ML20073S966
| ML20073S966 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 06/28/1994 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Northeast Nuclear Energy Co (NNECO) |
| Shared Package | |
| ML20073S967 | List: |
| References | |
| NPF-49-A-091 NUDOCS 9407070285 | |
| Download: ML20073S966 (14) | |
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i i fi I UNITED STATES 1[sg'..,al j
NUCLEAR REGULATORY COMMISSION g
j W ASHINGTON, O fl 20555-0001 NORTHEAST NUCLEAR ENEFGY COMPANY. ET AL.
DOCKET NO. M-421 MILLSTONE NUCLEAR POWEE_ STATION. UNIT NO. 3 AMENDMENT TO FACIllTY OPERATING LICENSE I
Amendment No. 91 License No. NPF-49 t-1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application ror amendment by Northeast Nuclear Energy Company, et al. (the licensee), dated February 10, 1992 as supplemented April 14, 1994 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; 8.
The facility will operate in tenfc with the application, the provisions of the Act, and the rulu.
regulations of the Commission; C.
,u e is re::r d !a assurance (i) that the activities authorized by
' a 'mendmant can be conducted without endangering the health and sa'e': of tie public, and (ii) that such activities will be g
con lui ed in compliance with the Commission's regulations; 0;
%.ssnnce of this amendment will rat be inimical to the common g
aefense and security or to the healti and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the 3mnission's regulations and all applicable requirements have been satisfied.
9407070285 940628 PDR ADOCK 05000423 P
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. 2.
Accordingly, the licens) is amended by changes to the Technical Specific.ations as indic*ttd in the attachment to this license amendment, and paragraph 2.C.(2) of facility Operating License No. NPF-49 is hereby amended to read as follows:
(2) Technical Snecifications The Technical Specifications contained in Appendix A, as ravised through Amendment No. 91
, and the Environmental Protection Plan contained in Appendix B both of wh;di are attached hereto are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Te.nnical Specifications and the Environmental Protection Plan.
3.
Th' license amendment is effective as of the date of its issuance, to be implemented within 30 days of 1:cuance.
FOR THE NUCLEAR REGULA ORY COMMISSION l
/L J6h F. Stu z, Director PrjectDirectorate1_)
4 j(D vision _of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 28. 1994
i ATTACHMENT TO LICENSE AMENDMENT NO. 91 FAClllTY OPERATING LICENSE NO, NPF-49 DOCKET NO. 50-411 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert v
v vi vi 3/4 3-1 3/4 3-1 3/4 3-8 3/4 3-8 3/4 3-9 3/4 3-9 3/4 3-15 3/4 3-15 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 B 3/4 3-2 B 3/4 3-2
1 Mil LINITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION EASE Position Indiction System -
Shutdown.....................
3/4 1-24 Rod Drop Time............................................
3/4 1-25 Shutdown Rod Insertion Limit.............................
3/4 1-26 Control Rod Insertion Limits............................
3/4 1-27 3/4.2 POWER D(LTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE..............................
3/4 2-1 Four Loops Operating.....................................
3/4 2-1 j
Three Loops 0perating....................................
3/4 2-3 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fo(Z).....................
3/4 2-5 Four Loops 0perating.....................................
3/4 2-5 Threr Loops 0perating....................................
3/4 2-12 3/4.2.3 RCS FLO4 RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACT04...................................................
J/4 2-19 i
Four Loops Operating...........,.........................
3/4 2-19 Threr Loo,1s 0perating....................................
3/4 2-22 3/4.2.4 QUADRANT POWER TILT RAT 10................................
3/4 2-24
?/4.2.5 DNB PA4AMETERS...........................................
3/4 2-27
_i TABLE 3.2-1 DH9 PARAMETERS........................................
3/4 2-28 3/4.3 INSTRUME([16IiQfi 3/4.3.1 REACT 0A TRIP SYSTEM INSTRUMENTATION......................
3/4 3 1 TABLE 3.3-1 REAC10R TitlP SYSTEM INSTRUMENTATION...................
3/4 3 2 TABLE 3.3-2 DELETED TABLE 4.3-1 REACTCR 1.llP SYSTEM INSTRUMENTATION SURVE1LLANCE 3/4 3 1o REQUIPEMruye.,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
3/4.3.2 ENGINEEP;it S/FETY FEATURES ACTUATION SYSTEM.
INSTRUMtkTAT10N..........................................
3/4 3 15 TABLE 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION..........................................
3/4 3-17 TABLE 3.3-4' ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETP0lNTS...........................
