ML20199G058

Amend 197 to License DPR-75,incorporating Into TSs Margin Recovery Portion of Fuel Upgrade Margin Recovery Program

ML20199G058
Person / Time
Site:	Salem
Issue date:	01/08/1999
From:	Bill Dean NRC (Affiliation Not Assigned)
To:
Shared Package
ML20199G065	List: ML20199G058 ML20199G079
References
NUDOCS 9901220170
Download: ML20199G058 (48)

Text

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UNITED STATES s

j NUCLEAR REGULATORY COMMISSION 3

2 WASHINGTON D.C. 20066 4001 PUBLIC SERVICE ELECTRIC & GAS COMPANY PHILADELPHIA ELECTRIC COMPANY l

DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-311 SALEM NUCLEAR GENERAT!NG STATION. UNIT NO. 2 6

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 197 License No. DPR-75 1.

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

A.

The application for amendment filed by the Public Service Electric & Gas Company, Philadelphia Electric Company, Delmarva Power and Light Company and Atlantic City Electric Company (the licensees) dated May 10,1996, as supplemented March 19 and August 29,1997, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and

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regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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 Commission's regulations set forth in 10 CFR Chapter I; 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 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

L 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating Ucense No. DPR-75 is hereby amended to read as follo.vs:

l 9901220170 990100 PDR ADOCK 05000311 P

PDR

b

< >. - (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendices A and B as revised through Amendment No.197, 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 Rs date of issuance, to be implemented within 60 days.

j FOR THE NUCLEAR REGULATORY COMMISSION hf William M. Dean, Director Project Directorate 1-2 Division of Reactor Projects - t/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance: January 8.1999

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ATTACHMENT TO LICENSE AMENDMENT NO.197 FACILITY OPERATING LICENSE NO. DPR-75 DOCKET NO. 50-311 Revise Appendix A as follows:

Remove Paoes Insert Paaes I

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IV IV XI XI 1-2 1-2 2-1 2-1 2-2 2-2 2-3 2-3 2-5 2-5 2-7 2-7 2-8 2-8 2-9 2-9 B 2-1 B 2-1 B25 B 2-5 B24 B24 3/4 1-1 3/4 1 1 3/4 1-2 3/4 1-2 3/4 1-4 3/4 1-4 3/4 1-5 3/4 1-5' 3/4 1-13 3/4 1-13 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 1-22 3/4 1-22 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 l

3/4 2 4 3/4 2-4 3/4 2-5 3/4 2-5 l

3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8

)

3/4 2-9 3/4 2-9 3/4 2-17 3/4 2-17 B 3/4 1-1 B 3/4 1-1 i

B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 8 3/4 2-2 B 3/4 2-3 B 3/4 2-3 B 3/4 2-4 8 3/4 2-4

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1 Remove Panes insert Panes I

B 3/4 2-5 B 3/4 2-5 B 3/4 4-1 B 3/4 4-1 54 5-4 S 24 6-24 6-24a 6 24a I

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O INDEX DEFINITIONS SECTIOR E. Mil 1.0 DEFINITIONS t

DEFINED TERMS l

1-1 h,

ACTION

...............................1-1 AXIAL FLUX DIFFERENCE 1-1 CHANNEL CALIBRATION 1-1 CHANNEL CHECK 1-1 CHANNEL FUNCTIONAL TEST 1

1-1 CONTAINMENT INTEGRITY 1-2 CORE ALTERATION 1-2 CORE OPERATING LIMITS REPORT.

I 1-2 l

20SE EQUIVALENT I-131 1-2 E-AVERAGE DISINTEGRATION ENERGY 1-3 l

ENGINEERED SAFETY PEATURE RESPONSE TIME 1-3 mEQUENCr NOTATION 1-3 FULLY WITHDRAWN 1-3

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GASEOUS RADWASTE TREATMENT SYSTEM 1-3 i

IDENTIFIED LEAKAGE 1-3

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HEMBER(S) OF THE PUBLIC 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCH) 1-4 OPERABLE - OPERABILITY 1-4 OPERATIONAL LODE 1-4 PHYSICE TESTS 1-5 PRESSURE BOUNDARY IZAKAGE 1

1-5 PROCESS CONTROL PROGRAM (PCP) 1-5 PURGE-PURGING

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1-5.

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QUADRANT POWER TILT RATIO 1-5 RATED THER>mI. POWER 1-5 REACTOR TRIP SYSTEM RESPONSE TIME 1-6 l

REPORTABLE EVENT i

1-6 SHUTDOWN MARGIN 1-6 SITE BOUNDARY 1-6 SOLIDIFICATION 1-6 SOURCE CHECK 1-6 STAGGERED TEST BASIS 1-6 THERMAL POWER 1-7 UNIDENTIFIED IZAKAGE 1-7 UNRESTRICTED AREA. ~.. ~ '.

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L VENTILATION EXHAUST TREATMENT SYSfEM 1-7 VENTING 1*7 l

SALEM - UNIT 2 I

Amendment No.197

l INDEX SAFETY LIMITS AND LIMITING SAFETY h7dTM titTTINGS SECTION g

2.1 SAFETY LIMITS Reactor Core.............................................. 2-1 Reactor Coolant Sys tem Pressure.......................... 2-3 l

2.2 LIMITING SAFITY SYSTEM. SETTINGS Reactor Trip System Instrumentation. CMinu............ 2-4 i

BASES SECTION PAGE 2.1 SAFETY LIMITS Reactor Core............................

.B 2-1 Reactor Coolant System Pressure..........

