ML20248B075
| ML20248B075 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 05/26/1998 |
| From: | Mckee P NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20248B080 | List: |
| References | |
| NUDOCS 9806010171 | |
| Download: ML20248B075 (21) | |
Text
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's NUCLEAR REGULATORY COMMISSION o
WASHINGTON, D.C. - "1
- ++
NORTHEAST NUCLEAR ENERGY COMPANY. ET AL DOCKET NO. 50-423 MILLSTONE NUCLEAR POWER STATION. UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 159 License No. NPF-49 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 October 15,1997, as supplemented January 23 and April 8, 1998, 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 ChapterI; 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 autnor!Ied 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.
9806010171 980526 PDR ADOCK 05000423 P
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2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2 C.(2) of Facility Operating License No. NPF-4g is hereby amended to read as follows:
(2)
Technical Soncifications i
The Technical Specifications contained in Appendix A, as revised through Amendment No.159, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of issuance, to be implemented within 60 days ofissuance.
FOR THE NUCLEAR REGULATORY COMMISSION
=
$ Phillip F. McKee Deputy Director for Licensing Special Projects Office Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of lasuance:
May 26, 1998 i
1 1
,e ATTACHMENT TO LICENSE AMENDMENT NO. 159 i
FACILITY OPERATING LICENSE NO. NPF-49 l
l t
DOCKET NO. SQ-d23 f
Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by amendment number and contain verticallines indicating the areas of change.
Remove Insert 2-4 2-4 2-5 2-5 24 24 2-7 2-7 2-8 2-8 2-10 2-10 2-12 2-12 B 2-3 B 2-3 B24 B24 B 2-4a
- 3/4 3-15 3/4 3-15 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4329 3/4 3-29 3/4 3-30 3/4 3-30 j
B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 l
- overflow page - no change 1
6 1
e SAFETY LIMITS AND LIMITING SAFETY SYSTEN SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Nominal Trip Setpoint values shown in Table 2.2-1.
l APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
a.
With a Reactor Trip System Instrumentation Channel or Interlock Channel Nominal Trip Setpoint inconsistent with the value shown in the Nominal Trip Setpoint column of Table 2.2-1, adjust the Setpoint consistent with the Nominal Trip Setpoint value.
b.
With a Reactor Trip System Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status.
MILLSTONE - UNIT 3 2-4 Amendment No.159 0550
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A 2.2 LINITING SAFETY SYSTEM SETTINGS BASES l
2.2.1 REACTOR TRIP SYSTEN INSTRUMENTATION SETPOINTS The Nominal Trip Setpoints specified in Table 2.2-1 are the nominal values at which the reactor trips are s'et for each functional unit. The l
Allowable Values (Nominal Trip Setpoints i the calibration tolerance) are considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to ensure that the core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.
The Setpoint for a Reactor Trip System or interlock function is considered to be consistent with the nominal value when the measured "as left" Setpoint is within the administrative 1y controlled ( ) calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged.
Measurement and Test Equipment accuracy is administrative 1y controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991. Operability determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty calculation.
The Allowable Value specified in Table 2.2-1 defines the limit beyond which a channel is inoperable.
If the process rack bistable setting is I
measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be operable.
The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels.
Inherent in the determination of the Nominal Trip Setpoints are the magnitudes of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance characteristics. Device drift in excess of the allowance that is more than occasional, may be indicative of more serious problems and would warrant further investigation.
The various reactor trip circuits automatically open the reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.
In addition to the redundant channels and trains, the design approach provides Reactor Trip System functional diversity. The NILLSTONE - UNIT 3 82-3 Amendment No. 159 0661
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l 2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued) functional capability at the specified trip setting is required for those anticipatory or diverse reactor trips for which no direct credit was assumed in i
the safety analysis to enhance the overall reliability of the Reactor Trip System. The Reactor Trip System initiates a turbine trip signal whenever reactor trip is initiated. This prevents the reactivity insertion that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation System.
Manual Reactor Trio The Reactor Trip System includes manual Reactor trip capability.
Power Ranae. Neutron Flux In each of the Power Range Neutron Flux channels there are two independent bistables, each with its own trip setting used for a High and Low Range trip setting. The Low Setpoint trip provides protection during subcritical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels. The High Setpoint trip is reduced during three loop operation to a value consistent with the safety analysis.
l The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.
