ML19242A071

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Amend 16 to License DPR-70,clarifying Moderator Temp Coefficient Specs,Reducing Required Range of Condensor Outlet Temp Detectors & Authorizing Removal of part-length Control Rods
ML19242A071
Person / Time
Site: Salem PSEG icon.png
Issue date: 06/01/1979
From: Schwencer A
Office of Nuclear Reactor Regulation
To:
Shared Package
ML19242A072 List:
References
NUDOCS 7907310183
Download: ML19242A071 (50)
v • d • e


Text

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UNITED STATES j

yN 7t NUCLEAR REGULATORY COMMISSION 7

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WASHINGTON, D. C. 20555

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PUBLIC SERVICE ELECTRIC AND GAS COMPANY PHILADELPHIA ELECTRIC COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STATION, UNIT NO. l_

AMENDMENT TO FACILITY OPERATING LICENSE Amendment NJ.16 License No. DPR-70 1.

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

A.

The application for amendment by Public Service Electric and Gas Company, et al. (the licensee) dated June 29, September 25, 1978, February 16, March 1, April 19 and 24,1979, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application,

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the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities

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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 Connission's regulations;

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D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with

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10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amendment by changes to the Technical

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Specifications as indicated in the attac'hment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. OPR-70 is hereby amended to read as follows:

"(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 16, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications."

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUC AR REGULATORY COMMISSION f.

v4MfAW A. Schwencer, Chief Operating Reactors Branch #1

,;_ g Division of Operating Reactors

Attachment:

Changes to the Technical Specifications Date of Issuance: June 1, 1979 i'.3.

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ATTACHMENT TO LICENSE AMENDMENT NO. 16 FACILITY OPERATING LICENSE NO. CPR-70 3.(

DOCKET NO. 50-272 7

Replace the following pages of the Appendix "A" ar.c' "B" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Paaes Appendix "A" III 1X 1-3 3/4 1-5 3/4 1-Sa (added) 3/4 1-18 3/4 1-20 3/4 1-26 3/4 2-3 3/4 2-7

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3/4 2-11 3/4 2-12 3/4 6-8 3/4 6-13 3/4 6-14 3/4 6-16 3/4 6-17 3/4 6-17a (added) 3/4 10-1 3/4 10-2 X

3/4 10-2a (delete) 3/4 10-3 B 3/4 1-1 B 3/4 1-2

.m B 3/4 1-3 B 3/4 1-4 8 3/4 2-1 B 3/4 2-4 5-4 Appendix "B" 2.1-1 2.1-3 2.1 -5 fy

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INDEX LIMITING CONDITIONS FOR OPERATION AN6 SURVEILLANCE REQUIREMENTS SECTION Page 3/4.0 APPLICABILITY.............................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTROL Shutdown Margin - T

> 200 F..........................

3/4 1-1 avg Shutdown Margin - T,yg < 200 F..........................

3/4 1-3 Boron Di1ution.........................................

3/4 1-4 Moderator Temperature Coefficient......................

3/4 1-5 Minimum Temperature for Criticality...............

3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown..................................

3/4 1-7 Flow Paths - 0perating.................................

3/4 1-8 Charging Pump - Shutdown...............................

3/4 1-10 C ha rg i n g Pump s - Ope ra t i n g.............................

3/4 1-11 Beric Acid Transfer Pumps - Shutdown...................

3/4 1-12 i

Boric Acid Transfer Pumps - Operating..................

3/4 1-13 Borated Water Sources - Shutdown.......................

3/4 1-14 Borated Water Sources - Operating......................

3/4 1-16

[.,

e 3/4.1.3 MOVABLE CONTROL ASSEMBLIES

+z Group Height...........................................

3/4 1-18 Position Indicator Channels............................

3/4 1-20 Rod Drop Time..........................................

3/4 1-21 Shutdown Rod Insertion Limit...........................

3/4 1-22 3:;g Control Rod Insertion Limits...........................

3/4 1-23

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SALEM - UNIT 1 III Amendment No. 16

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INDEX_

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Pace SECTION 3/4.2 POWER DISTRIBUTION LIMITS _

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3/42-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................

3/4 2-5 3/4.2.2 H E AT FLUX H OT C H ANN EL F ACT0R...........................

3/4.2.3 NUCLE AR ENTHALPY HOT CH ANNEL FACT 0R....................

3/42-9 r

3/4 2-11

$7 3/4.2.4 QUADRANT POWER TILT RAT!0..............................

3/4 2-13 3/4.2.5 DNB PARAMETERS.........................................

3/4.3 INSTRUMENT ATION_

3/4.3.1 RE ACTOR TRIP SYSTEM INSTRUMENT ATION....................

3/43-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM 3/4 3-14 INSTRUMENTATION........................................

3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...................

3/4 3-35 Movable Incore Detectors...............................

3/4 3-39 Seismic Instrumentation................................

3/4 3-40 Meteorological Instrumentation.........................

3/4 3-43 Remote Shutdown Instrumentation........................

3/4 3-46 Fire Detection Instrumentation.........................

3/4 3-49 l

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3/4.4 REACTOR COOLANT SYSTEM I

3/4.4.1 REACTOR COOLANT LOOPS 3/44-1 Normal Operation.......................................

3/4.4.2 SAFETY VALVES - SHUTD0WN...............................

3/44-4 3/4.4.3 SAFETY VALVES - 0PERATING..............................

3/4 4-5

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3/4 4-6 3/4.4.4 P RE S S U R I Z E R............................................

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3/4.4.5 STEAM GENERATORS.......................................

3/4 4-7 k

w 3/4.4.6 REACTOR C0OLANT SYSTEM LEAKAGE Leakage Detection Systems..............................

3/4 4-14 L*.'

Operational Leakage....................................

3/4 4-15 y

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SALEM - UNIT 1 IV Amendment No. 11

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INDEX BASES SECTION Page 3/4.0 APPLICABILITY..........................................

B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS e

3/4.1.1 80 RATION CONTR0L.....................................

B 3/4 1-1 i?