.B 2-2 2 2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip System Instrumentation detpoints..............B 2-3 l

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SALEM - UNIT 2 II Amencent No.197 i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEII. LANCE REQUIREMENTS SECTION J2gg i

i l

3/4.2 POWER DISTRIBUTION LIMITS j

3/4.2.1 AXIAL FLUX DIFFERENCE................................ 3/4 2-1 4'

.i 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR......................... 3 /4 2-5 i

3/4.2.3 NUCLEAR ENTHALFY HOT CHANNEL FACTOR................... 3/4 2-9 l

3/4.2.4 QUADRANT POWER TILT RATIO............................ 3 /4 2-13 i

3/4.2.5 DNB PARAME TERS....................................... 3 /4 2-16 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION.................. 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYS7.EM INSTRUMENTATION...................................... 3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation................. 3/4 3-38 Novable Incore Detectors............................. 3/4 3-42 Remote Shutdown Instrumentation...................... 3/4 3-4 3 Accident Monitoring Ins trumentation................... 3/4.3-50 Radioactive Liquid Effluent Honitoring Ins trumenta ti on...................................... 3 /4 3 - 5 3 Radioactive Gaseous Effluent Monitoring Instrumentation...................................... 3/4 3-59 3/4.3.4 TURBINE OVERSPEED PROTECTION......................... 3/4 3-65 i

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SALEM - UNIT 2 IV Amendment No.197

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4 INDEX BASES SECTION M

3/4.0 1.PPLICABILITY.........................................B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS T'

3/4.1.1 BORATION CONTROL......................................B 3/4 1-1 3/4.1.2 I I BORATION SYSTEMS......................................B 3/4 1-3 3/4.1.3 mVABLE CONTROL ASSEMBLIE S............................B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE

.................................B 3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR and j

3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR...................B 3/4 2-4 3/4.2.4 QUADRANT POWER TILT RATIO............................. B 3 /4 2-5 3/4.2.5 DNB PARAMETERS........................................B 3/4 2-5 6

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SALEM - UNIT 2 XI Amendment No.I97

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'o DEFINITIONS GONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

1.7.1 All penetrations required to be closed during accident conditions are either:

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Capable of being closed by an OPERABLE containment automatic a.

isolation valve systaa, or b.

Closed by manual valves, blind flangen or deactivated estomat4 c valves securi;d in their closed positions, except ac. Lialves that are opened under administrative centrol as permit.ted by Specification 3.6.3.

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1.7.2 All equipment hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, i

I 1.7.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and

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3 1.7.5 The sealing mechanism associated with each penetration (e.g.,

7 welds, bellows or 0-rings) is OPERABLE.

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1.B NOT USED CORE ALTERATION l

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1.9 CORE ALTERATION shall be the movement or manipulation of any component --

within the reactor pressure vessel with the vessel head removed and fuel in I

the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

CRE OPERATING LIMITS REPORT 1

l 1.9a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides core operating limits for the current operating reload cycle.

These cycle-specific core operating limits shall be detenmined for each reload cycle in accordance with Specification 6.9.1.9.

Unit operation within these operating limits is addressed in individual specifications.

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DOSE EOUIVALENT I-131

- t' 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The SALEM - UNIT 2 1-2 Amendment No. j g7 w<.

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2.0 SAFITY LD(ITS AND LIMITING SAPITY SYSTEM SETTINGS 2.1 SAPTTY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant temperature (T.,) shall not exceed the limits shown in Figures 2.1-1 for 4 loop operation.

l APPLICABILITY: M:CES 1 and 2.

ACTION:

I Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizar pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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SALEM - UNIT 2 2-1 Amendment No.197

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FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SALEM - UNIT 2 2-2 Amendment No.197

1 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.1 SAFETY LIMITS

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REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

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APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTICtt:

MES 1 and 2 Whenever the Reactor Coolant System pressure has excoaded 2735 psig, be in NOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

E DES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coclant System pressure to within its limit within 5 minutes.

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l SALEM - UNIT 2 2-3 Amendment No.197

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• l TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRfDENTATION _ TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABI2. VALUES

1. Manual Reactor Trip Not applicable Not applicable

2. Power Range, Neutron Flux Low setpoint - 5 25 % of RATED Low Setroint - s 26% of RATED THERMAL POWER THERMAL POWER High Setpoint

. s 109% of RATED High Setpoint - s 110% of RATED THERMAL POWER THERKE POWER i

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 5.5% of RATED THERMAL 10NER High Positive Rate a time constant 2 2 second with a time constant 2 2 second l

4. Power Range, Neutrora Flux, s 5% of RATED THERMAL POWER with s 5.5% of RATED THEMEAL POWER High Negative Rate a time constant 2 2 second with a time constant 2 2 second

5. Intermediate Range,l Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER Flux I

6. Source Range, Neu on Flux s 105 counts pe second s 1.3 x 10' counta per second

7. Overtemperature AT See Note 1

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See Note 3 l

8. Overpower AT See Note 2 See Note 4

9. Pressurizer Pressure--Low 2 1865 peig a 1855 peig l

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10. Pressurizer Pressure--High s 2385 psig s 2395 peig

11. Pressurizer Water Level--

s 92% of instrument span s 93% instrument span High

12. Loss of Flow 90% of design flow per loop

• 2 89% of design flow per loope 2

' Design flow is 82,$00 gpe per loop.

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SAIRM - UNIT 2 2-5 Amendment l'o.197 1

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TABLE 2.2-1 (Continued)

REACTOR TRIP SYS"'lM INSTRUMENTATION TRIP SETPOINTS NOTATION NOTE 1:

Overtemperature AT s AT (K -K 1+t, S (T-T') +K (P-P')-f (AI)]

y 3

1+t s 2

where:

AT = Indicated AT at RATED THERMAL POWER o

T

= Average temperature, *F

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T' Indicated T., at RATED THERMAL POWER s 577.9'F

=

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f P

= Pressuriser pressure, psig P'

= 2235 psig (indicated RCS nominal operating pressure) 1+t, S The function generated by the lead-lag controller for

=

1+t S T., dynamic compensation 2

a t &1 2

Time constants utilized in the lead-lag controller for

=

T.,

t = 30 secs, 1

• 4 ** **

2 S

= Laplace transform operator, Sec'1

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'A SALEN - UNIT 2 2-7 Amendment No.197 h__-....

TABLE 2. 2-1_(y itinued)

REACTOR TRIP SYSTEM INSTRUMENTJ TION TRIP SEPPOINTS NOTATION (Continued)

Operation with 4 Loops K

1.22

=

2 K.

0.02037

=

K

=

3 0.001020 and fa (AI) is a function of the indicated difference between top and botion detectors of the power-range nuclear ion chambers; with gains to be selecto'.

based on measured instrument response during plant startup tests such that:

f (1) for q,

q. between -23 percent and +13 percent, f (AI) = 0 I

t (where q, and q. are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + q. is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (q, q.)

exceeds

-23 percent, the AT trip setpoint shall be automatically reduced by 1.26 percent of its value at RATED THERMAL POWER.

