LR-N15-0124, Supplemental Information Needed for Acceptance of Requested Licensing Action Amendment Request Regarding Replacement of Source Range and Intermediate Range Neutron Monitoring Systmens
| ML15153A193 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 06/02/2015 |
| From: | Jamila Perry Public Service Enterprise Group |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LAR S15-02, LR-N15-0124 | |
| Download: ML15153A193 (31) | |
Text
PSEG Nuclear LLC P.O. Box 236, Hancocks Bridge, NJ 08038-0236 PSEG Nuclear JUN 1i2 2015 1 0 CFR 50.90 LR-N15-01 24 LAR S1 5-02 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Salem Nuclear Generating Station Units 1 and 2 Renewed Facility Operating License Nos. DPR-70 and 75 NRC Docket Nos. 50-272 and 50-31 1
Subject:
Supplemental Information Needed for Acceptance of Requested Licensing Action Re: Amendment Request Regarding Replacement of Source Range and Intermediate Range Neutron Monitoring Systems (TAG Nos. MF6065 and MF6066)
References 1. PSEG letter to NRC, "License Amendment Request Regarding Replacement of Source Range and Intermediate Range Neutron Monitoring Systems,"
dated April 3, 201 5 (ADAMS Accession No. ML15093A291 )
- 2. NRC letter to PSEG, "Salem Nuclear Generating Station, Unit Nos. 1 and 2 -
Supplemental Information Needed for Acceptance of Requested Licensing Action Re: Amendment Request Regarding Replacement of Source Range and Intermediate Range Neutron Monitoring Systems (TAG Nos. MF6065 and MF6066)," dated May 1 4, 2015 (ADAMS Accession No. ML15127A287)
In the Reference 1 letter, PSEG Nuclear LLC (PSEG) submitted a license amendment request for Salem Nuclear Generating Station (Salem), Unit,Nos. 1 and 2. The proposed amendment would revise technical specification (TS) 3/4.3. 1 , "Reactor Trip System Instrumentation," to support the planned replacement of the existing source range (SR) and intermediate range (IR) nuclear instrumentation.
In the Reference 2 letter, the U.S. Nuclear Regulatory Commission staff requested that PSEG supplement the application with information necessary to enable the NRC staff to begin its detailed technical review. The requested information is provided in Attachment 1.
PSEG has determined that the information provided in this submittal does not alter the conclusions reached in the 1 0 CFR 50.92 no significant hazards determination previously submitted. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
1 0 CFR 50. 90 Page 2 LR-N1 5-0124 There are no regulatory commitments contained in this letter.
If you have any questions or require additional information, please contact Mr. Brian Thomas at 856-339-2022.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on JUN 0 2 2015 (date}
Respectfully, CJrC? F 1?7-* _.
John F. Perry Site Vice Presiden Salem Nuclear Generating Station - Supplemental Information Needed for Acceptance of Requested Licensing Action Re: Amendment Request Regarding Replacement of Source Range and Intermediate Range Neutron Monitoring Systems (TAC Nos. MF6065 and MF6066) - Analysis of SC-NIS001 -01 and SC-NIS002-01 cc: Mr. D. Dorman, Administrator, Region I, NRC Ms. C. Parker, NRC Project Manager, Salem NRC Senior Resident Inspector, Salem Mr. P. Mulligan, Manager IV, NJBNE Mr. L. Marabella, Corporate Commitment Tracking Coordinator Mr. T. Cachaza, Salem Commitment Tracking Coordinator
LR-N1 5-01 24 Attachment 1 Supplemental Information Needed for Acceptance of Requested Licensing Action Re:
Amendment Request Regarding Replacement of Source Range and Intermediate Range Neutron Monitoring Systems (TAC Nos. MF6065 and MF6066)
LR-N 1 5-01 24
- 1. Provide the equipment qualification information for the detectors, amplifiers, and the control room electronics for harsh and mild environments.
a) For equipment in harsh environments, provide the qualification information that meet the guidance of Title 1 0 of the Code of Federal Regulations (1 0 CFR) Part 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants."
PSEG Response:
For the harsh environment, the Nuclear Instrumentation System (NIS) source and intermediate range detectors and associated cables inside containment were evaluated to the 1 OCFR50. 49 criteria above and were exempted from the Salem Equipment Qualification (EQ) Program requirements. The exempt classification status for the NIS detectors was submitted to the NRC as part of the design basis review for Salem EQ equipment. Based on this analysis, the source and intermediate range detectors in a harsh environment were exempted from the requirements of 1 0 CFR 50. 49 and not included in the Salem EQ program.