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~ E 3/4.1.2 B0 RATION SYSTEMS.................................... B 3/4 1-3 s

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3/4.1.3 M0VABLE CONTROL ASSEMBLIES...........................

B 3/A 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE................................

B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT

[Mk CHANNEL FACT 0RS......................................

B 3/4 2-4

[gfi g:c-3/4.2.4 QUADRANT POWER TILT RATI0............................

B 3/4 2-5

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3/4.2.5 DNB PARAMETERS.......................................

B 3/4 2-6 G'G5[N g..

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SALEM - UNIT 1 IX Amendment No.16

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a INDEX BASES t'

e SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............................

B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF) INSTRUMENTATION........

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION..............................

B 3/4 3-1 3/4.4 REACTOR COOLANT SYSTEM

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3/4.4.1 REACTOR COOLANT L00PS...................................

B 3/4 4-1 3/4.4.2 and 3/4.4.3 SAFETY VALVES......................

B 3/4 4-1 3/4.4.4 PRESSURIZER.............................................

B 3/4 4-2 3/4.4.5 STEAM GENERATORS........................................

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE..........................

B 3/4 4-3 k

3/4.4.7 CHEMISTRY...............................................

B 3/4 4-4 v.

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3/4.4.8 SPECIFIC ACTIVITY.......................................

B 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS......<......................

B 1/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4 4-12

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SALEM - UNIT 1 X

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DEFINITIONS CHANNEL FUNCTIONAL TEST 1.11 A CHANNEL FUNCTI0flAL TEST shall be the injection of a simulated signc1 into the channel as close to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CORE ALTERATION 1.12 CORE ALTERATION shall be the movement or manipulation of any com-ponent within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATION shall not preclude completion of movement of a cemponent to a safe conservative position.

SHUTDOWN MARGIN _

l.13 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present I 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 withdrcwn.

I IDENTIFIED LEAKAGE 1.14 IDENTIFIED LEAKAGE shall be:

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a.

Leakage (except CONTROLLED LEAKAGE) into closed systems, sut;1 as pump seal or valve packing leaks that are captured and conducted to a sump or collecting tank, or

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b.

Leakage into the containment atmrsphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE, or c.

Reactor coolant system leakage through a steam generator to the secondary system.

UNIDENTIFIED LEAKAGE 1.15 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

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5657TT

)'2 'l 5ALEM - UNIT l 1-3 Amendment No. 16

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'gEFINITIONS PRESSURE BOUNDARY LEAKAGE l.16 PRESSURE B0UNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

CONTROLLED LEAKAGE 1.17 CONTROLLED LEAKAGE shall be that seal water flow from the reactor l

coolant pump seals.

QUADRANT POWER TILT RATIO 1.18 QUADRANT POWER TILT RATIO shall be the ratic of the maximum upper excore detector calibrated output to the average of th upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (pCi/ gram) which alone would produce the same thyroid dose as the quantity and iso-topic mixture of 1-131,1-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be tnose listed in Table 111 of TID-14844, " Calculation of Distance Factors n

for Power and Test Reactor Sites."

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

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a.

A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified

.:s test interval into n equal subintervals, b.

The testing of one system, subsystem, train or other designated component at the beginning of eac4 subinterval.

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  1. G.C W SALEM - UNIT 1 1-4 Amendment No. 9 E-

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REACTIVITY CONTROL SYSTEMS MODERATOR' TEMPERATURE COEFFICIENT

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LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be:

a.

Less positive than 0 delta k/k/ F for the all rods withdrawn, beginning of cycle life (B0L), hot zero THERMAL POWER condition.

-4 b.

Less negative than -3.8 x 10 delta k/k/*F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.4.a - MODES 1 and 2* only#

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Specification 3.1.1.4.b - MODES 1, 2 and 3 only#

ACTICN:

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With the MTC more positive than the limit of 3.1.1.4.a, above, operations in MODES I and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than 0 delta k/k/ F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

mff 3.

A Special Report is prepared and submitted to the Ccmuission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its

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limit for the all rods withdrawn condition.

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b.

With the MTC more negative than the limit of 3.1.1.4.b, above,

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be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

  • With K greater than or equal to 1.0 j

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  1. See Special Test Exception 3.10.3

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SALEM - UNIT 1 3/4 1-5 Amendment No. 16

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1 REACTIVITY CONTROL SYSTEMS

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MODERATOR TEMPERATURE C0 EFFICIENT SURVEILLANCE REQUIREMENTS m.

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4.1.1.4 The MTC shall be determined to be within its limits during each fuel cycle as follows:

a.

The MTC shall be measured and compared to the BOL limit of Specification 3.1.1.4.a, above, prior to initial operation above 5% of RATED THERMAL POWER, af ter each fuel loading.

b.

The MTC shgli be measured at any THERMAL POWER and compared to

-2.9 x 10~ delta k/k/ F (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD af ter reaching an equilibrium b?ron concentration of 300 ppm.

In the event this comparison 4

indicates the MTC is more negative than -2.9 x 10 delta k/k/ F, the MTC shall be remeasured, and compared to the E0L MTC limit

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of specification 3.1.1.4.b, at least once per 14 EFPD during the remainder of the fuel cycle.

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4 SALEM - UNIT 1 3/4 1-Sa Amendment No.

16

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REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.5 The Reactor Coolant System lowest operating loop temperature (T,yg) shall be > 541 F.

APPLICABILITY: MODES 1 and 2.

ACTION:

With a Reactor Coolant System operating loop temperature (T) to within its limit with

< 541 F, restore (T H

a within the yi@xt 15 minutes.

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SURVEILLANCE REQUIREMENTS 4.1.1.5 The Reactor Coolant System temperature (Tavg) shall be determined to be > 541 F:

a.

Within 15 minutes prior to achieving reactor criticality, and b.

At least once per 30 minutes when the reactor is critical and theReactorCoolantSystemT.8Ereset.

is less than 551 F with the T

-T Deviation Alarm 8 avg ref With Keff > 1.0.

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ossg SALEM - UNIT 1 3/4 1-6

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) a.

At least once per 7 days by:

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1.

Verifying the boron concentration in each water source, 3;

2.