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7 (iii) for each percent that the magnitude of (q, - q.)

exceeds

+13 percent, the AT trip setpoint shall be automatically reduced by 2.63 percent of its value at RATED THERMAL POWER.

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SALEM - UNIT 2 2-8 Amendment No.197 l

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IABLE,2.2-1 (cc.tinued)

REACTOR TRIP SYS{gM INSTRUMENTATION TRIP SETP)1NTS NC'!ATION 1 Continued) i Note 2: Overpower AT s AT [N-N [ ts3

) T - % (T-T")-f,(AI)]

o 1+t,8 where:

AT =

Indicated AT at RATED THERN POWER o

T =

Average temperature, *F T" = Indicated T.,

at RATED Tk2RMAL POWER s 577.9'F N=

1.09

%=

0.02/'F for increasing average temperature and O for decrearsing average temperature N=

0.00149/*F for T > T"; & = 0 for T s T" l

t.S

=

The function generat..ed by the rate lag controller 1+t 5 for T., dynamic compensation 3

Time constant utilized in the rate lag controller 13

=

j for T.,13 = 10 secs.

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S =

I.aplace transform operator, Sec-1 f (AI) = -O for all AI -

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i Note 3:

l The channel's maximum trip point shall not exceed its computed trip j

point by more than 1.1 percent.

I Note 4: The channel's =mvisua trip point shall not exceed its computed trip point by more than 2.1 percent.

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1 SALEM - UNIT 2 -9 Amendment No.197 m

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2.1 SAFETY LIMITS BASES l

l 2.1.1 REACTOR CORE 1

I The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission

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I products to the re: actor coolant. Overheating of the fuel cladding is prevented i

by restricting fuel operation to within the nucleate boiling regime where the 1

heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

5

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Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and j

therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been j

related to DNB through correlations which have been developed to predict the l

1 f

DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, decided as the ratio of l

the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

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The DNB design basis is as follows: uncertainties in the WRB-1 and WRB-2 correlations, plant operating parameters, nuclear and thermal parameters, fuel

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j fabrication parameters, and computer codes are considered statistically such that there is at least a, 95 percent probability with 95 percent confidence level that DNBR will not occur on the most limiting fuel rod during Condition I and II events. This establishes a design DNBR value which must be met in plant safety analyses using values of input Parameters without uncertainties.

The curves of Figure '2.1-1 shows the loci of points of THERMAL POWER, l----

Reactor Coolant System pressure and-average temperature for which the minimum DNBR is no less than the design DNBR value, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

The curves are based on an enthalpy hot channel factor, F""as and l

a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in F*'as at reduced power based on the expression:

f, = F"",,

[1. 0 + PF., (1.0 - P)]

a Where: F"",, is the limit at RATED THERMAL POWER in the Core Operating Limits Report (COLR). _ _ _ _. _

-,-I PF n is the Power Factor Multiplier for fa, specified in the COLR, and g

P is THERMAL POWER a

RATED THERMAL POWER I

These limiting heat flux conditions are higher than those calculated for the range of all control rod positions from FULLY WITHDRAWN to l

s the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f (delta I) function of the Overtemperature trip. When the axial power SALEM - UNIT 2 B 2-1 Amendment No.197

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LIMITING SAFETY SYSTEM SETTINGS 4

t BASES 1

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Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint'does not require reactor protection system setpoint modification because the P-B setpoint and associated trip will prevent DNB during 3 loop

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operation exclusive of the overtemperature delta T setpoint. Three loop ji operation above the 4 loop P-8 has not been evaluated and is not permitted.

l 1

1 Overoower Delta T The Overpower delta T reactor tirip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required

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range for Overtemperature delta T protection, and provides a backup to the

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High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of wau e with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors. No credit was taken for operation of this trip in the accident analyses / however, its functional capability at the specified trip setting is requiru! by this specification to enhance the overall reliability of the Reactc c Protection Systea.

l Pressurizer Pressure The Pressurizer High and I,ow Pressure trips are provided. to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressu: e for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor

.I Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overell reliability of the Reactor Protection System.

l SALEM - UNIT 2 B 2-5 Amendment No.197

9 LDfITING SAITTY SYSTEM SETTINGS I

BASES i

Loss of Flow 4

4

• The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

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Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 90% of nominal full loop if flow. Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop 5 '

flow. This latter trip will prevent the minimum value of the DNBR from going below the design DNBR value during normal operational transients.

Steam Generator Water Level The Steam Generator Water I4 vel Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there will be sufficient water inventory in the steam

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generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

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SALEM - UNIT 2 B 2-6 Amendment No.j g7

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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTRQQ

)

SHUTDOWN MARGIN - T,y > 200*F LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN HARGIN shall be greater than or equal to 1.3% delta k/k.

l APPLICABILITY: HODES 1, 2*,

3, and 4.

..f ACTION:

~

With the SHUTDOWN MARGIN less than 1.3% delta k/k, immediately initiate and l

continue boration at 233 gpa of a solution containing a 6,560 ppa boron or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE REQUIREMENTS 1

4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.3% delta k/k:

l Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af ter detection of an inoperable control rod (s) and at a.

least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovsble or untrippable, the above required SHUTDOWN HARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s).

b.

When in ICDE 1 or H3DE 2 with Net greater than or equal to 1.0, at least once per,12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawalds within the limits in the COLR per Specification 3.1.3.5.

I When in MODE 2 with Nrr less than 1.0, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to c.

achieving reactor criticality by verifying that the predicted critical control rod position is within the limits in the COLR per l

Specification 3.1.3.5.

1

u. m.... -

,n

~

"~~v

• See Special Test Exception 3.10.1 SALEM - UNIT 2 3/4 1-1 Amendment No.197

w-

.- ----. ~ ~ - -

REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit in the COLR per l

Specification 3.1.3.5.

