ML20209F721
| ML20209F721 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 04/21/1987 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Northeast Nuclear Energy Co (NNECO), Connecticut Light & Power Co, Western Massachusetts Electric Co |
| Shared Package | |
| ML20209F725 | List: |
| References | |
| TAC 64674, DPR-65-A-116 NUDOCS 8704300359 | |
| Download: ML20209F721 (15) | |
Text
'
es ur UNITED STATES
[
g NUCLEAR REGULATORY COMMISSION t
-l WASHINGTON, D. C. 20555
- ...+
me NORTHEAST NUCLEAR ENERGY COMPANY THE CONNECTICUT LIGHT AND POWER COMPANY THE WESTERN MASSACHUSETTS ELECTRIC COMPANY DOCKET NO. 50-336 MILLSTONE NUCLEAR POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.116 License No. DPR-65 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Northeast Nuclear Energy Company, et al. (the licensee), dated February 6, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, 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; 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.
8704300359 870421 PDR ADOCK 05000336 P
1b ~
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.f 21 of Facility Operating License No. DPR-65 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.116, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/-
h'N. Stolz, D recto Pro ect Directorate T-4 i 'sion Reactor Pro;Ibets I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: April 21,1987 i
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i i
ATTACHMENT TO LICENSE AMENDMENT NO. 21 FACILITY OPERATING LICENSE.NO. DPR-65 I
DOCKET NO. 50-336 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain a vertical line indicating the area of change.
The corresponding i
overleaf pages are provided to maintain document completeness.
Remove Pages Insert Pages IV IV 3/4 1-31 3/4 3-2 3/4 3-2 3/4 3-4 3/4 3-4 3/4 3-7 3/4 3-7 8 3/4 1-3 B 3/4 1-3 8 3/4 1-5 B 3/4 1-5 I
-r..
INDEX SAfdTY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS t
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PAGE SECTION 2.1 SAFETY LIMITS l
2-1 I
Reactor Core................................................
2-1 Reactor Coolant System Pressure.............................
2.2 LIMITING SAFETY SYSTEM SETTINGS 2-3 l
l Rea ctor Tri p Setpoi nts......................................
BASES PAGE SECTION 2.1 SAFETY LIMITS B 2-1 Reactor Core................................................
B 2-3 Reactor Cool ant System Pressure.............................
2.2 LIMITING SAFETY SYSTEM SETTINGS B 2-4 Reactor Trip Setpoints......................................
MILLSTONE - UNIT 2 III Amendment No.104 l
b 2
INDEX
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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...........................................
3/4 0-1 3/4.1-REACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L.....................................
3/4 1-1 Shutdown Margin - T,yg > 200*F.......................
3/4 1-1 Shutdown Margin - T,yg <_ 200*F.......................
3/4 1-3 Boron Dilution...................................... 3/4 1-4 Moderator Temperature Coefficient (MTC).............. 3/4 1-5 Minimum Temperature for Criticali ty.................. 3/4 1-7 3/4.1.2 B0 RATION SYSTEMS.....................................
3/4 1-8 Fl ow P a t h s - S h u td ow n................................
3/4 1-8 Flow Paths - Operating...............................
3/4 1-10 Charging Pump - Shutdown.............................
3/4 1-12 Charging Pumps - Operating...........................
3/4 1-13 Boric Acid Pumps - Shutdown..........................
3/4 1-14 Bori c Acid Pumps - Operating.........................
3/4 1-15 Borated Water Sources - Shutdown.....................
3/4 1-16 Barated Water Sources - Operating.................... 3/4 1-18 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...........................
3/4 1-20 Full Length CEA Group Position....................... 3/4 1-20 Pos i ti on Indi cator Channel s..........................
3/4 1-24 CEA Drop Time........................................
3/4 1-26 Shutdown CEA Insertion Limit.........................
3/4 1-27 Regulating CEA Insertion Limits...................... 3/4 1-28 Control Rod Drive Mechanisms.........................
3/4 1-31 l
1 MILLSTONE - UNIT 2-IV Amendment No. 35,7pA,116 1
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REACTIVITY CONTROL SYSTEMS CONTROL ROD DRIVE MECHANISMS LIMITING CONDITION FOR OPERATION 3.1.3.7 The control rod drive mechanisms shall be de-energized.
APPLICABILITY: MODES 3*, 4, 5 and 6, whenever the RCS boron concentra-tion is less than refueling concentration of Specification
3.9.1. ACTION
With any of the control rod drive mechanisms energized, restore the mechanisms to their de-energized state within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or immediately open the reactor trip circuit breakers.