Verifying the water level of each water source, and 3.

Verifying the boric acid storage system solution temperature.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is < 35*F.

b ossrq SALEM - UNIT 1 3/4 1-17

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REACTIVITY CONTROL SYSTEMS 3/4.1.3. MOVABLE CONTROL ASSEMBLIES

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GROUP HEIGHT LIMITING CONDITION FOR OPERATION All full length (shutdown and control) rods, which are inserted 3.1.3.1 in the core, shall be OPERABLE and positioned within i12 steps (indicated position) of their bank demand position.

APPLICABILITY:

MODES 1* and 2*

ACTION:

With one or more full length rods inoperable due to being a.

immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b.

With more than one full length rod inoperable or misaligned from the bank demand pc.ition by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

With one full length rod.noperable due to causes other than c.

addressed by ACTION a, above, or misaligned from its group step counter demand height by more than i 12 steps (indicated position), POWER OPERATION may continue provided that within one hour either:

1.

The rod is restored to OPERABLE status within the above alignment requirements, or 2.

The rod is declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied.

POWER OPERATION may then continue provided that:

a)

An analysis of the potential ejected rod worth is performed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and the rod worth is determined to be < 0.95% Ak at zero power and < 0.21% Ak at RATED THERMAL POWER for the remainder of the fuel cycle, and seg g

  • See Special Test Exceptions 3.10.2 and 3.10.3.

SALEM - UNIT 1 3/4 1-18 Amendment No.16

4_'.

REACTIVITY CONTROL SYSTEMS

+.

LIMITING CONDITION FOR OPERATION (Continued) b)

The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and c)

A power distribution map is obtained frop the movable incore detectors and F (Z) and Ft'H are verified to be within theih limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, and d)

The THERMAL POWER level is reduced to < 75% of RATED THERMAL POWER within one hour and withTn the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the high neutron flux trip setpoint is reduced to < 85% of RATED THERMAL POWER, or e)

The remainder of the rods in the group with the inoperable rod are aligned to within + 12 steps of the inoperable rod within one hour whTle maintaining

3..e the rod sequence and insertion limits of Figures 3.1-1 and 3.1-2; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.5 during subsequent operation.

.6.

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full length rod shall be determined to be within the group demand limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when

=-

the Rod Position Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

4.1.3.1.2 Each full length rod not fully inserted shall be determined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

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p-d ge m SALEM - UNIT 1 3/4 1-19

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REACTIVITY CONTROL SYSTEMS

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POSITION' INDICATOR CHANNELS l-LIMITING CONDITION FOR OPERATION 4M

1; If 3.1.3.2 All shutdown and control rod position indicator channels and the demand position indication system shall be OPERABLE and capable of determining the control rod positions within + 12 steps.

APPLICABILITY:

MODES 1 and 2.

ACTION:

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With a maximum of one rod position indicator channel per group a.

inoperable either:

l.

Determ.ne the position of the non-indicating rod (s) in-directly by the movable incore detectors at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and inmediately af ter any motion of the non-

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indicating rod which exceeds 24 steps in one direction since the last determination of the rod's position, or 2.

Reduce THERMAL POWEP. TO < 50% of RATED THERMAL POWEP within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

b.

With a maximum of one demand position indicator per bank inoperable either:

1.

Verify that all rod position indicators for the affected bank are OPERABLE and that the most withdrawn rod and the least withdrawn rod of the bank are within a maximum of 12 steps o' each other at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or 2.

Reduce THERMAL POWER to < 50% of RATED THERMAL POWER within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.2 Each rod position indicator channel shall be determined to be

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OPERABLE by verifying the demand position indication system and the rod position indicator channels agree within 12 steps at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when th~e Rod Position Deviation Monitor is inoperable, then compare the de. mand position indication system and the rod position indicator channels at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

565297

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SALEM - UNIT 1

'3/4 1-25 Amendment No.,plb a

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F"" 5G 5234 SALEM - UNIT 1 3/4 1-26 Amendment No.16

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference pursuant to 4.2.1.3 above or by linear interpolation between the most recently measumed value and 0 percent at the end of the cycle life.

The provisions of Specification 4.0.4 are not applicable.

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56529o SALEM - UNIT 1 3/4 2-3 Amendment No.16

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POWER DISTRIBUTION LIMITS k SURVEILLANCE REQUIREMENTS (Continued) b) At least once per 31 EFPD, whichever occurs first. RTP 2. When the F is less than or equal to the F lircit for x the appropriate measured core plane, additional power C distribution maps shall be taken and F compared to RTP l F and F at least once per 31 EFPD x xp e. The F limits for RATED THERMAL POWER within specific core xy planes shall be: RTP 1. F 1 1.71 for all core planes containing bank "D" x control rods, and RTP 2. F 1 1.55 for all unrodded core planes. x f. The F limits of e, abovt, are not applicable in the fol-xy lowing core plane regions 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 + 2%, 32.1 + 2%, 46.4 + 2%, 60.6 + 2% and 74.9 + 2%, inclusive. 4. Core plane regions within + 2% of core height (+ 2.88 inches) about the bank demand position of the bank "D" control rods. Evaluating the effects of F, on F (Z) to determine if F (Z) g. x q q is within its limit whenever F exceeds ~F xy G= 4.2.2.3 When F (Z) is measured pursuant to specification 4.10.2.2, an 7 q overall measured F (Z) shall be obtained from a power distribution map q and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty. ~ 565292 SALEM - UNIT 1 3/4 2-7 Amendment No. 9,16 ..~ ' -

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Sta L ,n i POWER DISTRIBUTION LIMITS =.. QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION

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3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02. "~ APPLICABILITY: MODE 1 AB0VE 50% OF RATED THERMAL POWER

  • ACTION:

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but < 1.09: 1. Within 2 hours: i a) Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power P.ange Neutron Flux-High Trip betpoints within the next 4 hours. 2. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High Trip setpoints to < 55% of RATED THERMAL POWER within the next 4 hours. };{ 3 Identify and correct the cause of the out of limit con- = dition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed PEr provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER. b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod: 1. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in if E excess of 1.0, within 30 minutes. 2. Verify that the QUADRANT POWER TILT RATIO is within its ^ limit within 2 hours after exceeding the limit or 5652 %

  • $ee Special Test Exception 3.10.2.