Power Rance. Neutron Flux. Hiah Positive Rate The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.
1 1
1 NILLSTONE - UNIT 3 B2-4 Amendment No. J#,159 0551 l
2.2 LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Power Rance. Neutron Flur. Hiah Positive Rate
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The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.
Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for all rod ejection accidents.
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l MILLSTONE - UNIT 3 B 2 - 4a Amendment No. 117 159 0661
INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION LINITING CONDITION FOR OPERATION 3.3.2 The Er,31neered Safety Features Actuation System (ESFAS) instrumentation channels and interlocks shown in Table 3.3-3 shall be OPERABLE Setpoints set consistent with the values shown in the Nominal Trip Setpoint column of Table 3.3-4.
APPLICABILITY: As shown in Table 3.3-3.
ACTION:
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With an ESFAS Instrumentation Channel or Interlock Channel Nominal a.
Trip Setpoint inconsistent with the value shown in the Nominal Trip Setpoint column of Table 3.3-4, adjust the Setpoint consistent with the Nominal Trip Setpoint value.
b.
With an ESFAS Instrumentation Channel or Interlock Channel found to be inoperable, declare the channel inoperable and apply the applicable ACTION statement requirements of Table 3.3-3 until the channel is restored to OPERABLE status.
NILLSTONE - UNIT 3 3/4 3-15 Amendment No. 77,159 0552
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3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION
^
The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that:
1 the associated action and/or Reactor trip will be initiated when the par (am)eter monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed availat% in the facility design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.
The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
The Engineered Safety Features Actuation System Nominal Trip Setpoints specified in Table 3.3-4 are the nominal values of which the bistables are set for each functional unit.
The Allowable Values (Nominal Trip Setpoints i the calibration tolerance) are considered the Limiting Safety System Settings as identified in 10CFR50.36 and have been selected to mitigate the consequences of accidents. A Setpoint is considered to be consistent with the nominal value when the measured "as left" Setpoint is within the administrative 1y controlled ( )
calibration tolerance identified in plant procedures (which specifies the difference between the Allowable Value and Nominal Trip Setpoint). Additionally, the Nominal Trip Setpoints may be adjusted in the conservative direction provided the calibration tolerance remains unchanged.
Measurement and Test Equipment accuracy is administrative 1y controlled by plant procedures and is included in the plant uncertainty calculations as defined in WCAP-10991. Operability determinations are based on the use of Measurement and Test Equipment that conforms with the accuracy used in the plant uncertainty j
calculation.
The Allowable Value specified in Table 3.3-4 defines the limit beyond which a channel is inoperable. If the process rack bistable setting is measured within the "as left" calibration tolerance, which specifies the difference between the Allowable Value and Nominal Trip Setpoint, then the channel is considered to be operable.
MILLSTONE - UNIT 3 B 3/4 3-1 Amendment No.159 0555
4 3
INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
The methodology, as defined in WCAP-10991 to derive the Nominal Trip Setpoints, is based upon combining all of the uncertainties in the channels.
Inherent in the determination of the Nominal Trip Setpoints are the magnitudes j
l of these channel uncertainties. Sensors and other instrumentation utilized in these channels should be capable of operating within the allowances of these uncertainty magnitudes. Occasional drift in excess of the allowance may be determined to be acceptable based on the other device performance characteristics. Device drift in excess of the allowance that is more than
{
i occasional, may be indicative of more serious problems and would warrant further j
investigation.
i The above Bases do not apply to the two radiation monitors in the ESF Table (Item 3C and Item 7E).
For these radiation monitors the allowable values
)
are essentially nominal values. Due to the uncertainties involved in radiological parameters, the methodologies of WCAP-10991 were not applied.
Actual trip setpoints will be reestablished below the~ allowable value based on calibration accuracies and good practices.
The measurement for response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Features actuation associated with each channel is completed 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 Response Characteristics of Protection Sensors in U.S. Plants," August 1983.
l i
i NILLSTONE - UNIT 3 8 3/4 3-2 Amendment No. 7, JJ 159 l
. 0666