Text

Amend 201 to License DPR-70,consisting of Changes to TSs
	ML20203C705
Person / Time
Site:	Salem
Issue date:	11/26/1997
From:	Stolz J; Office of Nuclear Reactor Regulation
To:	; Public Service Electric & Gas Co, Philadelphia Electric Co, Delmarva Power & Light Co, Atlantic City Electric Co
Shared Package
ML20203C709	List: ML20203C705; ML20203C722;
References
	DPR-70-A-201 NUDOCS 9712150397
	Download: ML20203C705 (45)
	v • d • e

..-. -

~

o meo p

i UNITED STATES

[

NUCLEAR REGULATORY COMMISNON WASHINGTON, D.C. SeteH001 g

(

PUBLIC SERVICE ELECTRIC & GAS COMPANY

,o PHILADELPHIA ELECTRIC COM" Nil DELMARVA POWER AND LIGHT COMPANY i -

ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-272 SALEM NUCLEAR GENERATING STMION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 201 License No. DPR-70 1.

The Nuclear Regulatory Comission (the Comission 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 Enerav Act of 1954, as amended (the Act), and the Comission's rules and.egulations 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 Comission; C.

There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) ion's regulations set forth in 10 CFR that such activities will be conducted in compliance with the Comiss Chapter I; J

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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-70 is hereby amended to read as follows:

J 000 72

--rvw a

e v---

r,v.-

j 2-(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Ap>endices A and B, as revised through Amendment No. 201. are here)y incorporated in the license.

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

3.

This license amendment 1: effective as of its date of issuance, to be implemented prior to entry into Mode 2 from the current outage.

FOR THE NUCLEAR REGULATORY COMMISSION oh F. Stolz, Direc r ject Directorate

-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: November 26, 1997

AIJACHMENT 10 LICENSf AMENDMEffT NO. 201 FACILITY OPESATlWG LICENSE NO. DPR-70 QQQ(ET NO. 50-272 Revise Appandix A as follows:

Remove Panes Insert Panes I

I 11 11 1-2 1-2 2-1 2-1 2-2 2-2 0-3 2-3

)

2-5 2-5 2-8 2-8 2-9

.2-9 B 2-1 B 2-1 8 2-2 B 2-2 B 2-5 8 2-5 B 2-6 B 2-6 3/4 1-1 3/4 1-1 3/4 1-2 3/4 1-2 3/4 1-5 3/4 1-5 3/4 1-5a 3/4 1-Sa 3/4 1-18 3/4 1-18 3/4 1-23 3/4 1-23 3/4 1-24 3/4 1-24 3/4 1-25 3/4 1-25 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-B 3/4 2-9 3/4 2-9 3/4 2-14 3/4 2-14 8 3/4 1-1 B 3/4 1-1 B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 8 ?/4-2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B-3/4 2-3 B 3/4 2-3 B 3/4 2-4 B 3/4 2-4 8 3/4 2-5 8 3/4 2-5 B 3/4 4-1 B 3/4 4-1 5-5 5-5 6-24.

6-24 6 24a

INDEX DE72NITIONS SECTION PAGE L0 DEFINITIONS DEFINED TERMS 1-1 ACTION.

1-1 AXIAL FLUX DIFFERENCE 1-1 CHANNEL CALIBRATION 1-1 CHANNEL CHECK 1-1 CHANNEL FUNCTIONAL TEST 1-1 CONTAINMENT INTEGRITY 1-2 CORE ALTERATION 1-2 CORE OPERATING LIMITS REDORT 1-2 DOSE EQUIVALENT I 131 1-2 E AVERAGE DISINTEGRATION ENERGi*

1-3 ENGINEERED SAFETY FEATURE RESPONSE TIME 1-3 FREQUENCY NOTATION.

1-3 FULLY WITHDRAWN 1-3 GASEOUS RADWASTE TREATMENT SYSTEM 1-3 IDENTIFIED LEAKAGE.

1-3 MEMBER (S ) OF THE PUBLIC 1-4 OFFSITE DOSE CALCULATION MANUAL (ODCM) 1-4 OPERABLE - OPERABILITY.

1-4 OPERATIONAL MODE 14 PHYSICS TESTS 1-5 PRESSURE BOUNDARY LEAKAGE 1-5 PROCESS CONTROL PROGRAM (PCP) 1-5 PURGE-PURGING 1-5 QUADRANT POWER TILT RATIO 1-5 RATED THERMAL POWER 1-5 REACTOR TRIP SYSTEM RESPONLE TIME 1-6 REPORTABLE EVFNT,

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 LEAKAGE 1-7 UNRESTRICTED AREA 1-7 VENTILATION EXHAUST TREATMENT SYSTEM.

1-7 VENTING 1-7 o

SALEM - UNIT 1 I

Amendment No. 201 1

(

1H2EX SAFETY LIMITS AND LIMITIN3 SAFETY SYSTEM SETTINGS SECTION P]Ce,E 2.1 SAFETY LIMITS 1

Reactor Core.