SALEM - UNIT 1 3/4 2-11 Amendment No. 9, 16

POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION (Continued) reduce THERMAL POWER te less than 50% of RATED THERMAL POWER within the next 2 hours and reduce the Power Range Neutron Flux-High trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours. 3. Identify and correct the cause of the out of limit con-dition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours e until verified acceptable at 95% or greater RATED TH?'i>L POWER. c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the m.salignment of either a shut-down or control rod: 1. Reduce THERMAL POWER to less than 50% of RATED THEKMAL POWER within 2 hours and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55% of RATED THERMAL POWER within the next 4 hours. 2. Identify and correct the cause of the out of limit con-dition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER mvy proceed provided that the QU DRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours or until verified at 95% or greater RATED THERMAL POWER. SURVEILLANCE REQUIREMENTS 4.2.4 The QUADRANT POWER TILT RATIO shall be determined to be within the limit above 50% of RATED THERMAL POWER by: a. Calculating the ratio at least once per 7 days when the alarm is OPERABLE. b. Calculating the ratio at least once per 12 hours during steady state operation when the alarm is,, inoperable. Using the movable incore detectors to determine the QUADRANT POWER TILT RATIO at least once per 12 hours when one Power Range Channel is inoperable and THERMAL POWER is > 75 percent of RATED THERMAL POWER. SALEM - UNIT 1 3/4 2-12 l3(3l3;3gyeg.:.endment No.16

A CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature shall not exceed 120 F. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containmert average air temperature > 120 F, reduce the average air temperature to within the limit within 8 hours, or be in at least HOT-STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.6.1.5 The primary containment average air temperature shall be the arithmetical average of the temperatures at any 5 of the 10 following locations and shall be determined at least once per 24 hours: Location (Containment Perimeter) a. Elev. 84' - North f. Elev. 136' - South b. Elev. 106' - North g. Elev. 84' - South c. Elev. 78' - Northeast h. Elev. 106' - Southwest d. Elev. 84' - East i. Elev. 121' - West e. Elev. 106' - East j. Elev. 136' - West r. 565296 SALEM - UNIT 1 3/4 6-7

I 84 CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITIONS FOR OPERATION 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the structural integrity of the containment not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200 F. SURVEILLANCE REQUIREMENTS 4.6.1.6.1 Containment Surfaces The structural integrity of the exposed ~~ accessible interior and exterior surfaces of the containment, including i; the liner plate, shall be determined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of these surfaces. This inspection shall be performed prior to the Type A containment leakage rate test to verify no apparent changes in appearance or other abnormal degradation. 4.6.1.6.2 Reports Any abnormal degradation of the containment structure detected during the above required inspections shall be reported to the Commission pursuant to Specification 6.9.1. lhis report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective action taken. f-565297 ~ SALEM - UNIT 1 3/4 6-8 Amendment No.16

..= ;- CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.6.3.1.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by: Vertfying that on a Phase A containment isolation test signal, a. each Phase A tsolation valve actuates to its isolation position. b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position, Verifying that on a feedwater isolation test signal, each c. feedwater isolation valve isolates to its isolation position. d. Verifying that on a Containment Purge and Pressure-Vacuum Relief isolation test signal, each Purge and Pressure-Vacuum Relief valve actuates to its isolation position. 4.6.3.1.3 At least once per 18 months, verify that on a main steam i-l isolation test signal, each main steam isolation valve specified in Table 3.6-1 actuates to its isolation position. L'.'.' 565298 SALEM - UNIT 1 3/4 6-13 Amendment No.16

TABLE 3.6-1 CONTAINMENT ISOLATION VALVES h VALVE NUMBER FUNCTION ISOLATION TIME h A. PHASE "A" ISOLATION 5 H 1. 1 PR 17* Pressurizer Relief Tk.-Gas Analyzer Conn. < 10 Sec. 2. 1 PR 18* Pressurizer Relief Tk.-Gas Analyzer Conn. [10Sec. 3. 1 NT 25* Pressurizer Relief Th.-N Connection < 10 Sec. 7 4. 1 WR 80* Pressurizer Relief Tk.-Primary Water Conn. [10Sec. 5. 1 CV 3 CVCS - Letdown Line < 10 Sec. 6. 1 CV 4 CVCS - Letdown Line 7 10 Sec. 7. 1 CV 5 CVCS - Letdown Line 7 10 Sec. 8. 1 CV 7 CVCS - Letdown Line 7 10 Sec. 9. 1 CV 68## CVCS - Charging Line [10Sec. 10. 1 CV 69## CVCS - Charging Line 1 10 Sec. 11. 1 CV 284 CVCS - RCP Seals 1 10 Sec. m2 12. 1 CV 116 CVCS - RCP Seals 1 10 Sec. 13. 1 CC 215 Comp. Cooling to Excess Letdown Hx 1 10 Sec. L 14. 1 CC 113 Comp. Cooling to Excess LetdowTi Hx 1 10 Sec. 15. 1 WL 96* RC Drain Tk - Gas Analyzer Conn. < 10 Sec. 16. 1 WL 97* RC Drain Tk - Gas Analyzer Ccon. [10Sec. ^ 17. 1 WL 98* RC Drain Tk - Vent Header Conn. 1 10 Sec. 18. 1 WL 99* RC Drain Tk - Vert Header Conn. < 10 Sec. 19. 1 WL 108* RC Drain Tk - N Connection 7 10 Sec. E 20. 1 WL 12* RCDrainTankPbmps 7 10 Sec. 21. 1 WL 13* RC Drain Tank Pumps 7 10 Sec. g4 m 22. 1 NT 32* f.ccumulator N2 Supply [10Sec. C1 23. 1 SJ 123* SI Test Line < 10 Sec. N 24. 1 SJ 60* SI Test Line 7 10 Sec. 'T E 25. 1 SJ 53* SI Test Line 7 10 Sec. O 26. 1 SS 103* Accumulator Sampling 7 10 Sec. K 27. 1 SS 27* Accumulator Sampling i 10 Sec. 28. 1 SS 104* RC Sampling i 10 Sec. 29. 1 SS 33* RC Sampling 1 10 Sec. 30. 11 SS 94* SG Blowdown Sampling 1 10 Sec. ynn g