SURVEILLANCE REQUIREMENTS 4.1.3.7 The control rod drive mechanisms shall be verified to be de-energized at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
l
- The control rod drive mechanisms may be energized for MODE 3 as long as 4 reactor coolant pumps are OPERATING, the reactor coolant system temperature is greater than 500 F, the pressurizer pressure is greater than 2000 psia and the high power trip is operable.
MILLSTONE - UNIT 2 3/4 1-31 Amendment No.116 i
h 1.
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICABILITY: As shown in Table 3.3-1.
ACTION; As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the modes and at the frequencies shown in Table 4.3-1.
4.3.1.1.2 The logic for the bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
l 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.
4.3.1.1.4 The response time of all REACTOR TRIP SYSTEM resistance temperature I detectors (RTD) shall be verified to be less than or equal to the value specified in Table 3.3-2 within one month of operation for newly installed RTD's and once every 18 months thereafter.
MILLSTONE - UNIT 2 3/4 3-1 knendment No.7 2 4
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TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION 3
7-MINIMUM 5;
TOTAL N0.
CHANNELS CHANNELS APPLICABLE g
M FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION 1.
1 2
1, 2 and.*
1 2.
Power Level - High 4
2 (f) 3 1,2,3(d) 2 3.
Reactor Coolant Flow - Low 4
2(a) 3 1, 2 (e) 2 4.
Pressurizer Pressure - High 4
2 3
1, 2 2
5.
Containment Pressure - High 4
2 3
1, 2 2
Y 6.
Steam Generator Pressure - Low 4 2(b) 3 1, 2 2
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7.
Steam Generator Water Level - Low 4
2 3
1, 2 2
8.
Local Power Density - High 4
2(c) 3 1
2 9.
Thennal Margin / Low Pressure 4
2(a) 3 1, 2 (e) 2 k
- 10. Loss of Turbine--Hydraulic g.
Fluid Pressure - Low 4
2(c) 3' 1
3 a
W 0
n
TABLE 3.3-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION 2
v>
MINIMUM N
T6 TOTAL NO.
CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i
e
- 11. Wide Range Logarithmic Neutron Flux Monitor - Shutdown 4
0 2
3, 4, 5 4
m
- 12. Underspeed - Reactor 4
2(a) 3 1,2(e) 2 Coolant Pumps u
b Wb 3I e
a-1
i TABLE 3.3-1 (Continued) l TABLE NOTATION
- With the protective system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.
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(a) Trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 5% of RATED THERMAL POWER..
-(b) Trip may be manually bypassed below 600 psia; bypass shall be-l automatically removed at or above 600 psia.
(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 15% of RATED THERMAL i;-
POWER.
(d) Trip does not need to be OPERABLE if all the control rod drive mechanisms are de-energized or if the RCS boron concentration is
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greater than or equal to"the refabling: concentration of3Specifica-tion 3.9.1.
(e)- Trip may be bypassed during testing pursuant to Special Test Excep-tion 3.10.3.
(f) AT Power input to trip may be bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 5% of RATED THERMAL POWER.
ACTION STATEMENTS With the number of channels OPERABLE one less than required ACTION 1 by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and/or open the protective system trip breakers.
With the number of OPERABLE channels one less than the ACTION 2 Total Number of Channels and with the THERMAL POWER level:
a.
< 5% of RATED THERMAL POWER, imediately place the Tnoperable channel in the bypassed condition; restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5% of RATED THERMAL POWER.
b.
> 5% of RATED THERMAL POWER, operation may continue with the inoperable channel in the bypassed condition, provided the following conditions are satisfied:
MILLSTONE - UNIT 2 3/4 3-4 Amendment No. 9,38,72,116
TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS
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=
O5 CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE g
FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED I
[
1.
Manual Reactor Trip N.A.
N.A.
S/U(1)
N.A.
2.
Power Level - High a.
Nuclear Power S
D(2),M(3),Q M
1, 2, 3*
l b.
AT Power S
D(4),O M
1 3.
Reactor Coolant Flow - Low S
R M
1, 2 4.
Pressurizer Pressure - High S
R M
1, 2 5.
Containment Pressure - High S
R M
1, 2 6.
Steam Generator Pressure - Low S
R M
1, 2 7.
Steam Generator Water Level - Low S
R M
1, 2 p
8.
Local Power Density - High S
R M
1 m
2 5
9.
