pool

- - Alarm / trip setpoint. will be per the Operational Dose Calculation Manual when purge exhaust operations are in progress 9

8 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 9

ANALOG CHANNEL MODES FOR WHICH CHANNEL CHANNEL OPERATIONAL SURVEILLANCE E

INSTRUMENT CHECK CALIBRATION TEST IS REQUIRED

-4 w

1.

AREA MONITORS a.

Spent Fuel Pool Area (RM-G8)

S R

M b.

Reactor Building Manipulator S

R M

6 Crane Area (RM-G17A or RM-G17B) c.

Reactor Building Area 1.

High Range (RM-G7)

S R***

M 1, 2, 3 & 4 ii.

High Range (RM-G18)

S R***

M 1, 2, 3 & 4 2.

PROCESS MONITORS Spent Fuel Pool Exhaust Area -

a.

Ventilation System (RM-A6) 1.

Gaseous Activity S

R M

ii.

Particulate Activity S

R M

{

b.

Containment i.

Gaseous Activity T

- Purge & Exhaust Isolation (RM-A4)

S R

M 6

l ii.

Particulate and Gaseous l

Activity - RCS Leakage S

R M

1, 2, 3 & 4 Detection (RM-A2) c.

Control Room Isolation (RM-A1) S R

M All MODES d.

Noble Gas Effluent Monitors g

(High Range) m5

i. Main Plant Vent (RM-A13)

S R

M 1, 2, 3 & 4 5

ii. Main Steam Lines "4

(RM-G19A, B, C)

S R

H 1, 2, 3 & 4 iii. Reactor Building Purge 2P Supply & Exhaust System (RM-A14)

S R

M 1, 2, 3 & 4 "With fuel in the storage pool or building

- 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 gama source.

INSTRUMENTATION MOVABLE INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The movable incore detection system shall be OPERABLE with:

a.

At least 75% of the detector thimbles, b.

A minimum of 2 detector thimbles per core quadrant, and

' fficient movable detectors, drive, and readout equipment to map Su c.

these thimbles.

APPLICABILITY: When the movable incore detection system is used for:

a.

Recalibration of the excore neutron flux detection system, b.

Monitoring the QUADRANT POWER TILT RATIO using a full-core flux map per Specification 4.2.4.2, or H, F (Z) and F c.

Measurement of F q

xy ACTION:

With the movable incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.2 The movable incore detection system shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , by normalizing each detector output when required for:

a.

Recalibration of the excore neutron flux detection system, or b.

Monitoring the QUADRANT POWER TILT RATIO, or NaH, F (Z), and Fxy' c.

Measurement of F q

SUMMER - UNIT 1 3/4 3-46

3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY r,sures that the release of radioactive materials from the containment atmosphere will be restricted to those leakaoe paths and associated leak rates assumed in the accident analyses.

This restric-tion, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates (including those used in demonstrating a 30 day water seal) ensure that the total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, P.

As an added conservatism, the measured overall inte-grated leakage rate is further limited to less than or equal to 0.75 L or possibie,asapplicable,duringperformanceoftheperiodictesttoacc8untfor 0.75 L degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates are consistent with l

the requirements of Appendix "J" of 10 CFR 50.

l 3/4.6.1.3 REACTOR BUILDING AIR LOCKS The limitations on closure and leak rate for the reactor building air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate.

Surveillance testing of the air lock seals provide assurance that the overall air lock leakage will not become excessive due to seal damage during the intervals between air lock leakage tests.

l l

l SlM4ER - UNIT 1 B 3/4 6-1 Amendment No. 47

CONTAINMENT SYSTEMS BASES 3/4.6.1.4 INTERNAL PRESSURE l.

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 3.5 psig and l

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 positive 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.

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 force loss, the lift off force for a tendon is compared with the force predicted for the tendon times a reducti:n factor of 0.95.

This resulting force is referred to as the 95% Base Value.

The predicted tendon force is equal to the 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 5% reduction on the predicted force is intended to compensate for both uncertainties in the prediction techniques for the losses and for inaccuracies in the lift-off force measurements.

SUMMER - UNIT 1 B 3/4 6-2 Amendment No. 49

ADMINISTRATIVE CONTROLS RECORDS 6.5.2.10 Records of NSRC activities shall be prepared, approved and distributed as indicated below:

+

a.

Minutes of each NSRC meeting shall be prepared, approved and forwarded to the Vice President, Nuclear Operations within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be pre-pared, approved and forwarded to the Vice President, Nuclear Opera-l tions within 14 days following completion of the review.

c.

Audit summary reports encompassed by Section 6.5.2.8 above, shall be forwarded to the NSRC and to the Vice President, Nuclear Operations.

Full audits shall be forwarded to the management positions responsi-ble for the areas audited within 30 days after completion of the audit by the auditing organization.

5 6.5.3 TECHNICAL REVIEW AND CONTROL ACTIVITIES

=

6.5.3.1 Activities which affect nuclear safety shall be conducted as follows:

I a.

Procedures required by Technical Specification 6.8 and other proce-dures which affect plant nuclear safety, and changes thereto, shall be 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 which prepared the procedure or procedure change.

Procedures other than Administrative Frocedures will be approved as delineated in writing by the Director, Nuclear Plant Operations.

The Director, Nuclear Plant Operations will approve administrative procedures, security implement-l ing procedures and emergency plan implementing procedures.

Temporary approval to procedures which clearly do not 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 Senior Reactor Operator's License.

For changes to procedures which may involve a change in intent of the approved procedures, the person authorized above to approve the proce-dure shall approve the change.

b.

Proposed changes or modifications to plant nuclear safety-related structures, systems and components shall be reviewed as designated by the Director, Nuclear Plant Operations.

Each such modification shall be designed as authorized by Technical Services and shall be reviewed l

by an individual / group other than the individual / group which designed the modification, but who may be from the same organization as the individual / group which designed the modifications.

Implementation of modifications to plant nuclear safety-related structures, systems and components shall be concurred in by the Director, Nuclear Plant Operations.

SUMMER - UNIT 1 6-11 Amendment No. M, 49 1-

ADMINISTRATIVE CONTROLS Type of container (e.g., LSA, Type A, Type B, Large Quantity), and e.

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, i

The radioactive effluent release reports shall include any changes to the Process Control Program (PCP) made during the reporting period.

MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, in-cluding documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, l

U.S. Nuclear Regulatory Commission, Washington, D.C.

20555, with 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 6.9.1.11 The F limit for Rated Therr. r'ower (FRTP) shall be provided to xy x

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 Performance Branch, U. S. Nuclear Regulatory Commission, Washington, D.C.

20555 for all core planes containing bank "D" 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 some 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.

Any information needed to support F will be by request from the NRC and x

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.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the rollowing records shall be retained for at least the minimum period indicated.

SbMMER - UNIT 1 6-18 Amendment No. 35, 49

-

ML20211A654

Text

Attachment:

Navigation menu

Search