2-1 I

Reactor Coolant System Pressure 2-3 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip System Instrumentation Setpoints 2-4 BASES EEFTION FAQE 2.1 SAFETY LIMITS Reactor Core.

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

.... B 2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Trip Syotem Instrumentation Setpoints

. B 2-3 o

SALEM - UNIT 1 II Amendment No. 201

..._.__..___.____.__________.__.__...__m DEFINITIONS CONTAINMENT INTEGRITY l

1.7 -CONTAINMENT INTEGRITY shall exist s* tens j

1.7.i A.ts penetrations required to be closed during' accident conditions-are either:

1 4

a.

Capable of being closed by an OPERABLE containment I

. automatic isolation valve system, or l

b.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions,-except for valves that are open under administrative control as permitted by Specification 3.6.3.1.

1.7.7 All equipment-hatches are closed and sealed, 1.7.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.7.4 The containment leakage rates are with.

the limits of Specification 3.6.1.2, and 1.7.5 The sealing mechanism associated with each penetration (e.g.,

f

~

welds, bellows or 0 rings) is CPERABLE.

1.8 NOT USED CORE ALTERATION 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 the vessel.

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

CORE OPERATING LIMITS REPORT 4

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 determined for each reload cycle in accordance with Specification 6.9.1.9.

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

POSE EOUIVA MNT I-131 1.10 DOSE EQUIVALENT"I-131 shall-be that concentration of 1-131 (microcuries per gram) which alone would produce the same thyroid dose as the quantity and

' isotopic mixture of I-131, 1-132, I-133, I-134, and I 135 actually present.

The-SALEM - UNIT 1 1-2 Amendment No. 201

~

.. _... _ _ _.. _ _.. _ _ _ _... _. _. _.. _ _ _ _ _ _. = _ _ _ _ _ _.. _ _.

I 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS f

' 2.1 RAFETY LIMITS REACTOR CORE I

i 2.1.1 The combination of THERMAL POWER, pressuriser pressure, and the highest operating loop coolant tenperature -(T,,)

shall not exceed the limits shown in Figure 2.1-1 for 4 loop operation.

L i.

APPLICABILITY: MODES 1 and 2.

.[

ACTIONt S

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer I

pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

.,i i

e l

I v

I 4

i

)

I a

f

. SALEN'i UNIT 1 21

. Amendment No. 201

..___.m-_,..

,_ 2

. ~..,,

2. ~ -..

i b/u 9 tismo 46 seo *w 650 NCMAE2 I

NN S

!z 640 ase es.4 4 # 888 8) 8w3 isesnai ff 630 b"

men

\\

g p 620 mum moo pp 6

DNS

\\

h610 h

e

=

a: 600 o nei 43 o

W IW 590 dC onomm

* * *' " e ACC M W 560 gg (13aso2 1

$70 it 3assai g W

560 C

at s arn 550 '

00 01 0.2 03 04 05 06 07 08'09 1.0 11 12 13 FRACTION OF RAMD THERMAL POWER FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT - FOUR LOOPS IN OPERATION SALEM - UNIT 1 22 Ame.:dment No.- 201

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PEACTOR COOLANT SYSTEM PRESS M 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY: MODES 1, 2,

3, 4 and 5.

EQlf 8 MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 prig, be in HOT STANDBY with the Reactor Coolant System pressure within ity limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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

l I

l l

SALEM - UNIT 1 2-3 Amendment No. 201

4 i

TABLE 2.2-1 i

t REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

]'

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

}

1.

Manual Reactor Trip Not Applicable Not Applicable a

2.

Power Range, Neutron Flux Low Setpoint - s 25% of RATED Low Setpoint - s 26% of RATED TICRMAL POWER THERMAL POWER i

i f

High Setpoint - s 109% of RATED High Setpoint - s 110% of RATED i

THERMAL POWER THERMAL POWER 1

I

{

3.

Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 5.5% of RATED THERMAL POWER

]

High Positive Rate a time constant a 2 seconds with a time constant a 2.. seconds

, 'l 4.

Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 5.5% of RATED THERMAL POWER.

.)

j j

High Negative Rate a time constant a 2 seconds with a time constant a 2 seconds i

j 5.

Intermediate' Range Neutron s 25% of RATED THERMAL POWER s 30% of RATED THERMAL POWER i

Flux f

'i r

i 6.

Source Range, Neutron Flux s 10 counts per second s 1.3 x 10 counts per second f

5 5

I 7.

Overtemperature AT See Note 1 See Note 3 4

8.

Overpower AT See Note 2 See Note 4

'3 i

9.

Pressurizer Pressure--Low 2 1865 psig 2 1855 psig l

10.

Pressurizer Pressure--High s 2385 psig s 2395 psig

'f 3

s I

11.

Pressurizer Nater Level--High s 92% of instrument span s 93% of' instrument span 12.