When in ) ODES 3 or 4, 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 consideration of

-g e.

the following factors:

1.

Reactor coolant system boron concentration, g

l

. 7 2.

t Control rod position, 3.

Reactor coolant system average temperature, 3

4.

Fuel burnup based on gross thermal energy generation, i

i i

5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted, values to demonstrate agreement.within i 1% delta k/k at least once per 31 Effective Full Power Days (EFPD). This comparison shall consider at least 4

those f actors stated in Specification 4.1.1.1.1.e, above. The predicted l

reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

i 1

_m.

7~---'

~

' ~ ~ ~ ~'

SALEM - UNIT 2 3/4 1-2 Amendment No.197

>+

REACTIVITY CONTROL SYSTEMS HODERATOR TEMPERATURE COEPTICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATIi4G LIMITS REPORT (COLR). The mavimum upper limit shall be less positive than or equal to 0 ak/k/*F.

APPLICABILITY: Beginning of Cycle Life (BOL) Limit - W OES 1 and 2* only#

End of Cycle Life (EOL) TAmi t - HODES 1, 2 and 3 only#

ACTION:

With the MTC more positive than the BOL limit specified in the COLR, l

a.

operations in HODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than j

the BOL limit specified in the COLR within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in l

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 in the COLR per j

Specifiention 3.1.3.5.

s 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.

3.

In lieu of any other report required by Specification 6.9.1,-a - -------

Special Report is prepared and submitted to the Commission pursuant to specification 6.9.2 within 10 days, describing the value of the measured HTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive HTC to within its limit for the all rods withdrawn condition.

i b.

With the MTC more negative than the EOL limit specified in the COLR, l

be in NOT SHU'IDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

~

P-

• With Ntr greater than or equal to 1.0

.x.2 c... Je -n ' N.-

"A

1. see special Test Exception 3.10.3 SALEM - UNIT 2 3/4 1-4 Amendment No 3 97

REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT SURVEILLANCE REQUIREMENTS 4.1.1.3 The HTC shall be determined to be within its limits during each fuel cycle as follows:

The HTC shall be measured and compared to the BOL limit specified in l

a.

the COLR prior to initial operation above 5% of RATED THERMAL POWER, I

after each fuel loading.

b.

h' The HTC shall be measured at any THERMAL POWER and compared to the 300 ppa surveillance limit specified in the COLR (all rods l

1 withdrawn, RATED THERMAL POWER condition) within 7 EFPD after

{

reaching an equilibrium boron concentration of 300 ppa. In the event this comparison indicates the HTC is more negative than the 300 ppa surveillance limit specified in the COLR, the HTC shall be l

remeasured, and compared to the EOL HTC limit specified in the COLR I

at least once per 14 EFPD during the remainder of the fuel cycle.

a-

... T~..

SALEM - UNIT 2 3/4 1-5 Amendment No.197

~~

~

e REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT LIMITING CONDITION FOR OPERATION

==

3.1.3.1 All full length (shutdown and control) rods, shall be OPERABLE and positioned within i 18 steps (indicated position) when reactor power is s 85%

RATED THERMAL POWER, or i 12 steps (indicated position) when reactor power is

> 85% RATED THERMAL POWER, of their group step counter demand position within one hour after rod motion.

APPLICABILITY: 3CDES 1* and 2*

ACTION:

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

as a result of excessive friction or mechanical interference or known to be untrippabla, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within I hour 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 mis-aligned from the group step counter demand position by more than i 18 steps (indicated position) at s 85% RATED THERMAL POWER or i 12 steps (indicated position) at > 85% RATED THERMAL POWER, be in HOT STANDBI 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 inoperable due to causes other than c.

addressed by ACTION a, above, or mis-aligned from its group step counter demand position by more than i 18 steps (indicated position)

~ ~ ' ~

at s 85% RATED THERMAL POWER or i 12 steps (indicated position) at

> 85% RATED THERMAL POWER, 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 remainder of the rods in the bank with the inoperable rod are aligncd to within i 18 steps (indicated position) at s 85%

RATED THERMAL POWER or i 12 steps (indicated position) at >85%

RATED THERMAL POWER, of the inoperable rod while maintaining the rod sequence and insertion limits in the COLR per Specification 3.1.3.5. The THERMAL POWER level shall J>e --

7-restricted pursuant to specification 3.1.3.5 during subsequent operation, or 3.

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:

• See Special Test Exceptions 3.10.2 and 3.10.3.

SALEM - UNIT 2 3/4 1-13 Amendment No.197

e REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.5 The control banks shall be limited in physical insertien as specified l

in the CORE OPERATING LIMITS REPORT (COLR).

l APPLICABILITY: E DES 1*, and 2*#

ACTION:

k' With the control banks ireerted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, either:

Restore the control banks to within the limits within two hours, or a.

b.

Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank position using the insertion limits specified in the COLR, or l

Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

SURVEILLANCE REQUIREMENTS i

4.1.3.5 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by use of the group demand counters and verified by the analog rod position indicators *= except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod 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 /> **. -

• See Special Test Exceptions 3.10.2 and 3.10.3

• *For power levels below 50% one hour thermal " soak time" is permitted.

During this soak time, the absolute value of rod motion is limited to six steps.

1. With Keff greater than or equal to 1.0

-'~~-----~~'m

-- ~ ~ ~ - - ' ~. ~

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~ " '~~ & ?:~

- " '~

' ' " ~

^~;"'~

1

)

2 SALEM - UNIT 2 3/4 1-20 Amendment No.197

e THIS PAGE INTENTIONALLY IMT BIANK

. 3. 44,

4...

.,4....2.,..,,.,%...em s.

,s.,4. -,-..~<

- -L:..w

.. 4

.~.,...o_....

.,. ~

SALEM - UNIT 2 3/4 1-21 Amendment No.197

a

'O I

l l

THIS PAGE INTENTIONALLY LEFT BIANK 1

(

--..-.- l l

l t

i

-q-

- - - - - ~

SALEM - UNIT 2 3/4 1-22 Amendment No.197

---

a==

/

3/4.2 POWER DISTRIBUTION LIMITS 2/4,2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE shall be maintained within the target band about the target flux difference as specified in the CORE OPERATING LIMITS REPORT (CORL).