3/4.3-26 MILLSTONE - UNIT 3 y
Amendment No. pp. JP 77 91 0197
INDEX LIMITING C0WITION FOR OPERATION AM SURVEILLANCE REQUIREMENTS SECTION 1%L TABLE 3.3-5 DELETED TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......
3/4 3-36 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring for Plant Operations.......
3/4 3 42 TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS.................
3/4 3-43 TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS 3/4 3-45 Movable Incore Detectors 3/4 3-46 Seismic Inctrumentation................
3/43-47 TABLE 3.3-7 SEISMIC M0h!TORING INSTRUMENTATION..........
3 A 3 -48 TARLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.....................
3/4 3-49 Meteorological Instrumentation 3/4 3-50 TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION.......
3/4 3 51 TABLE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 3/4 3-52 Remote Shutdown Instrumentation............
3f4 3-53 TABLE 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION............
3/4 3-54 TABLE 4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............
3/4 3-58 l
Accident Monitoring Instrumentation..........
3/4 3-59 TABLE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION.........
3/4 3 60 l
TABLE 4.3-7 ACCIDENT MONITORING INSTP.'JMENTATION SURVEILLANCE REQUIREMENTS....................
3/4 3-62 TABLE 3.3-11 DELETED Loose-Part Detection System.............
3/4 3-68 Radioactive Liquid Effluent Monitoring Instrumentation 3/4 3-69 TABLE 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION...................
3/4 3-70 TABLE 4.3 8 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS......
3/4 3-72 Radioactive Gassous Effluent Monitoring Instrumentation 3/4 3-74 t
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MILLSTONE - UNIT 3 vi Amendment No. pf,91 em
3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRl5FWJTIQN Itulitus ensaitTinti rna nornartnu 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
canvrtt i auer nrnistorurne 4.3.1.1 Each Reactor Trip System instrumentation channel and 1. '.arlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
4.3.1.2 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.*
Neutron detectors and speed sensors are exempt from response time testing.
l Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel (to include input relays to both trains) per function such that all channels are tested at Ic.ast 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.
- Except that the surveillance requirement due no later than June 13, 1993, may be deferred until the next refueling outage, but no later than September 30, 1993, whichever is earlier.
MILLSTONE - UNIT 3 3/4 3-1 Amendment No. JJ, 77, 91 01H
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4 INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIQH LIMITING CONDITION FOR OPERATION 3.3.2 The Engineered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3 3 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3 4.
l APPLICABILITY: As shown in Table 3.3 3.
ACTION:
a.
With an ESFAS instrumentation er Interlock Trip Setpoint trip less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3 4, adjust the Setpoint consistent with the Trip Setpcint value, b.
With an ESFAS Instrumentation or Interlock Trip Setpoint less conservative than the value shown in the Allowable Value column of Table 3.3 4, either:
1.
Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3 4, and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.21 was satisfied for the affected channel, or 2.
Declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3 3 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.
Equation 2.2 1 Z + R + S s TA Where:
I The value from Column I of Table 3.3 4 for the affected
- channel, R=The'asmeasured"value(inpercentspan)ofrackerrorforthe affected channel, S = Either the 'as measured' value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3 4 for the affected channel, and TA = The value from Column TA (Total Allowance) of Table 3.3-4 for the affected channel.
c.
With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3 3.
MILLSTONE - UNIT 3 3/4 3-15 Amendment No.
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MILLSTONE - UNIT 3 3/43-35 Amendment No. /' 91 8879
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INSTRUMENTATION BASES REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) the sensor from its calibration point or the value specified in Table 3.3-4, in percent span, from the analysis assumptions. Use of Equation 3.3 1 allows for a sensor drift factor, an increased rack drift factor, and provides a threshold value for REPORTABLE EVENTS.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Tria Setpoints are the magnitudes of these channel uncertainties. Sensor and instrumentatioa utilized in these channels are expected to be capable of ric (
operating within the allowances of these uncertainty magnitudes. R6ck drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that there is a small statistical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is coms.".ated within the time limit assumed in the safety analyses. The RTS and ESF response times are included in the Operating Procedure OP 3273 " Technical Requirements--Supplementary Technical Specifications." Any changes to the RTS and ESF response times shall be in accordance with Section 50.59 of 10CFR50 and approved by the Plant Operations Review Committee. No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either:
(1) in place, onsite, or offsite test measurements, or (2) utilizing replacement sensors with certified response time. Detector response times may be measured by the in situ on line noise analysis-response time degradation method described'in the Westinghouse Topical Report, "The Use of Process Noise Measurements To Determine Re.e ose Characteristics of Protection Sensors in U.S. Phnts," August 1983.
l MILLSTONE - UNIT 3 83/43-2 Amendment No. ['91 wn l
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