Loss of Flow a 90% of design flow per loop

2 89% of design flow per loop *

[

l

Design flow is 82,500 gpm per loop.

i.

j SALEM - UKIT 1 2-5 Amendment No. M

i 4

,,+ -

TABLE 2.2-1 (Continued)

REACTOR TRIP SYS.*EM INSTRUMENTATION TRIP SETPOINTS NgjwTION (Continued)

Operation with 4 Loops K1 1.22

=

K2 0.02037

=

K3 0.001020

=

and f (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chanbers; with gains to be selected based on measured instrument response during plant startup tests such thats l

(1) for qt - go between -23 percent and +13 percent, ft (AI) =0 (where qt and go are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + go is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (qt go) exceeds

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

(iii) for each percent that the magnitude of (qt - go) exceeds

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

SALEM - UNIT 1 2-8 Amendment No. 201

-- -.. ~.

TADLE 2.2-1 (Cpntinued)

REACTOR TRIP SYFTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued) e NOTE 2: Overpower: AT s AT, IK -K3 7s T - rg (T-T")-f ( AI))

3 2

4 1+ 7 S 3

9

=

where:

AT, = Indicated AT at RATED THERMAL POWER T

= Average temperature,

'F 7"

= Indicated T, at RATED THERMAL POWER s 577.9'F p

l K4 = 1.09 K3 = 0.02/'F for increasing average temperature and 0 for decreasing average temperature K

= 0.00149/*F for T > T"; K6* O for T s T" 6

r,J

= The function generated by the rate lag controller 1+ f S for 7,3 dynamic compensation 3

73

= Time constant utilized in the rate lag controller for T

73 = 10 secs.

ag S

= Laplace transform operator, Sec'I f ( AI)

= 0 for all AI 2

NOTE 3: The channel's maximum trip point shall not exceed its computed trip point by more than 1.1 percent.

NOTE 4: The channel's maximum trip point shall not exceed its computed trip point by more than 2.1 percent.

SALEM - UNIT 1 2-9 Amendment No. 201

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boi regime could result in excessive cladding temperatures because of the

' et of departure from nucleate boiling (DNB) and the resultant sharp eeduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through correlations which have been g

developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

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

The DNB design basis is as follows: uncertainties in the WRB-1 and WRB-2 correlations, plant operating parameters, nuclear and thernal parameters, fuel fabrication parameters, and sputer 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 or 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, 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.

t SALEM - UNIT 1 B 2-1 Amendment No. 201

.. ~ _..-.

I SAFETY LIMITS BASES The curves are based on an enthalpy hot channel factor, F "as, and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in r"aa at reduced power based on the expressions F "an [1. 0 + Pran (1. 0 - P ) )

r"an

=

where F "aw is the lindt at RATED THERHAL POWER (RTP) specified in the Core Operating Limits Report (COLR).

PraH is the Power factor Multiplier for r"aH spe. #ied in the COLR, and P is THERMAL POWER RATED THERMAL POWER These limiting heat flux conditions ace higher than those calculated for the range of all control rod positions from rods FULLY WITHDRAWN to the l

maximum allowable control rod insertion assuming tho axial power imbalance is within the limits of the ft(AI) function of the overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature AT trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Reactor Coolant System piping and fittings are designed to ANSI B 31.1 1955 Edition while the valves are designed to ANSI B 16.5, MSS-SP-66-1964, or ASME Section III-1968, which permit maximum transient pressures of up to 120% (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prict to initial operation.

Salem - Unit 1 B 2-2 Amendment No. 201

~.

-.. - _ - _ - - ~. -

i LIMITING SAFETY SYSTEM SETT M 4

operation with a reactor coolant loop out of service below the 4 loop P-8 set point does not require reactor protection system cet point modification because the P-8 set point and associated trip will prevent DNB during 3_ loop operation exclusive of the overtemperature AT set point. Three loop operation above the 4 loop P-8 set point has not been evaluated and is not permitted.

Overoower AT i

The overpower AT reactor trip provides assurance of fuel integrity, e_.g.,

no melting, under all possible overpower conditions, limits the required range for overtemperature AT protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes in density and heat capacity of water 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 required by this specification to enhance the overall reliability of the Reactor Protection System.

i Pressurizer Pressure i

The Pressurizer High and Low 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 in therefore set lowe than the set pressure for these valves j.

(2485 psig). The Low Pressure trip provideo protection by tripping the reactor in the event of a loss of reactor coolant pressure, i

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor coolant System overpressurization by limiting the water level to a volume

' sufficient to retain a steam bubble and prevent water relief t

f SALEM - UNIT 1 B 2-5 Amendment No. 201

.~J..

n-.,%

-,.,--,i

.n

,n,.,

,.c,

-,-,n,.

n - - -,,.

.n,.-,n n.,

c,-

i l

1 i

.l i

LIMITING SAFETY SYSTEM SETTINGS l

AMisa s

throug's the pressuriser 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.

i overall reliability of the Reactor Protection System.