-~

APPLICABILITY: NODE 1 ABOVE 50% RATED THERMAL POWER

• ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the a.

target band about the target flux difference as specified in the COLR and with THERMAL POWER:

1.

Above 90% of RATED THERMAL POWER, within 15 minutes:

a)

Either restore the indicated AFD to within the target band limits, or b)

Reduce THERMAL POWER to less than 90% of RATED THERMAL POWER.

2.

Between 50% and 90% of RATED THERMAL POWER:

a)

POWER OPERATION may continue provided:

1)

The indicated AFD has not been outside of the target band as specified in the COLR for more than l' --~

' - ~

hour penalty deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 2)

The indicated AFD is within the limits as specified in the COLR. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b)

Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification -

~~-~~~~~T^

4.3.1.1.1 provided the indicated AFD is maintained

~"*'T-

- within the limits as specified in the COLR. A total of 16 l

hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

~*See Special Test Exception 3.10.2 SALEM - UNIT 2 3/4 2-1 Amendment No.19.7

l i

i i

l POWER DISTRIBUTION LIMITS I,IMITING CONDITION FOR OPERATION (Continued) b.

THERMAI, POWER shall not be increased above 90% of RATED THERMAL i

s POWER unless the indicated AFD is within the target band as specified in the COLR and ACTION 2.a) 1), above has been satisfied.

I THERMAL POWER shall not be increased above 50% of RATED THERMAL c.

POWER unless the indicated AFD has not been outside of the l

h target band as specified in the COLR for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty I

deviation cumulative during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Power increases above 50% of RATED THERMAL POWER do not require being within the target band provided the accumulative penalty deviation is not violated.

SURVEILLANCE RIQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

Monitoring the indicated AFD for each OPERABLE excore channel:

a.

l 1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for

^ ~ - - -

each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when l

at least 2 or more OPERABLE excore channels are indicating the AFD to be outside the target band.

Penalty deviation outside of the target band shall I

be accumulated on a time basis of:

a.

One minute penalty deviation for each one minute of POWER

~a

~ ~ - - ~

=

b OPERATION outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

~

b.

One-half minute penalty deviation for each one minute of POWER OPERATION outside of the target band at THERMAL POWER levels below 50% of RATED THERMAL POWER.

SALEM - UNIT 2 3/4 2-2 Amendment No.197

~ _ _ _.

\\

l i

THIS PAGE INTENTIONALLY I. EFT BLANK i

l l

1 1

i I

i 1

4 l

...w..

=.

.- n -. - -

Gs z,.

_.4 1,

4~--+ % + - c - e -

-6 SALEM - UNIT 2 3/4 2-4 Amendment No.197

i POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - FJE)

LIMITING CONDITION FOR OPERATION 3.2.2 Fg (a) $

• K(s) for P >

0.5, and y

i

'}

~

Fe (8) S M

• E(s) for P >

0.5, and

~

(. 5 i

Where F " = the F limit at RATED THERMAL POWER (RTP) specified in g

g 7'

the CORE OPERATING LIMITS REPORT (COLR),

P

=

THERMAL POWER

, and RATED THERMAL POWER K(s) = the normalized F (s) as a function of core height g

as specified in the COLR.

APPLICABILITY: ICDE 1 ACTION:

With F (E) exceeding its limit:

g j.__..-...

Reduce THERMAL POWER at least 1% for each 1%. F (E) exceeds the.

a.

o 1

Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION any proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER f

OPERATION may proceed provided the Overpower delta T Trip Setpoints

)~

have been reduced at least 14 for each 1% F (E) exceeds the limit.

a The overpower delta T Trip setpoint reduction shall be performed

}

with the reactor in at least NOT STANDBY.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required

___. by a. above;. THERMAL POWER may then be increased provided F (E) is -

g demonstrated through incore mapping to be within its limit.,

y--

+

Amendment No.197 SALEM - UNIT 2 3/4 2-5

s 1

POWER DISTRIBUTION LIMITS i

SURVEIIIANCE REQUIRDENTS 1

4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F,y shall be evaluated to determine if F (Z) is within its limit by:

4 a

I Using the novable incore detectors to obtain a power a.

distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Increasing the measured F,y component of the power distribution I

map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

Comparing the F,y computed (PC,y) obtained in b, above to:

c.

1.

The F,y limits for RATED THERMAL POWER (E*",y ) for the appropriate measured core planns given in e. and f.,

below, and 2.

The relationship:

F'n = F"",

(1 + PF,(1-P) )

l where F',y is the limit for fractienal THERMAL POWER operation expressed as a function of F"",y, PF, is the power factor multiplier for F in the CORL, and P is

~ ~~' -~~-

n the fraction of RATED THERMAL POWER at which F,y was measured.

d.

Rameasuring F,y according to the following schedule:

1.

When Fe,y is greater than the F"",y limit for the apprbpriate measured core plane but less than the F',y relationship, additional power distribution maps shall be taken and Fe,y compared to P*",,

and F',y

,)--~11ther within ~24 hours af ter exceeding by 20% of RATED' ~

-~ ~ ~i '-

THERMAL POWER or greater, the THERMAL POWER at which PC,y was last determined, or SALEM - UNIT 2 3/4 2-6 Amendment No.197

4 l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b)

At least once per 31 EFPD, whichever occurs first.

2.

When the F, is less than or equal to the F*", limit for the C

appropriate measured core plane, additional power distribution

[

maps shall be taken and f*y compared to 7*", and F( at I

least once per 31 EFPD.

e.

The F limit for Rated Thermal Power (f*"n) shall be provided for n

all core planes containing bank "D" control rods and all unrodded core planes in the COLR per specification 6.9.1.9.

g f.

The F limits of e., above, are not applicable in the following n

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 % i 2 %, 32.1 % i 2 %, 4 6. 4 % i 2 %,

60.6% i 2% and 74.9% i 2%, inclusive.

~-

~-

4.

Core plane regions within i 2% of core height (1 2.88 inches) about the bank demand position of the bank "D" centrol rods.

g.