TABLE 3.6-1 (Continued) ">G CONTAINMENT ISOLATION VALVES

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3 I "=- Q VALVE NUMBER FUNCTION ISOLATION TIME li A. PHASE "A" ISOLATION (Continued) P i! 31. 12 SS 94* SG Blowdown Sampling < 10 Sec. 32. 13 SS 94* SG Blowdown Sampling 7 10 Sec. n 33. 14 SS 94* SG Blowdown Sampling 7 10 Sec. p 34. 1 SS 107* Pressurizer Liquid Sampling [10Sec. 35. 1 SS 49* Pressurizer Liquid Sampling < 10 Sec. 36. 1 SS 110* Pressurizer Steam Sampling [10Sec. 37. 1 SS 64* Pressurizer Steam Sampling < 10 Sec. 2 38. 1 VC 7 Containment Radiation Sampling i10Sec. w 39. 1 VC 8 Containment Radiation Sampling < 10 Sec. m L 40. 1 VC 11 Containment Radiation Sampling i10Sec. U m 41. 1 VC 12 Containment Radiation Sampling 1 10 Sec. 42. 11 CA 330 Instrument Air Supply 1 10 Sec. H 43. 12 CA 330 Instrument Air Supply < 10 Sec. 44. 1 DR 29 Demineralized Water Supply i10Sec. I! 45. 11 GB 4 Steam Generator Blowdown 1 10 sec. 3 46. 12 GB 4 Steam Generator Blowdown < 10 sec. 47. 13 GB 4 Steam Generator Blowdown 7 10 sec. 48. 14 GB 4 Steam Generator Blowdown 7 10 sec. 49. 1 WL.16 Containment Sump Discharge 7 10 sec. 50. 1 WL 17 Containment St.mp Discharge [10sec. UI 51. 1 FP 147* Fire Protection System i 10 sec. O Cll B. PHASE "B" ISOLATION CJ < 10 sec. O 1. 1 CC 118 Component Cooling to RCP 7 10 sec. O 2. 1 CC 117 Component Cooling to RCP 3. 1 CC 187 Component Cooling to RCP [10sec. < 10 sec. 4. 1 CC 136 Component Cooling to RCP 710 sec. 5. 1 CC 190 Component Cooling to RCP 6. 1 CC 131 Component Cooling to RCP [ 10 sec. N w .mv m 6

TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES G 2c VALVE NUMBER FUNCTION ISOLATION TIME B E C. MAIN STEAM ISOLATION 1. 11 MS 7# Main Steam Drain < 10 Sec. m 2. 12 MS 7# Main Steam Drain 7 10 Sec. 3. 13 MS 7# Main Steam Drain 710 Sec. 4. 14 MS 7# Main Steam Orain 7 10 Sec. 5. 11 MS 18# Main Steam Bypass [ 10 Sec. 6. 12 MS 18# Main Steam Bypass 1 10 Sec. 7. 13 MS 18# Main Steam Bypass 1 10 Sec. 8. 14 MS 18# Main Steam Bypass 1 10 Sec. D. FEEDWATER ISOLATION R ^ 1. 11 BF 19# Main Feedwater Isolation 1 8 Sec. T 2. 12 BF 19# Main Feedwater Isolation 1 8 Sec. g 3. 13 BF 19# Main Feedwater Isolation 1 8 Sec. 4. 14 BF 19# Main Feedwater Isolation < 8 Sec. a 5. 11 BF 40# Main Feedwater Isolation 7 8 sec. 6. 12 BF 40# Main Feedwater Isolation 7 8 Sec. 7. 13 BF 40# Main Feedwater Isolation 7 8 Sec. 8. 14 BF 40# Main Feedwater Isolation [8Sec. CJI E. CONTAINMENT PURGE O AND PRESSURE-VACUUM RELIEF N C/l n E. C+0 1. 1 VC 1* Purge Supply < 2 Sec. I 2. 1 VC 2*# Purge Supply [2Sec. h E 3. 1 VC 3*# Purge Exhaust 1 2 Sec. 4. 1 VC 4* Purge Exhaust i 2 Sec. 2 P 5. 1 VC 5* Pressure-Vacuum Relief < 2 Sec. g 6. 1 VC 6*# Pressure-Vacuum Relief [2Sec. mr r nu .g .w g.

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TABLE 3.6-1 (Continued) CONTAINMENT ISOLATION VALVES VALVE NUMBER FUNCTION ISOLATION TIME g [ F. MANUAL 1. (2 valves)# Pressurizer Dead-Weight Calibrator Not Applicable 2. 11 CV 98# CVCS - RCP Seals Not Applicable f 3. 12 CV 98# CVCS - RCP Seals Not Applicable ~' 4. 13 CV 98# CVCS - RCP Seals Not Applicable 5. 14 CV 98# CVCS - RCP Seals Not Applicable 6. 1 SJ 71# CVCS Flushing Connection Not Applicable 7. 11 SS 93*# Steam Generator Sampling Not Applicable 8. 12 SS 93*# Steam Generator Sampling Not Applicable w 9. 13 SS 93*# Steam Generator Sampling Not Applicable 1 10. 14 SS 93*# Steam Generator Sampling Not Applicable m 11. 1 SA 118# Compressed Air Supply Not Applicable i 12. 1 WL 190# Refueling Canal Supply Not Applicable 13. 1 SF 36# Refueling Canal Supply Not Applicable 14. 1 WL 191# Refueling Canal Discharge Not Applicable s 15. 1 SF 22# Refueling Canal Discharge Not Applicable 16. 1 VC 9*f Containment Radiation Sampling Not Applicable 17. I VC 10*# Containment Radiation Sampling Not Applicable 18. 1 VC 13*# Containment Radiation Sampling Not Applicable I 19. 1.VC 14*// Containment Radiation Sampling Not Applicable 20. Fuel Transfer Tube Not Applicable N, Cl am a Cn E Cd r* n E N t n u