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

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

t i

i I

i

~

SALEM (-LUNIT_1 B26

' Amendment No. 201'

....-. 2 u...

.u..

,__.m.

_. _ _.. _... _... _,,.. ~. _

_._.._______.____.___..___._.___m_,

1 3/4.1 REACTIVITY CCNTROL SYSTEMS 3/4.1.1' BORATION CONTROL 1

EHUTDOWN MARGIN - Tava > 200'F I

LIMITING CONDITION FOR OPERATION i

3.1.1.1 The SHUTDOWN MARGIN shall be a 1.3% Ak/k.

I APPLICABILITY: MODES 1, 2*,

3, and 4.

l ACTION:

With the SHUTDOWN MARGIN < 1.3% Ak/k, itindiately initiate and continue boration_at a 33 gpm of a solution containing a 6,560 ppm boron or equivalent j

until the required SHUTDOWN MARGIN is restored.

t SURVEILLANCE REQUIREMENTS i

4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be a 1.3% Ak/k a.

Within one hour after detection of an inoperable control rod (s)and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereaf ter while the rod (s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be. increased by an amount at least equal to the withdrawn worth of the immovable or untrippable control rod (s),

8 b.

When in MODES 1 or 2 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits in the COLR per Specification 3.1.3.5.

When in MODE 2,

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor c.

criticality by verifying that the predicted critical control rod position is within the limits in the COLR per specification 3.1.3.5.

l

?

See Special Test Exception 3.10.1

- With 4 a,1.0

1. With Km i 1 0

[

SALL

7 TIT 1 3/4 1 Amendment No. 201

.:. =

a.--

REACTIVITY CONTROL SYSTEMS SURVEILLMICE REQUIREMENTS (Continued) d.

Prior to initial operation above 5% RATED THERMAL POWER after cach fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit in the COLR per Specification 3.1.3.5.

e.

When in HODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1.

Reactor coolant system boron concentration,-

2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Puel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.2.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within 12% Ak/k at least once per 31 Effective Pull Power Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4.1.1.1.1.e, above. The predicted 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.

l r

1 l

SALEM - UNIT 1 3/4 1-2 Amendment No. 201

i l

l I

REACTIVITY COtfTROL SYSTEMS i

MODEPATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.4 The moderator temperature coefficient (MTC) shall be within the limits specified in the CORE OPERATING LIMITS REPORT (COLR).

The maximum upper limit shall be less positive than or equal to 0 Ak/k/*F.

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

End of Cycle Life (EOL) Limit MODES 1, 2 and 3 only#

ACTION:

a.

With the MTC more positive than the BOL limit specified in the COLR, operations in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to less positive than the BOL limit specified in the COLR within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdrawal limits l

shall be in addition to the insertion limits in the COLR per Specification 3.1.3.5.

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 *.tithdrawn condition.

3, 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 MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition, b.

With the MTC more negative than the EOL limit specified in the COLR, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With N, greater than or equal to 1.0

See Special Test Exception 3.10.3 SALEM - UNIT 1 3/4 1-5 Amendment No. 201

}

EEACTIVITY CONTROL SYSTEMS MODEPATOR TEMPERATURE COEFFICID7T SURVE!L1JJiCE REQUIREMDITS 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 specified in the COLR, prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading, b.

The MTC shall be measured at any THERMAL POWER and compared to the 300 ppm surveillance limit specified in the COLR (all rods withdrawn, RATED THERMAL POWER condition) within 7 EFPD after reaching an equilibrium boron concentration of 300 ppm.

In the event thi> comparison indicates the MTC is more neaative than the 300 ppm surveillance limit specified in the COLR, the MTC shall be remeasured, and compared to the EOL MTC limit specified in the COLR, at least once per 14 EFFD during the remainder of the fuel cycle.

l SALEM - UNIT 1 3/4 1-Sa Amendment No. 201

REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLTES GROUP HEIGHT LIMITING CONDITION FOR OPERATION 3.1.3.1 All full length (shutdown and control) rods, shall be OPERABLE and positioned within i 12 steps (indicated position) of their group step counter demand position within one hour after rod motion.

APPLICABILITY: MODES 1* and 2*

ACTION:

a.

_With one or more full length rods inoperable due to being 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 i hour and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , b.

With more than one full length rod inoperable or mis-aligneo from the group step counter demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With one full length rod inoperable due to causes other than addressed by ACTION a, above, or mis-aligned from its group step counter demand position by more than t 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 remainder of the rods in the bank with the inoperable rod are aligned to within 2 12 steps 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 be 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 1 3/4 1-18 Amendment No. 201

REACTIVITY CONTPOL SYSTEMS EQ)1 TION INDICATION SYSTEM SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.5 The control banks shall be limited in physical insertion as specified in the CORE OPERATING LIMITS REPORT (COLR).

APPLICABILITY: MODES 1*, and 2*#

ACTION:

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

a.