Evaluating the effects of F on F (Z) to determine if F (Z) is y

g o

within its limit whenever I*, exceeds F*y.

4.2.2.3 When F,(2) is measured pursuant to specification 4.10.2.2, an overall asasured F (Z) shall be obtained from a power distribution map and increased o

i

)

by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

SALEM - UNIT 2 3/4 2-7 Amendment No.197

THIS PAGE INTENTIONALI.Y I. EFT WM I

[

i F

9 1

j l

~ i r

l v

...m;...,.

1 l

t i

i I

l SALEM - ITnIT 2 3/4 2-8 Amendment No.197 4

2

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR P".

LIMITING CONDITION FOR OPERATION 3.2.3 f

shall be limited by the following relationship:

u fu "' I ", [1.0 + PF, (1.0 - P)]

Where 7", is the limit at RATED THERMAL POWER in the Core Operating Limits Report (COLR).

PF, is the Power Factor Multiplier for F"a specified in thE f.cLR, and P is THERMAL POWER RATED THERMAL POWER APPL 7F* ABILITY:

HDDE 1 ACTh0N:

With 7*, exceeding its limit:

Reduce THERMAL POWER to less tAaz. 50% of RATED THERMAL POWER within a.

I 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power h w,e Neutron Flux-High Trip Setpoints to 5 55% of RATED THERMAL '.Owi.x within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Demonstrate thra in-cor s mapping that 7*, is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter exceedi'.g the 11. sit er reduce THERMAL POWER to less than 5% of RATED THERLAL POWER within the next.2 Aours,- and -_--- - - -- -~ !

Identify and correct the cause of the out of limit cendition prior c.

to increasing TH'.RMAL POWER above the reduced lir.it required by a.

or b. above; sv' sequent POWER OPERATION may proceed provided that a

F*, is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THEPMAL POWER, at a nominal *15% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95%

or greater RATED THERMAL POWER.

f_ -

--. L

---

j-l m_.,.._

,-.m..e

.v.

~

SALEM - UNIT 2 3/4 2-9 Amendment No.197 b

e i

TABLE 3.2-1 DNB PARAMETERS PARAMETER LIMITS 4 Loops in

[I Operation

)f Reactor Coolant System T.,

s 582.9'F l

4 Pressurizer Pressure a 2200 psia l

j J

Reactor Coolant Systram Total Flow Rate a 341,000 gpa l

i l

j i

i 4

e

-l Ei 1

\\

l 4

4 j

g.

Limit not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% RATED THERMAL POWER.

1. Includes a 2.4% flow uncertainty plus a 0.1% measurement uncertainty l

due to feedwater venturi fouling.

SALEM - UNIT 2 3/4 2-17 Amend. ment No.197

-.. ~... -

l 3/4.1 REACTIVITY CONTROL SYSTEMS i

4 1

BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN i'

A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients jj associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently suberitical to preclude inadvertent criticality in the shutdown condition.

Ig F'

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T.n. The most restrictive condition occurs at EOL, with T., at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% Ak/k it initially required to control the reactivity l*

transient. Accordingly, *Ao SWTDOWN MARGIN requirement is based upon this limiting condition and its censistent with FSAR safety analysis assumptions.

With T., less than or e@cl. to 200*F, the reactivity transients resulting from a, postulated steam line break cooldown are minimal and a in Ak/k shutdown margin provides adequate protection.

3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (MTC)

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.

i

~

~

~.

1

. - ~ -

a - e SALEM - UNIT 2 B 3/4 1-1 Amendment No.197

/

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (MTC) (Continued) 4 The MTC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MTC values at conditions other g

than those explicit 3y stated will require extrapolation to those conditions in order to permit an accurate comparison.

ti 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 analysis to nominal operating conditions. These corrections involved: (1) a conversion of the MDC used in the FSAR analysis to its equivalent MTC, based on the rate of change of moderator density with temperature at RATED THERMAL POWER conditions, and (2) subtracting from this value the largest differences in MTC observed between EOL, all rods withdrawn, PATED THERMAL POWER conditions, and those most adverse conditions of moderator temperature and pressure, rod insertion, axial power skewing, and xenon concentration that can occur in normal operation and lead to a significantly more negative EOL MTC at RATED THERMAL POWER.

These corrections transformed the MDC value used in the FSAR analysis into the limiting End Of Cycle Life (EOL) HTC value. The 300 ppa surveillance limit MTC value represents a conservative value at a core condition of 300 ppa equilibrium boron concentration that is obtained by correcting the limiting EOL MTC for burnup and born concentration.

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.

3/4.1.1.4 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 allowable setpoint, 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its

_ _ _,.. _ _ _. _ minimum RT,,n temperature.

3 --

t

-. _., u - 4.- 4.-

-m -e w-SALEM - UNIT 2 B 3/4 1-2 Amendment No.197

REACTIVITY CONTROL SYSTEgg BASES 3/4,1.2 BORATION 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)

I.

separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature 2 350*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an A

assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay and cooldown l

to 200*F.

The maximum expected boration capability (minimum boration volume) requirement is established to conservatively bound expected operating conditions throughout core operating life. The analysis assumes that the most reactive control rod is not inserted into the core. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires borated water from a boric acid tank in accordance with TS Figure 3.1-3, and additional makeup from either:

(1) the second boric acid tank and/or batching, or (2) a maximum of 41,000 gallons of 2,300 ppm borated water from the refueling water storage tank. With the refueling water storage tank as the only borated water source, a maximum of 73,000 gallons of 2,300 ppm borated water is required. However, to be consistent with the ECCS requirements, the RWST is required to have a minimum contained volume of 350,000 gallons during operations in HODES 1, 2, 3 and 4.

The boric acid tanks, pumps, valves, and piping contain a boric acid solution

^-~

concentration of between 3.75% and 4% by weight. To ensure that the boric acid remains in solution, the tank fluid temperature and the process pipe wall temperatures are monitored to ensure a temperature of 63*F, or above is maintained. The tank fluid and pipe wall temperatures are monitored in the main control room. A 5'F margin is provided to ensure the boron will not precipitate out.