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TABLE 3.6-1 (Continued) l$ _ CONTAINMENT ISOLATION VALVES 9 [ VALVE NUMBER FUNCTION ISOLATION TIME El G. CHECK 1. 1 CV 74 CVCS - Charging Line Not Applicable 2. 1 CV 296 CVCS - RCP Seals Not Applicable 3. 1 CC 186 Component Cooling to RCP Not Applicable 4. 1 CC 208 Component Cooling to RCP Not Applicable 5. 1 NT 26 Pressurizer Relief Tk. - Nitrogon Conn. Not Applicable 6. 1 WR 81 Pressurizer Relief Tk. - Primary Water Not Applicable Conn. R a ? G w

  • May be opened on an intermittent basis under administrative control.
  1. Not subject to Type C leakage tests.
    1. Either valve 1 CV 68 or 1 CV 69 must be OPERABLE.

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r a i CONTAINMENT SYSTEMS 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS g; LIMITING CONDITION FOR OPERATION {.{ 3.6.4.1 Two independent containment hydrogen analyzers shall be OPERABLE. APPLICABILITY: MODES 1 and 2. ACTION: With one hydrogen analyzer inoperable, restore the inoperable analyzer to OPERABLE status within 30 days or be in at least HCT STANDBY within the next 6 hours. 1... SURVEILLANCE REQUIREMENTS 4.6.4.1 Each hydrogen analyzer shall be demonstrated OPERABLE at least $ =.. Once per 92 days on a STAGGERED TEST BASIS by perfonning a CHANNEL =' CALIBRATION using sample gases containing: y G Zero volume percent hydrogen, balance purging air without any a. free hydrogen, and b. Two volume percent hydrogen, balance air. ~ e. SALEM - UNIT 1 3/4 6-18 .n E

3/4.10 SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN '.*r LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided the reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s),and APPLICABILITY: MODE 2. ACTION: a. With any full length control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at > 10 gpm of 20,100 ppm boric acid solution or its equivalent untiT the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. =.. =.. b. With all full length control rods inserted and the reactor sub-critical by less than the above reactivity equivalent, immediately initiate and continue boration at > 10 gpm of 20,100 ppm boric acid solution or its equivalent until the SHUTDOWM MARGIN required by Specification 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS = 4.10.1.1 The position of each full length and part length rod either i partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each full length rod not fully inserted shall be demonstrated capable of full, insertion when tripped from at least the 50% withdrawn position within 24 hours prio* to reducing the SHUTDOWN MARGIN to less z..=.x than the limits of Specification 3.1.1.1. 555." f. 5G5305 SALEM-UNIT 1 3/4 10-1 Amendment No.16

AW SPECIAL TEST EXCEPTIONS

~~;

GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION r,*"" 3.10.2 The group height, insertion and power distribution limits of Specifications 3.1.3.1, 3.1. 3. 4, 3.1. 3. 5, 3.2.1, and 3.2. 4 may be suspended during the performance of PHYSICS TESTS provided: a. The THERMAL POWER is maintained < 85% of RATED THERMAL POWER, and b. The limits of Specifications 3.2.2 and 3.2.3 are maintained and determined at the frequencies specified in Specification f.10.2.2 below. APPLIr. ABILITY: MODE 1 ACTION: With any of the limits of Specifications 3.2.2 or 3.2.3 being exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.4, 3.1.3.5, 3.2.1 and 3.2.4 are suspended, either: a. Reduce THERMAL POWER sufficient to satisfy the ACTION require-ments of Specifications 3.2.2 and 3.2.3, or b. Be in HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined to be < 85% of RATED THERMAL POWER at least once per hour during PHYSICS TE5'TS. }[b 4.10.2.2 The Surveillance Requirements of Specifications 4.2.2 and "1 4.2.3 shall be performed at the following frequencies during PHYSICS s}$ 9=: TESTS: =- a. Specification 4.2.2 - At least once per 12 hours. (If 3 b. Specification 4.2.3 - At least onge per l? hours. s 565306 SALEM-UNIT 1 3/4 10-2 Amendment No. 5, g, 16

e.e l SPECIAL TEST EXCEPTIONS PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.3 The limitations of Specifications 3.1.1.4, 3.1.3.1, 3.1.3.4, and 3.1.3.5 may be suspended during the performanca of PHYSICS TESTS provided: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and b. The reactor trip setpoints on the OPERABLE Intermediate and Power Range Nuclear Channels are set at < 25% of RATED THERMAL POWER. APPLICABILITY: MODE 2. ACTION: With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers. SURVEILLANCE REQUIREMENTS 4.10.3.1 The THERMAL POWER shall be determined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TES'TS. 4.10.3.2 Each Intermediate and Power Rangt Channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating PHYSICS TESTS.

== .._1..

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e... f'. 5653C7 ~ ~ SALEM-UNIT 1 3/4 10-3 Amendment No.16

f f.' t t SPECIAL TEST EXCEPTION NO FLOW TESTS 2-- LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specification 3.4.1.1 may be suspended during the performance of startup and PHYSICS TESTS, provided: a. The THERMAL POWER does not exceed the P-7 Interlock Setpoint, and b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power Range Channels are set ; 25% of RATED THERMAL POWER APPLICABILITY: During operation below the P-7 Interlock Setpoint. ACTION: With the THERMAL POWER greater than the P-7 Interlock Setpoint, immediately open the t' actor trip breakers. SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be determined to be less than P-7 Interlock Setpoint at least once per hour during startup and PHYSICS TESTS. 4.10.4.2 Each Intermediate, Power Range Channel and P-7 Interlock shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating startup or PHYSICS TESTS. c 5G53C8 SALEM-UNIT 1 3/4 10-4

w. r.- e /d 3/4.1 REACTIVITY CONTROL SYSTEMS ~7 }+. BASES _ _. 4 3/4.1.1 BORAT10N CONTROL 1-3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGlH A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all aperating conditions, 2) the reactivity tranrients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will oe maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition. SHUTDOWN MARGIN requirements vwy throughout core life as a function of fuel depletion, RCS boron concen. ' tion, and RCS T The most restrictive condition occurs at EOL, with T at no i dd operating temperature, and is associater' with a postuNed steam line break accident ~ and resulting uncontrolled %S cooldown. In the analysis of this accident,

