ML070810419

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Technical Specification for Amendment 278 Power Range Neutron Flux High Negative Rate Trip
ML070810419
Person / Time
Site: Salem PSEG icon.png
Issue date: 03/19/2007
From:
Plant Licensing Branch III-2
To:
Ennis R, NRR/DORL, 415-1420
Shared Package
ML070530283 List:
References
TAC MD1490
Download: ML070810419 (8)
v • d • e


Text

(1) Maximum Power Level PSEG Nuclear LLC is authorized to operate the facility at a steady state reactor core power level not in excess of 3459 megawatts (one hundred percent of rated core power).

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.278 are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(3) Deleted Per Amendment 22, 11-20-79 (4) Less than Four Loop Operation PSEG Nuclear LLC shall not operate the reactor at power levels above P-7 (as defined in Table 3.3-1 of Specification 3.3.1.1 of Appendix A to this license) with less than four (4) reactor coolant loops in operation until safety analyses for less than four loop operation have been submitted by the licensees and approval for less than four loop operation at power levels above P-7 has been granted by the Commission by Amendment of this license.

(5) PSEG Nuclear LLC shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, and as approved in the NRC Safety Evaluation Report dated November 20, 1979, and in its supplements, subject to the following provision:

PSEG Nuclear LLC may make changes to the approved fire protection program without prior, approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Amendment No.278

INDEX BASES SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE AND 3/4.3.2 ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION .................................. B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION ...................... B 3/4 3-la 3/4.3.4 TURBINE OVERSPEED PROTECTION .................... B 3/4 3-4 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION .............................. ....... B 3/4 4-1 3/4.4.2 SAFETY VALVES ...................... .............. B 3/4 4-la 3/4.4.3 RELIEF VALVES ...................... .............. B 3/4 4-la 3/4.4.4 PRESSURIZER ......................... .............. B 3/4 4-2 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY .............. B 3/4 4-2 3/4 .4 .6 REACTOR COOLANT SYSTEM LEAKAGE ..... ............. B 3/4 4-4a 3/4 . 4 .7 DELETED 3/4 .4 .8 SPECIFIC ACTIVITY ............................... B 3/4 4-5 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ..................... B 3/4 4-6 3/4 .4 .10 STRUCTURAL INTEGRITY ............................. B 3/4 4-17 3/4.4.11 BLANK ........................................... B 3/4 4-17 3/4.4.12 REACTOR VESSEL HEAD VENTS ....................... B 3/4 4-17 SALEM - UNIT 1 Xll Amendment No. 278

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

1. Manual Reactor Trip Not-Applicable Not Applicable
2. Power Range, Neutron Fluix Low Setpoint -
  • 25% of RATED Low Setpoint -
  • 26% of RATED THERMAL POWER THERMAL POWER High Setpoint -
  • 109% of RATED High Setpoint -
  • 110% of RATED THERMAL POWER THERMAL POWER
3. Power Range, Neutron Fluix,
  • 5% of RATED THERMAL POWER with
  • 5.5% of RATED THERMAL POWER High Positive Rate a time constant Ž 2 seconds with a time constant Ž 2 seconds
4. Deleted
5. Intermediate Range, Neut iron 25% of RATED THERMAL POWER
  • 30% of RATED THERMAL POWER Flux
6. Source Range, Neutron F!.ux 105 counts per second 1.3 x 105 counts per second
7. Overtemperature AT See Note 1 See Note 3
8. Overpower AT See Note 2 See Note 4
9. Pressurizer Pressure--Lo 1865 psig Ž 1855 psig
10. Pressurizer Pressure--Hi gh 2385 psig 5 2395 psig
11. Pressurizer Water Level- -High 5 92% of instrument span .* 93% of instrument span
12. Loss of Flow 90% of design flow per loop* 89% of design flow per loop*
  • Design flow is 82,500 gpm p er io<

SALEM - UNIT 1 2-5 Amendment No 278

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 FxY shall be evaluated to determine if Fo(Z) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map:
1. When THERMAL POWER is
  • 25%, but > 5% of RATED THERMAL POWER, or
2. When the Power Distribution Monitoring System (PDMS) is inoperable; and increasing the Measured FQ(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR.
b. Using the PDMS or the moveable incore detectors when THERMAL POWER is > 25% of RATED THERMAL POWER, and increasing the measured FQ(Z) by the applicable manufacturing and measurement uncertainties as specified in the COLR.
c. Comparing the FxY computed (Fxyc) obtained in b, above to:
1. The FY limits for RATED THERMAL POWER (FXyRTP) for the appropriate measured core planes given in e and f below, and
2. The relationship:

L =FxyRTP [I+PFy (l-P)i where FYL is the limit for fractional THERMAL POWER operation expressed as a function of FXyR1 PFy is the power factor multiplier for Fy in the COLR, and P is the fraction of RATED THERMAL POWER at which Fxy was measured.

d. Remeasuring FY according to the following schedule:
1. When Fxyc is greater than the FXyRTP limit for the appropriate measured core plane but less than the FxyL relationship, additional core power distribution measurements shall be taken and Fxyc compared to FxyRTP and FxyL :

a) Either within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding by 20% of RATED THERMAL POWER or greater, the THERMAL POWER at which Fyc was last determined, or SALEM - UNIT 1 3/4 2-6 Amendment No. 278

TABLE 3.3-1 R EACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NUMBER CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 2 i. 2 1,2 and
  • 12
2. Power Range, Neutron Flux 4 2 3 1,2, and 3* 2
3. Power Range, Neutron Flux 4 2 3 1,2 2 High Positive Rate
4. Deleted
5. Intermediate Range, Neutron Flux 2 1 1,2 and
  • 3
6. Source Range, Neutron Flux A. Startup 2 1 2 2## and
  • 4 B. Shutdown 2 0 1 3,4, and 5 5
7. Overtemperature AT 4 2 3 1,2 6
8. Overpower AT 4 2 3 1,2 6
9. Pressurizer Pressure-Low 4 2 3 1,2 6
10. Pressurizer Pressure--High 4 2 3 1,2 6 278 SALEM - UNIT 1 3/4 3-2 Amendment No.

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT RESPONSE TIME

1. Manual Reactor Trip NOT APPLICABLE
2. Power Range, Neutron Flux
  • 0.5 seconds*
3. Power Range, Neutron Flux, NOT APPLICABLE High Positive Rate
4. Deleted
5. Intermediate Range, Neutron Flux NOT APPLICABLE
6. Source Range, Neutron Flux NOT APPLICABLE
7. Overtemperature AT
  • 5.75 seconds*
8. Overpower AT NOT APPLICABLE
9. Pressurizer Pressure--Low
  • 2.0 seconds
10. Pressurizer Pressure--High
  • 2.0 seconds
11. Pressurizer Water Level--High NOT APPLICABLE
  • Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

SALEM - UNIT 1 3/4 3-9 Amendment No.278

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip Switcl N.A. N.A. R(9) 1, 2, and *
2. Power Range, Neutron Flux S D(2), M(3) Q 1, 2, and 3*

and Q(6)

3. Power Range, Neutron Flux, N.A. Q 1, 2 High Positive Rate
4. Deleted 1 '
5. Intermediate Range, Neutron Flux S R(6) S/U( 1, 2 and *
6. Source Range, Neutron Flux S(7)

R (6) Q and S/U(') 2, 3, 4, 5 and *

7. Overtemperature AT S R Q 1, 2
8. Overpower AT S R Q 1, 2
9. Pressurizer Pressure--Low S R Q 1, 2
10. Pressurizer Pressure--High S R Q 1, 2
11. Pressurizer Water Level--High S R Q 1, 2
12. Loss of Flow - Single Loop S R Q 1 SALEM - UNIT 1 3/4 3-I11 Amendment No278

ADMINISTRATIVE CONTROLS 6.9.1.5 Reports required on an annual basis shall include:

a. DELETED
b. DELETED
c. The results of any specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be included: (1) Reactor power history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (2) Results of the last isotopic analysis for radioiodine performed prior to exceeding the limit, results of analysis while the limit was exceeded and results of one analysis after the radioiodine activity was reduced to less than the limit. Each result should include date and time of sampling and the radioiodine concentrations; (3) Clean-up system flow history starting 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> prior to the first sample in which the limit was exceeded; (4) Graph of the 1-131 concentration and one other radioiodine isotope concentration in microcuries per gram as a function of time for the duration of the specific activity above the steady state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit.

MONTHLY OPERATING REPORT 6,9.1.6 DELETED SALEM - UNIT 1 6-21 Amendment No. 278

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