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

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

SURVEILLANCE REQUIREMENTS 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months **.

e 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.

With 4 greater than or equal to 1.0 SALEM - UNIT 1 3/4 1-23 Amendment No. 201

i THIS PAGE INTENTIONALLY LEFT BLANK f

I a

A SALEM-- UNIT 1 3/4 1 24 Amendment No.'201

t I

e THIS PAGE INTENTIONALLY LEFT BLANK SALEM - UNIT 1.-

3/4 1 25 Amendment No. 201

.. _ _ ~

3/4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE (AFD)

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

APPLICABILITY:

MODE 1 ABOVE 50% RATED THERMAL POWER

ACTION:

a.

With the indicated AXIAL FLUX DIFFERENCE outside of the limits 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 limits specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the prev.1 us 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and 2)

The indicated AFD is within the limits specified in the COLR. Otherwise, reduce THERMAL POWEn to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to a 55% of RATED THERMAL POWER within th; next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b)

Surveillance testing of the Power Range Neutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limits specified in the COLR.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operaticn may be accumulated with the AFD outside of the target band during this testing without penalty deviation.

i

See Special Test Exception 3.10.2 l

SALEM - UNIT 1 3/4 2-1 Amendment No. 201 1

{

~.

i POWER DISTRIBUTION LIMITS l

P LIMITING CONDITION FOR OPERATION (Continued) i b.

THERMAL POWER shall not be increased above 906 of RATED THERMAL POWER unless the indicated AFD is within the limits specified in the COLR_and ACTION 2.a)1), above has been.atisfied.

c.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the lindts specified in the COLR for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation l

cumulative during the previcas 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVE!LLANCE RBQP,2REMENTS 4.2.1.1 The indicated AXIAL PLUX DIFTERENCE shall be determined to be within its limits during F0WER OPERATION above 15% of RATED THERMAL POWER by a.

Monitoring the indicated AFD for each OPERABLE excore channel 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status.

b.

Monitoring and logging the indicated AXIAL TLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 houra and at least once per 30 minutes thereafter, when the AXIAL PLUX DIFFERENCE Monitor Alarm is inoperable. The logged values of the indicated AXIAL TLUX DIFFERENCE shall be assumed to exist during the interval pre eding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its limits when at least 2 of 4 or 2 of 3 OPERABLE excore channels are indicating the AFD to be outside of the target band.

Penalty deviation outside of the target band 3

l shall be accumulated on a < me basis of:

a.

One minute penalty deviation for each one minute of POWER OPERATION outside of tht. limits 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 limits at THERMAL POWER levels below 50%

of RATED THERMAL POWER.

SALEM - UNIT 1 3/4 2-?

Amendment No. 201 4

.r

-u'T-**r"i-ef-'-Y-Get t'"'t

'#" DC igrv -g'+WW*WT'F' rerun'Nfw'sPrv= TF W9 f "Y-?"W#7'W7N-'-F'T*=qWed-ewr-w1'T-kw 7we yT"VW M> g r,menem--M9emm-ms,g-ywew-f 9-w gi.w 7.nyy-y--i.cy-w

,y-,..e wT hwC'-h9-

i I

i THIS PAGE INTENTIONALLY LEFT BLANK r

SALEM - UNIT 1 3/4 2-4 Amendment No. 201

+ --

- -. - ~ ~. -

f e

POWER DISTRIBt?TICri! LIMITS i

sauT FLtfX HOT CHANNEL FACTOR-Fn(2)

LIMITING CONDITION FOR OPERATION

-3.2.2 Fg(z) shall be limited by the following relationships:

I Fg(z)g yRTP

K(z) for P > 0.5,'and o

P RTr Fg(z) sF o

K(z) for P g 0.5, 0.5 t

Where: FRTPga the Fg limit at RATED THERMAL POWER (RTP) specified in the CORE OPERATING LIMITS REPORT (COLR),

P=

THERMAL POWER and RATED THERMAL POWER K(z) the normalized Fg(z) as a function of core height as

=

specified in the COLR.

. g APPLICABILITY: MODE 1 ACTION:

With Fn(s) exceeding its limits a.

Reduce THERMAL POWER at least 1% for each 1% Fn(z) exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip setpoints have been reduced at least 1% for each 1% Fn(z) exceeds the limit.

The Overpower AT Trip Setpoint reduction shall be performed with the reactor in at least HOT STANDBY.

r b.-

Identify and correct the cause of the out-of limit condition prior to increasing THERMAL POWER ebove the reduced limit required by a.

2 above: THERMAL POWER may then-be increased provided Fg(z) is

' demonstrated through incore mapping to be within its limit.

i I

l.

i SALEM - UNIT.1.

3/4 2-5' Amendment No. 201 V

.-a-wwkwN-,..,-

-,wgy m 4.--my-i eer www4gng www pm

,q,.rA."

g-m--

.-w,.,

p.

y y y-g4.+

4se-p g

9

-g-e--

w-6 ge9 y

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 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 Fn(Z) is within its limit by:

I a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER, b,

Ineteasing the measured Fay component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to arcount for measurement uncertainties.