Should ambient temperature decrease below 63*F, the boric _ acid tank heaters, in conjunction with boric acid pump recirculation, are capable of maintaining the boric acid in the tank and in the pump at or about 63*F.

A small amount of boric acid in the flowpath between the boric acid recirculation line and the suction line to the charging pump will precipitate out,-but it will not ~

~

cause flow blockage even with temperatures below 50*F.

s..,

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-2 Wi e the RCS temperature below 350*F, one injection system is acceptable without single failure consideration on the basis of the stable resetivity condition of the reactor and the additional restrictions prohibiting CORE OPERATIONS and positive reactivity change in the event the single injection

{

system becomes inoperable.

SALEM - UNIT 2 B 3/4 1-3 Amendment No.197 l

f

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c.

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3/4,2 POWER DISTRIBUTION LIMITS i

BASES i

3 i

The specifications of this section provide assurance of fuel integrity

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during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) l events by: (a) 1-meeting the DNB Design Criteria during normal operation and in l

short term transients, and (b) limiting the fission gas release, fuel pellet l

temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition -

)

l I events provides assurance that the initial conditions assumed for the IACA j

analyses are met and the ECCS acceptance criteria limit of 2200'F is not 1

exceeded.

4 i

The definitions of hot channel factors as used in these specifications are as follows:

i F (Z)

Heat Flux Hot Channel Factor, is defined as the unwimum local heat e

flux on the surface of a fuel rod at core elevation E divided by the j

average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

F*'a Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

Fy(Z)

Radial Peaking Factor is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

_ _ _ _.. - ~ _ -

The limits on AXIAL FLUX DIFFERENCE assure that the Fe(E) upper bound envelope of the F limit specified in the CORE OPERATING LIMITS REPORT (COLR) l g

times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium menon conditions with the part length control rods withdrawn from the core. 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 staat/ state operation at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER

- -is the target flux difference at RATED THERMAL POWER for the associated core e

burnup donditions. Target flux differences for other THERMAL POWER levels are

_"~

obtained by multiplying the RATED THERMAL POWER value by the appropriate " k

--b-fractional THERMAL POWER level.

The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

SALEM - UNIT 2 B 3/4 2-1 Amendment No. 197

W'

_ - _ - ~.. _..

POWER DISTRIBUTION LIMITS BASES Although it is intended that the plant will be operated with the AXIAL FI.UX DIFFIRENCE within the target band in the COLR per Specification 3.2.1 l

about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL PC fER levels. This deviation will not affect the xenon redistribution sufUciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly, a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> penalty deviation limit cumulative during the g~

previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided for operation outside of the target band but Y,

within the limits specified in the COLR while at THERMAL POWER levels between l

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 penalty of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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. An alarm is received any time the auctioneered high flux difference exceeds the target band limits. Time outside the target band is graphically presented on the strip chart.

Figure B 3/4 2-1 shows a typical monthly target band.

~~

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SALEM - UNIT 2 B 3/4 2-2 Amendment No.197

INFORMATION ONLY*

1 4

1 1

Percent of Rated Thermal Power 1

4) i-10 0 %

l 90%

80%

70 %

<G-1....._.

40%

30%

i 0

-20%

-10%

0 10 %

20%

INDICATED AXIAL FLUX DIFFERENCE i

Figure B SM 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER ___

....-~..--.nn..

-~ _ ;

Lz,_ g.,_ y y._

• REFER TO COLR FIGURE 2 FOR ACTUAL LIMITS SALEM - UNIT 2 B 3/4 2-3 Amendment No.197

I' POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTRALPY HOT CHANNEL AND RADIAL PEAKING FACTORS - F;(Z) AND F*u i

The limits on heat flux and nuclear enthalpy hot channel factors and RCS i

flow rate ensure that ?.s the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad l

temperature will no'; exceed the 22OO'F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable but will normally only (b

be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

l Control rod in a single group move together with no individual rod a.

insertion differing from the group demand position by more than the allowed rod misalignment.

j 4

4 b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

J 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 1

DIFFERENCE, is maintained within the limits.

The relaxation in F*, as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

F"a will be

- -- 1 maintained within its limits provided conditions a through d above, are 4

maintained.

)

When an Fa measurement is taken, both experimental error and manufacturing tolerance must be allowed for.

Five percent is the appropriate allowance for a full core map taken with the incore detector flux mapping i

system and 3% is the appropriate allowance for manufacturing tolerance.

When F*, is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. The specified limit for 3*, also contains an 8% allowance for

_ uncertainties which mean that normal operation will result in F*a sP"" n/1.OB.

-- p Where F,s is the limit at RATED THERMAL POWER (RTP) specified in the CORE g

OPERATING LIMITS REPORT (COLR). The St allowance is based on the foll'owing '

considerations :

SALEM - UNIT 2 B 3/4 2-4 Amendment No.197

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL AND RADIAL PEAKING FACTCRS - F (Z) AND fan (Continued) g abnormal perturbations in the radial power shape, such as from rod a.

misalignment, effect f, more directly than F.

a g

t b.

although rod movement has a direct influence upon limiting F g to within its limit, such control is not readily available to limit f n, and

+

a errors in prediction for control power shape detected during startup c.

physics test can be compensated for in F by restricting axial flux g

dis tributions. This compensation for fan is less rapidly available.

The radial peaking factor F,y(Z) is measured periodically to provide assurance that the hot channel factor F (Z), remains within its limit. The F limit g

y for RATED THERMAL POWER F"",y as provided in COLR per specification 6.9.1.9, was determined from expected power control maneuvers over the full range of burnup conditions in the core. -

3/4.2.4 OUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

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~4 SALEM - UNIT 2 B 3/4 2-5 Amendment No.197

_c._

3/4.4 REACTOR COOLANT SYSTEM RASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, meet the DNB design criteria during all normal operations and l

anticipated transients.

In MODES 1 and 2 with less than all coolant loops in operation, this specification requires that the plant be in at least EOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal k

for removing decay heats but, single failure considerations require all loops be in operation whenever the rod control system is energized and at least one 3

loop be in operation when the rod control system is doenergized.