== a minimum SHUTDOWN MARGIN of 1.6% ak/k is in.tially e cmaired to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement . =.. is based upon this limitNg condition and is consistent with FSAR safety analysis assumotions. V,th Tavg i 200 F, the reactivity transients resulting from a postsla: steam line break cooldown are minimal and a 1% ak/k shutdown margin provides adequate protection. .g 3/4.1.1.3 BORC N DU. Ui10N [ %Q A minimum flow rate of at least 3000 GPM provides adequate mixing, jM prevents stratification and ensures that reactivity changes will be gig gradual during boron concentration reductions in the Reactor Coolant 1_ System. A flow rate of at least 3000 GPM will circulate an equivalent ~ Reactor Coolant System volume of 12,612 cubic feet in approximately 30 minutes. The reactivity change rate associated with boron reductions E will therefore be within the capability for operator recognition and D control.

~.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) 5. The limitations on MTC are provided to ensure that the value of this coefficient remains within the limiting condition assumed in the accident and transient analyses. g. 5653C9 4 SALEM - UNIT 1 B 3/4 1-1 Amendment No.16 m

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison. The most negative MTC value equivalent to the most positive moderator density coefficient (MDC), was obtained by incrementally correcting the MDC used in the FSAR analyses to nominal operating conditions. These corrections involved subtracting the incremental changes in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions. This valge of the MDC was then transformed into the limiting -4 MTC value - 3.8 x 10 ok/k/ F. The MTC value of -2.9 x 10 ak/k/ F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtgined by making these corrections to the limiting MTC value -3.8 x 10 k/k/ f, The surveillance requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains with its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup. 3/4.1.1,6 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 F. This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in an OPEFABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTNDT temperature. g g 31# SALEM - UNIT 1 B 3/4 1-2 Amendment No.16 h.

REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORAT10N SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators. With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable Nt-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injcction system failures during the repair period. The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 1.6% ok/k after xenon decay and cooldown to 200 F. The maximum boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 5106 gallons of 20,100 ppm borated water from the boric acid storage tanks or 75,000 gallons of 2000 ppm borated water f rom the refueling water storage tank. However, to be consistent with the ECCS requirements, the RWST is required to have a minimum contained volume of 350,000 gallons 3 during operations in MODES 1, 2, 3 and 4. With the RCS temperature below 200 F, one injection system is l' acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restric-tions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable. i-The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200 F to 140'F. This condition requires either 835 gallons of 20,100 ppm borated water from the boric acid storage tanks or 9690 gallons of 2000 ppm borated water from the refueling water storage tank. The contained water volume limits include allowance for water not available because of discharge line location and other physical SGZ1) characteristics. The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6. p_ SALEM - UNIT 1 B 3/4 1-3 Amendment No. 16

t REACTIVITY CGNTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of a rod ejection accident. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and' insertion limits. The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the triginal criteria are met. Misalignment of a rod requires measure-ment of peaking factors or a restriction in THERMAL POWER; either of these restrictions provide assurance of fuel rod integrity during continued The reactivity worth of a misaligned rod is limited for the operation. remainder of the fuel cycle to prevent exceeding the assumptions used in ~ the accident analysis. The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses. Measurement with Tavg 2.541'F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions. I Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent verifications required if an automatic monitoring channel

  1. E-is inoperable. These verification frequencies are adequate for assuring that the applicable LC0's are satisfied.

bh ( 1 SALEM - UNIT 1 B 3/4 1-4 Amendment No. 16

. ~.., 3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integ-rity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events oy: (a) maintaining the minimum DNBR in the core > 1.30 during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded. The definitions of hot channel factors as used in these specifi-cations are as follows: F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local 0 heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for man-ufacturing tolerances on fuel pellets and rods. F Nuclear Enthalpy Rise Hot Channel Factor, is defined as the g ratio of the integral of linear power along the rod with the highest integrated power to the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) The lim.ts on AXI AL FLUX DIFFERENCE assure that the F (Z) upper n bound envelope of 2.32 times the normalized axial peaking Yactor is not 1. exceeded during either normal operati n or in the event of xenon redis-a tribution following power changes. Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. The value of the target flux difference obtained under these conditions ser divided by the fraction of RATED THERMAL POWER is the target flux g7-difference at RATED THERMAL POWER for the associated core burnup

5 conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriata

~~ fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflqct core burnup considerations. 565313 J SALEM - UNIT 1 B 3/4 2-1 Amendment No.16

POWER DISTRIBUTION LIMITS BASES Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the +5% target band about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to dcviate outside of the target band at re-duced THERMAL POWER levels. This deviation will not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL P0hER (with the AFD within the target band) provided the time duration of the devi-ation is limited. Accordingly, a 1 hour penalty deviation limit cumJ-lative during the previous 24 hours is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER. For THERMAL POWER levels between 15% and 50% of rated THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penTity of 2 hours actual time reflects this reduced significance. Provisions for monitoring the AFD are derived from the plant nuclear instrumentation system through the AFD Monitor Alarm. A control room recorder continuously displays the auctioneered high flux difference and the target band limits as a function of power level. A first alarm is received any time the auctioneered high flux difference exceeds the target band limits. A second alarm is received if the AFD exceeds its allowable limits for a cumulative time of one hour during any 24 hour + - - time period starting with the occurrence of the first alarm. Time out-side the target band is graphica.y presented on the strip chart. r Figure B 3/4 2-1 shows a typical monthly target band. sa r 565314 SALEM - UNIT 1 B 3/4 2-2 . me n