C c.

Comparing the F,. computed (F,y) obtained in b, above tot 1.

The F,y limits for RATED THERMAL POWER (FRTP,y ) for the appropriate measured core planes given in e and f below, and 2.

The relationships F,yl = F,RTP [ g, pp,y (3,p))

L where F,y is the limit for fractional THERMAL POWER operation expressed as a function of FR TP,,, PP,y is the power f actor multiplier for F,y in the COLR, and P ie the fraction of RATED THERMAL POWER at which F,y was measured.

d.

Remeasuring F, according to the following schedule:

C 1.

When Fay is greater than the F"IP,y limit for the L

appropriate measured core plane but less than the F,y relationship, additional power distribution maps shall be C

taken and F,y compared to FRTP,y and F',y a)

Either within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at C

which F, was last determined, or SALEM - UNIT 1 3/4 2-6 Amendment No. 201

POWER DISTRIBUTION I IMITS SURVEILLANCE REQUIREMENTS (Continued) b)

At least once per 31 EFFD, whichever occurs first.

C 2.

When the F,y is less than or equal to the FRTP,, limit f or the appropriate measured core plane, additional power distribution maps shall be taken and F, compared to FRTP,,

C L

and F,y at least once per 31 EFPD.

e.

The F,y limit for Rated Thermal Powv. (FRTP,y) shall be provided for all core planes containing bank "D" control rods and all unrodded core planes in the COLR per specification 6.9.1.9.

f.

The F,y limits of e, above, are not applicable in the following core plane regions as measured in percent of core height from the bottom of the fuels 1.

Lower core region from 0 to 15% inclusive.

2.

Upper core region from 65 to 100% inclusive.

3.

Grid plane regions at 17.8 t 2%, 32.1 12%, 46.4 12%, 60.6 12%,

and 74.9 12% inclusive.

4.

Core plane regions within 12% of core height (12.88 inches) about the bank demand position of the bank "D" control rods.

g.

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

within its limit whenever

,, exceeds F

,y.

SALEM - UNIT 1 3/4 2-7 Amendment No. 201

THIS PAGE INTENTICIALLY LEFT BLANK.

SALD4 - UNIT 1 3/4.7-8 Amendment No. 201 I

l POWER DISTRIBUTION LI111TS N

NUCLEAR ENTHALPY HOT CHANNEL FACTOR F AH LIMITING CONDITION FOR OPERATION N

3.2.3 F

shall be limited by the following relationship F ap = pkTPau (1.0 + PFAH (1. 0 - P) )

N Where F"TP is the limit at RATED THERMAL POWER (RTP) specified in AH the CORE OPERATING LIMITS REPORT (COLR).

N PFAH is the Power Factor Multiplier for F AH sPecified in the COLR, and P is THERMAL POWER RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

N With F AH exceeding its limit a.

Reduce THERMAL POWER to less than 50% of RATED THSRMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux High Trip Setpoints to s 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , N

b.

Demonstrate thru in-core mapping that F is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and c.

Identify and correct the cause nf the out of limit condition prior to increasing THERMAL PO'4R above the reduced limit required by a, or b. above; subsequent POWER OPERATION may proceed provided that N

F u is demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95%

or greater RATED THERMAL POWER.

wd EM - UNIT 1 3/4 2-9 Amendment No. 201

l t

p TABLE 3.2-1 s

DNB PARAMETERS PARAMETER LIMITS 4 Loops Operat' Reactor Coolant System T.vo 5 582.f

Pressurizer Pressure 2 2200 psi.

Reactor coolant System Flow 2 341,000 gpn..

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

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

uncertainty due to feedwater venturi fouling.

SALEM - UNIT 1 3/4 2-14 Amendment No. 201

l' l

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made suberitical from all operating conditions, 2) the reactivity transients 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.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T,g.

The most restrictive condition occurs at EOL, with Tes at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled hCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.3% Ak/k is initially required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysin assumptions.

With T, s 200'F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% Ak/k shutdown margin provides adequete protection.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limitations en MTC are provided to ensure that the value of this coeff.cient remains within the limiting condition assumed in the accident and transient analyses.

SALEM - UNIT 1 B 3/4 1-1 Amendment No. 201

)

' ' ' ~ ~ ' ~ ~ ~ ~ ~ '

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 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 dif fert. ices in MTC observed between EOL, all rods withdrawn, RATED 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) MTC value.

The 300 ppm surveillance limit MTC value represents a conservative value at a core condition of 300 ppm equilibrium boron concentration that is obtained by correcting the limiting EOL MTC for burnup and boron 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 boren concentration associated with fuel burnup.

3/4.1.1.5 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 ant.lyzed tempera:ure range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its allowable setpoint, 4) the preosurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTNDT temperature.