1 In MODE 4, a single reactor coolant loop or RER loop provides sufficient heat removal for removing decay heats but, single failure considerations require that at least 2 loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requires that two RER loops be OPERABLE.

In MODE 5, single failure considerations require that two RER loops be OPERABLE. For support systems: Service Water (SW) and Component Cooling (CC),

component redundancy is necessary to enst'.re no single active component f ailure will cause the loss of Decay Heat Removal. One piping path of SW and CC is adequate when it supports both RHR loops. The support systeam needed before entering into the desired configuration (e.g., one service water loop out for maintenance in Modes 5 and 6) are controlled by procedures, and include the following:

• A requirement that two RKR, two CC and two SW pumps, powered from two different vital buses be kept operable

• A listing of the active (air / motor operated) valves in the affected flow path to be locked open or disabled Note that four filled reactor coolant loops, with at least two steam!'

~ ~ ~ ~ ~

generators with at least their secondary side water level greater than or equal to 5% (narrow range), may be substituted for one residual heat removal loop. This ensures that a single failure does not cause a loss of decay heat removal.

The operation of one Reactor Coolant Pump or one RER Pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during Boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with Boron concentration reductions will, therefore, be within the capability of operator I

recognition and control.

d 9

The restrictions on starting a Reactor Coolant Pump below P-7 with one or h ~ ~ more RCS cold legs less than or equal to 312*F are provided to prevent RCS

' ~ ~ ~"

pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10CFR Part 50.

The RCS will be l

protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer (thereby providing a volume into which the primary coolant can expand, or (2) by restricting the starting of Reactor Coolant Pumps to those times when i

secondary water temperature in each steam generator is less than 50*F above

{

each of the RCS cold leg tamperatures.

l SALEM - UNIT 2 3 3/4 4-1 Amendment No.197

' - = = = - = = = - - - ~ ~

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment is designed and shall be maintained for a maximum internal pressure of 47 psig. Containment air temperatures up to 351.3*F are acceptable providing the containment pressure is in accordance with that described in the UFSAR.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 193 fuel assemblies.

Each assembly shall consist of a matrix of zircaloy or EIRLO clad fuel rods with an initial composition of natural or slightly enriched uranium dioxide as fuel material.

Limited substitutions of zirconium a611oy or stainless steel filler rods for fuel rods, in accordance with NRC-approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.

A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

CONTROL ROD 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 percant cadmium.

All control rods shall be clad with stainless steel tubing.

5.4 RIACTOR COOLANT SYSTEM DESIGN FEATURE AND TEMPERATURE The reactor coolant system is designed and shall be maintained:

5.4.1 In accordance with the code requirement specified in Section 4.1 a.

of the FSAR, with allowance for normal degradation pursuant to the applicable surveillance Requirements,

~

b.

_ For a pressure of 2485_ psig, and. _

For a temperature of 650*F, except for the pressurizer which c.

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is 680*F.

VOLUME

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5.4.2 The total water and steam volume of the reactor coolant system is 1

12,446 1 426 cubic feet at a nominal T., of 573.0*F.

l SALEM - UNIT 2 5-4 Amendment No.197

]

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ADMINISTRATIVE CONTROLS d.

Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

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

f.

Solidification agent or absorbent (e.g., cement, urea forealdehyde).

The Radioactive Effluent Release Reports shall include a list of descriptions of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during

(

the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

6.9.1.9 CORE OPERATING LIMITS REPORT (COLR) a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLE for the following:

1.

Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 2.

Control Bank Insertion Limits for Specification 3/4.1.3.5,

^

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~ ~ ~ 3.

Axial Flux Difference Limits and target band for Specification

~

~~

3/4.2.1, 4.

Heat Flux Hot Channel Factor, F, its variation with core g

height, K(z), a.d Po.i9r Factor Multiplier PF Specification y,

3/4.2.2, and 5.

Nuclear Enthalpy Hot Channel Factor, and Power Factor Multiplier, PFa for Specification 3/4.2.3.

b.

The analytical methods used to determine the core operating.

>ts shall be those previously reviewed and approved by the NRC,

~~~

specifically those described in the following documents:

~~

~

E~

1.

WCAP-9272-P-A, Westinchouse Reload Safety Evaluation Methodoloerv, July 1985 (W Proprietary), Methodology for Specifications listed in 6.9.1.9 a. Approved by safety Evaluation dated May 28, 1985.

SALEM - UNIT 2 6-24 Amendment No.197

ADMINISTRATIVE CONTROLS 2.

WCAP-8385, Power Distribution Control and Load Followina Procedures - Topical Report, September 1974 (W Proprietary)

Methodology for Specification 3/4.2.1 Axial Flux Difference Approved by Safety Evaluation dated January 31, 1978.

3.

WCAP-10054-P-A, Rev.1, Westinchouse Small Break ECCS Evaluation Model Usinc NOTRUMP Code, August 1985 E l

Proprietary), Methodology for Specification 3/4.2.2 Heat Flux

'3 Hot Channel Factor. Approved for Salem by NRC letter dated August 25, 1993.

I 4.

WCAP-10266-P-A, Rev. 2, The 1981 Version of Westinchouse Evaluation Model Usina BASH Code Rev. 2. March 1987 3 Proprietary) Hethodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved by Safety Evaluation dated November 13, 1986, c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, j

nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any mid-cycle revisions or supplements shall be provided upon issuance for each reload cycle to the NRC.

j SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Document Control Desk, Washington, D.C. 20555, with a copy to the

_ Administrator, USNRC Region I within the time period specified for each report.

6.9.3 Violations of the requirements of the fire protection program described in the Updated Final Safety Analysir Report which would have adversely affected the ability to achieve and maintain safe shutdown in the event of a fire shall be submitted to the U. S. Nuclear Regulatory Commission, Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator of the Regional Office of the NRC via the I.icensee Event Report System within 30 days.

6.9.4 When a report is required by ACTION 8 OR 9 of Table 3.3-11 " Accident Monitoring Instrumentation", a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of '

g

_ monitoring for inadequate core cooling, the cause of the inoperability,-and cc the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 2 6-24a Amendment No.197 i

j 1

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