Thermal Margin / Low Pressure S
R M
1, 2
- z P
- 10. Loss of Turbine--Hydraulic d
Fluid Pressure - Low N.A.
N.A.
S/U(1)
N.A.
os I
TABLE 4.3-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
r-U CHANNEL MODES IN WHICH E
CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE
'," FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED E 11. Wide Range Logarithmic Neutron S
N.A.
S/U(1) 3, 4, 5 and
- M Flux Monitor ro
- 12. Underspeed - Reactor S
R M
1, 2 Coolant Pumps
- 13. Reactor Protection System Logic N.A.
N.A.
M and S/U(1) 1, 2
- 14. Reactor Trip Breakers N.A.
N.A.
M 1, 2 and
- i M
W b
$g
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0
REACTIVITY CONTROL SYSTEMS BASES 3/4.1.2 BORATION SYSTEMS (Continued)
The boron capability required below 200'F is based upon providing a 2% ak/k SHUTDOWN MARGIN at 140 F during refueling with all full and part length control rods withdrawn. This condition requires either 5,050 gallons of 6.25% boric acid solution from the boric acid tanks or 57,000 gallons of 1720 ppm borated water from the refueling water storage tank.
A minimum boron concentration of 1720 ppm is required in the RWST at all times in order to satisfy safety analysis assumptions for boron dilu-tion incidents and other transients using the RWST as a borated water source.
3/4.1.3 M0VABLE CONTROL ASSEMBLIES f
-The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident and an uncontrolled CEA withdrawal from subcriticality are limited to accept-able levels.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.
The ACTION statements applicable to an immovable or untrippable CEA and to a large misalignment (> 20 steps) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a potsible loss of mechanical functional capability of the CEAs and in the event of an immovable or untriopable CEA, the loss of SHUTDOWN MARGIN.
For small misalignments (< 20 steps) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the time dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION state-ment associated with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to restore the CEA to within its alignment requirements prior to initiating a reduction in THERMAL POWER.
The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution.
Overpower margin is provided to protect the core in the event of a large misalignment (> 20 steps) of a CEA. However, this misalignment would cause distortion of the core power distribution. The reactor MILLSTONE - UNIT 2 B 3/4 1-3 Amendment No. 38,67,72, 116
REACTIVIT!Y CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued) protective system would not detect the degradation in radial peaking factors and since variations in other system parameters (e.g., pressure and coolant temperature) may not be sufficient to cause trips, it is possible that the reactor could be operating with process variables less conservative than those assumed in generating LC0 and LSSS setpoints.
Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt and significant reduction in THERMAL POWER prior to attempting realignment of the misaligned CEA.
The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA.
Conformance with these alignment requirements bring the core, within a short period of time, to a configuration consistent with that assumed in generating LC0 and LSSS setpoints. However, extended operation with CEAs significantly inserted in the core may) lead to perturbations in 1) local burnup, 2) peaking factors and 3 available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.
Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.
Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit. The CEA " Full In" and " Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.
Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.
CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
The maximum CEA drop time permitted by Specification 3.1.3.4 is the assumed CEA drop time used in the accident analyses. Measurement with T,yg 1 515'F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
MILLSTONE - UNIT 2 B 3/4 1-4 Amendment No. 38
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l REACTIVITY CONTROL SYSTEMS BASES 3/4.1.3 MOVABLE CONTROL ASSEMBLIES (Continued)
The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the core operating in an essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during MODES 1 and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA insertion permitted by the Long Term Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded burnup assumption but will still provide sufficient reactivity control. The Transient Insertion Limits of Specification 3.1.3.6 are provided to ensure that (1) acceptable power distribution limits are raintained, (2) the minimum J
SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels; however, long term operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configura-y tion.
The control rod drive mechanism requirement of Specification 3.1.3.7 is provided to assure that the consequences of an uncontrolled CEA with-drawal from subcritical transient will stay within acceptable levels.
This specification assures that reactor coolant system conditions exist which are consistent with the plant safety analysis prior to energizing the control rod drive mechanisms. The accident is precluded when condi-tions exist which are inconsistent with the safety analysis since de-energized drive mechanisms cannot withdraw a CEA. The drive mechanisms may be energized with the boron concentration greater than or equal to the refueling concentration since, under these conditions, adequate SHUT-D0WN MARGIN is maintained,even if all CEAs are fully withdrawn from the Core.
MILLSTONE - UNIT 2 B 3/4 1-5 Amendment No. 38,116
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