SALEM - UNIT 1 B 3/4 1-2 Amendment No.201

l REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTj211 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, and 5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature a 350'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The

'f boration capability of either flow path is suf ficient to provide a SHUTDOWN g

MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay l

and cooldown 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 a' EOL from full power equilibrium xenon conditions and requires borated water from a boric acid tank in accordance with TS Figure 3.1-2, and additional makeup from either:

(1) the second boric acid tank and/or batching, or (2) a maximum of 41,800 gallons of 2,300 ppm borated water from the refueling water stotage tank.

With the refueling water storage tank as the only boratcJ water source, a maximum of 73,800 gallons of 2,100 ppm borated water is required.

However, to be consistent with the ECCS requirements, the RWST is required to nave a minimum contained volume of 350,000 gallons during operations in MODES 1, 2, 3

and 4.

The boric acid tanks, pumps, valves, and piping contain a boric acid solution concentration of between 3.75% and 4.0% 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 rcom. 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 above 63*F.

A small amount of beric acid in the flow path between the boric acid recirculation line and the auction line to the charging pump will precipitate out, but it will not cause flow blockage even with temperatures below 50*F.

With the RCS temperature below 350*F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

SALEM - UNIT 1 B 3/4 1-3 Amendment No. 201

l 3 /4. 2 POWER DISTRIBtTTION LIMITS o

BASES 1

The specifications of this section provide assurance of fuel integrity during Condition 1 (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) meeting the DNB design criterion during normal operation and in short teim 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 specifications are ar follows:

Fn(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

N F 3n 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.

Fn(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 Fg(Z) upper bound envelope of the Fq limit specified in the Cote Operating Limits Report (COLR) times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following powar 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 le.els.

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 burnup conditions.

Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER Jevel. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

SALEM - UNIT 1 B 3/4 2-1 Amendment No. 201

I l

POWER DISTRIBUTION LIMITS BAS"S Although it is intended that the plant will be operated with the AXIAL FLUX DIFFERENCE within the target band in the COLR per Specification 3.2.1 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 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 POWER (with the AFD within the target band) provided the time duration of the deviation is limited. Accordingly. 4 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 aours is provided for operation outside of the target band but within the limits specified in the COLR 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 penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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.

SALEM - UNIT 1 B 3/4 2-2 Amendment No. 201

INFORMATION ONLY*

Percent of Rated Thermal Power 10 0 %

90%

80%

70 %

(

Target Flux Difference 60%

50%

40%

30%

20%

10 %

0

-20%

-10%

0 10 %

20%

INDICATED AXIAL FLUX DIFFERENCE Figure B 3M 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE YERSUS THERMAL PCWER Refer to COLR Figure 2 for Actual Limits SALEM - UNIT 1 B 3/4 2-3 Amendment No. 201

r-..

.~.-------n

=.

EQWJMIS_TRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL AND PADIAL N

PFtWING FACTORS - Fg(Z), F gng p gg) 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 exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

e Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

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

a.

Control rod in a single group move together with no individual rod insertion differing by more than i 12 steps from the group demand

position, b.

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

c.

The control rod insertion limits of Specifications 3.1.3.4 and 3.1.3.5 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

N The relaxation in F AH as a function of THERMAL POWER allows changes in N

the radial cower shape for all permissible rod insertion limits.

F H will be maintained within its limits provided conditions a thru d above, are maintained.

When an Fq measurement is taken, both experimental error and manufacturing tolerance must be allowed for 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3%

is the appropriate allowance for manufacturing tolerance.

N 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 N

system. The specified limit for F also Contains an 8% allowance for uncertainties which mean that normal operation will result in Fa g pRP

/1.08 where FRTP is the limit of RATED THERMAL POWER (RTP)

N AH H

specified in the CORE OPERATING LIMITS REPORT (COLR). The 8% allowance is based on the following considerations:

SALEM - UNIT 1 B 3/4 2-4 Amendment No. 201

POWER DISTRIBUTION LIMITS BASES a.

abnorum1 perturbations in the radial power shape, such as from rod misalignment, effect FN nere directly Fq, b.

although rod movement has a direct influence upon limiting Fo to within its limit, such control is not readily available to limit N

F 33g c.

errors in prediction for control power shape detected during startup physico tests can be compensated for in Fq by restricting axial flux N

distributions.

This compensation for F H is less readily available.

The radial peaking f actor F,y(z) is measured periodically to provide assurance that the hot channel factor, Fn(z), remains within its limit. The Fay limit for Rated Thermal Power (FRTP

), as provid$d in the COLR per opecification ay 6. 9.1. 9, was determined from expected powar control maneuvers over the full range of burnup conditions in the core.

3/4.2.4 OUADRANT POWER TILT PATIO 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 inade during startup testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts. A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in Fq is depleted. The limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fq is reinstated by reducing the power by 3 percent from RATED THERMAL POWER for each percent of tilt in excess of 1.0.