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5 ~ - ~ ~ ~ ~ . _.....+ =._ _. _ _..__m_ =._..=._.i=._.n_2-:-=.=..*=..- :=...=.~ -..L.- .t=.._- = _, _.__,a__=.+._._-- t _ =.. - -.= 2_ t=- 30% 20% 10% 0 +1 +20% +30% INDICATED AXIAL FLUX DIFFERENCE 4 Figure B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER SALEM - UNIT 1 B 3/4 2-3 ' 565315

1 81 I POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/1.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS-F (Z) and F q H The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not cxceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit. Each of these hot channel factors are measurable but will normally only be deterrnined periodically as specified in Specifications 4.2.2 and 4.2.3. This feriodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided: Control rod in a single group move together with no individual a. rod insertion differing by more than +_ 12 steps from the group demvnd position. b. Control rod groups are sequenced with overlapping group.; as described in Specification 3.i.3.5. The control rod insertion limits of Specifications 3.1.3.4 and c. 3.1.3.5 are maintained. d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. The relaxation in F" as a function of THERMAL POWER allows changes ig the radial power shaph"for all permissible rod insertion limits. will be maintained within its limits provided conditions a thru d F abb4ve, are maintained. When an F measurement is taken, both experimental error and man-n ufacturing tolerance must be allowed for. 5% is the appropriate allowance far a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance. When F" is measured, experimental error must be allowed for and 4% istheapprbh;riateallowanceforafullcoremaptakenwiththeincore g The specified limit for F also contain; an 8% detection system. alloyanceforur.certaintieswhichmeanthath$rmaloperationwillresult in F The 8% al Mwance is based on the following considera- " "Qg < l.55/1.0F;. 565316 SALEM - UNIT 1 B 3/4 2-4 Amendment No.16

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s b n DESIGN FEATURES [.7 bt1 0 DESIGN PRESSURE AND TEMPERATURE

j..

5.2.2 The reactor containment building is designed and shall be main-tained for a maximum internal pressure of 47 psig and an air temperature of 271 F. 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel assembly containing 264 fuel rods clad with Zircaloy -4. Each fuel rod shall have a nominci active fuel length of 143.7 inches and contain a maximum total weight of 1766 grams uranium. The initial core loading shall have a maximum enrichment of 3.35 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 3.5 weight percent U-235. CONTROL RCD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length and no part length control rod assemblies. The full length control rod assemblies shall contain a nominal 142 inches of absorber material. The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium. All control rods shall be clad with stainless steel tubing. t. 5.4 REACTOR C0OLANT SYSTEM DESIGN PRESSURE AND TEtiPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained: 5G5318 SALEM - UNIT 1 5-4 Amendment No.16

2.0 LIMITlh; CONDITIONS FOR OPERATION 2.1 T HERMAL 2.1. l a MAXIMUM T ACROS5 CONDENSER DURING NORMAL OPERATION Ctjective To limit thermal stress to the aquatic ecosystem by limiting the maximum L '. across the condenser during normal operation. Speci fication The maximum AT across the condchser shall not exceed 16.5 F during normal operation with all six circulating water pumps operating. In the event that the specification is exceeded corrective action shall be taken to reduce the iT to within specification. Such corrective action could include cleaning condenser water box;s er re, duction of unit power level, unless an emergency need for power exists. Monitoring Requi rerent The temperature differential across the condenser shall be monitored every hour utilizing the computer prir. tout of the intake and discharge temperature measurements. The intake temperature is measured at each of the two inlets to each condenser shell. The discharge temperature is measured .t a point downstream of the condenser in each of the two 84-inch ID discharge lines from each condenser shell. Instrumentation accuracy is +0.5 F between the range of 32 F and 150 F. If the plant computer is out of service, the intake and discharge temperatures shall b3 monitored every two hours utilizing local reading instrumentation until the plant computer is returned to service. 565319 2.1-1 Amendment No.16

2. Tne maximur ai across the condenser shall not exceed 16.5'F for more than 72 consecutive hours for reasons of pump failure. 3. At no time will the aT across the condenser exceed 27.5 F. In the event that either specification is exceeded, corrective action shall be taken to reduce the aT to within specification. Such correc-tive action could include cleaning condenser water boxes or reduction of limit power level, unless an emergency need for power exists. Monitoring Requi rement The temperature differential across the condenser shall be monitored every hour utilizing the computer printout of the intake and discharge ter@erature measurements. The intake temperature is measured at each of the two inlets to each condenser shell. The discharge temperature is measured at a point downstream of the condenser in each of the two 84-inch ID discharge lines from each condenser shell. Instrumentation accuracy is +0.5 F between the range of 32 F and 150'F. If the plant conputer is out of service, the intake and discharge temperatures shall be monitored every two hours utilizing local reading instrunntation until the plant computer is returned to service. 565320 Bases The condenser coeling water system was designed to operate with a AT that would minimize *.hermal stress to organisms. The U.S. Environmenta' Protection Agency has set a limit of 27.5 F as a maximum aT permitted under the NPDES 2.1-3 Amendment No. 16

2. In the event that fewer than six circulating water pumps are in operation, the maximum condenser discharge water temperature shall not exceed 115"F for more than eight consecutive hours within any 24 hour period. 3. In the event specifications 2.1.2.1 or 2.1.2.2 are exceeded corrective action shall be taken to reduce the condenser discharge water tempera-ture to wit 1in specification. Such corrective action could include cleaning condenser water boxes or reduction of unit power level, unless an emergency need for poe r exists. Monitorina Requirement Discharge temperature shall be monitored every hour utilizing the average of the corr.puter printout of the discharge temperature measurements. The discharge temperature is measured at a point downstream of the condenser in each of the two 84-inch ID discharge lines from each condenser shell. Instrumentation accuracy is +0.5 F between the range of 32 F and 150 F. If the plant computer is out of service, the discharge temperature shall be monitored every two hours utilizing local reading instrumentation until the plant computer is returned to service. 565324 Bases Ichthyological Associates (IA) studies performed from June 1968 through December 1973 show 25 records of river temperatures > 84 F. Twenty-one of 2.1-5 Amendn.eqt No.16}}

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