SALEM - UNIT 1 B 3/4 2-5 Amendment No 201

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation, and m et the DNB criterion during all normal operations and 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 HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, a single reactor coolant loop provides sufficient heat removal for removing decay heat; but, single failure considerations require all loops be in operation whenever the rod control system is energized and at least one loop be in operation when the rod control system is deenergized.

In MODE 4, a single reactor coolant loop or RHR loop provides sufficient

' heat removal for removing decay heat; but, single failure censiderations 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 RHR loops be OPERABLE. The provisions of Sections 3.4.1.4 and 3.9.8.2 (paragraph (b) of footnote (+)] which permit one service water header to be out of service, are based on the following:

I.

The period of time during which plani operatirns rely upon the provisions of this footnote shall be lioited to a cumulative 45 days for any single outage, and 2.

The Gas Turbine sk-ll be operable, as a backup to the diesel generators, in the event of a loss of offsite power, to supply the applicable loads.

The basis for OPERABILITY is one successful startup af the Gas Turbine no more than 14 days prior to the beginning of the Unit outage.

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 eate associated with Boron concentration reductions will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs less than or equal to 312*F are provided to prevent RCS pressure transie:cs, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10CFR Part 50.

The RCS will be protected against overpressure transients and will not exceed the limits of Appendix 0 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 secondary water temperature in each steam generator is less than 50*F above each of the RCS cold leg temperatures.

SALEM - UNIT 1 B 3/4 4-1 Amendment No. 201

DESIGH FEATURES 3.

In accordance with the code requirements specified in Section 4.1 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 650'F, except for the pressurizer which is-680'F.

VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,446 1 426 cubic feet at a nominal Tavg of 573*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

5.6 FUEL STORAGE CRITICALITY 5.6.1.1 The new fuel storage racks are designed and shall be maintained with:

a.

A maximum K.n equivalent of 0.95 with the storage racks flooded with unborated water.

b.

A nominal 21.0 inch center-to-center distance between fuel assemblies, c.

A maximum unirradiated fuel assembly enrichment of 4.5 w/o U-235, 5.6.1.2 The spent fuel storage racks are designed and shall be maintained with:

a.

A maximum K.n equivalent of 0.95 with the storage racks filled with unborated water.

b.

A nominal 10.5 inch center-to-center distance between fue.

assemblies stored in Region 1 (flux trap type) racks.

c.

A nominal 9.05 inch center-to-center distance between fuel assemblies stored in Region 2 (non-flux trap) racks, d.

Fuel assemblies stored in Region 1 racks shall meet one of the following storage constraints.

1.

Unirradiated fuel assemblies with a maximum enrichment of 4.25 w/o U-235 have unrestricted storage.

SALEM - UNIT 1 5-5 Amendment No. 201

i i

ADMINISTRATIVE CONTROLS d.

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

e.

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

and f.

Solidification agent or absorbent (e.g.,

cement, urea formaldehyde).

The Radioactive Effluent Release Reports shall include a list of descriptions of unplanned releases from the site to UNRESTRICTED AREAS of radioacti/e 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. der,tified by the land use censur 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 COLR for the following:

1.

Moderator Temperature Coefficient Beginning of Life (BOL) and End of Life (EOL) limits and 300 ppm surveillance limit for Specification 3/4.1.1.4, 2.

Control Bank Insertion Limits for Specification 3/4.1.3.5, 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 o

height, K(z), and Power Factor Multiplier PF,,,

Specification 3/4.2.2, and 5.

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

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, Westinchouse Peload Safety Evaluation Methodoloov, July 1985 (H Proprietary),

Methodology for Specifications listed in 6.9.1.9.a.

Approved by Safety Evaluation dated May 28, 1985.

SALEM - UNIT 1 6-24 Amendment No. 201 l

7 l

l ADMINISTRATIVE CONTROLS 2.

WCAP-8385, Power Distribution Control and Load Followino Procedures - ToDical ReDort, September 1974 (H Proprietary)

Methodology for Specification 3/4.2.1 Axial Flux Dif ference.

Approved by Safety Evaluation dated January 31, 1978.

3.

WCAP-10054-P-A, Rev.

1, Westinahouse small Break ECCS Evaluation Model Usino NOTRUMP Code, August 1985 (E Proprietary),

Methodology for Specification 3/4.2.2 Heat Flux Hot Channel Factor. Approved for Salem by NRC letter dated August 25, 1993.

4.

WCAP-10266-P-A, Rev. 2, The 1981 Version of Westinchouse Evaluation Model Usino BASH Code, Rev. 2. March 1987 (W Proprietary) Methodology for Specifict. tion 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 applic.able limits (e.g.,

fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS!

limits, 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.

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 A'"inistrator, 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 Analysis 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 Licensee 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 monitoring for inadequate core cooling, the cause of the inoperability, and the plans and schedule for restoring the instrument channels to OPERABLE status.

SALEM - UNIT 1 6-24a Amendment No.201

.

ML20203C705

Text

Attachment:

Navigation menu

Search