Although this equipment was exempted from the requirements of 10 CFR 50.49, PSEG Specification S-C-DE-NIS-021 0 states that the qualification of this NIS equipment shall be demonstrated in accordance with S-C-ZZ-SDC-1 419, "Salem Generating Station Environmental Design Criteria, " which requires testing per IEEE-323-1 974.
b) For equipment in non-harsh environments, provide the qualification information that protects the detectors, amplifiers, and the control room electronics, for example, by meeting the guidance of Regulatory Guide (RG) 1 . 209, "Guidelines for Environmental Qualification of Safety-Related Computer-Based Instrumentation and Control Systems in Nuclear Power Plants. " (ADAMS Accession No. ML070190294)
PSEG Response:
There are no computer based control components within the replacement N IS that are critical to performance of its intended functions (Refer to the response to Question 3 for more information). For components installed in a mild environment, the new replacement NIS equipment is being supplied to meet the mild environment criteria identified in PSEG Specification S-C-DE-NIS-021 0 as follows:
Normal Operating and Abnormal Conditions Parameter Normal Min Normal Max DBA Max Temperature 55°F 85°F 85°F Humidity 20% 90% 90%
Pressure 0. 1 25"WC 0. 1 25"WC 0. 1 25"WC Radiation, Gamma Rate N/A 0.25 Not Calculated (mr/hr)
Radiation, Gamma TID {r_l N/A 87. 7 884 LR-N 15-0124 Attachment 1 If generic qualification documentation is used, explain how the generic qualification meets the site-specific qualification requirements for the applicable location of the equipment.
PSEG Response:
As documented in PSEG Specification S-C-DE-NIS-0210, if previous generic qualification documentation is provided by the vendor, the vendor is to provide a summary report or certificate of compliance for all testing performed. This documentation will be reviewed as part of the design change process to ensure it envelopes the PSEG criteria.
- 2. Please document how the equipment meets the guidance of RG 1.180, "Guidelines for Evaluating Electromagnetic and Radio-Frequency Interference in Safety-Related Instrumentation and Control Systems," (ADAMS Accession No. ML003740218) for electromagnetic and radio-frequency interference, or provide justification for meeting an alternative.
PSEG Response:
The replacement NIS components provided by Thermo Fisher Scientific are equipped with design features to limit their susceptibility to EMI/RFI. The proposed replacement NIS will use only analog and logic components. Linear power supplies will be used due to the low power requirements of the simple electronics. These linear power supplies have no oscillating components and a very high immunity to EMI/RFI. Bistable circuits are filtered to limit spurious effects. Bistable outputs are also isolated from the other cabinets. The logic outputs are electrically isolated from other circuits.
Testing, qualification, and analysis of the replacement NIS is in accordance with EPRI TR-1 02323, Guidelines for Electromagnetic Compatibility Testing of Power Plant Equipment, Rev 2. As documented in RG 1.180, Revision 1, the NRC staff accepted the Electric Power Research Institute (EPRI) topical report TR-102323, in a Safety Evaluation Report (SER) by letter dated April 17, 1996, as one method of addressing issues of electromagnetic compatibility (EMC).
- 3. Please state whether or not equipment uses software or embedded electronic components with software. If such components are used, then document how the equipment meets the software common cause failure guidance in Branch Technical Position (BTP) 7-19, "Guidance for Evaluation of Diversity and Defense-in-Depth in Digital Computer-Based Instrumentation and Control Systems," (ADAMS Accession No. ML070380094) or provide justification for meeting an alternative.
PSEG Response:
The only microprocessor based device associated with the replacement NIS is the Shutdown Margin Monitor (SDMM). The remaining portion of the NIS system is analog.
The shutdown margin monitor continuously measures and displays the count rate from the neutron flux monitor. The indication and alarm function provided by the SDMM is located in the Control Equipment Room. It is separate and independent from the audible count rate indication provided in the Main Control Room. The SDMM does not provide a function that
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LR-N15-01 24 The power to the SDMM assembly is via a fuse in the SDMM that provides the isolation between the SDMM and the power source associated with the Source Range (SR) Monitor.
The circuit isolation between SDMM and SR is provided by two relays. The SDMM provides
+/-1 2v to LEOs for local indication of the status of the instrument power and does not interface with any other components. The SR signal processor provides a pulse signal to the SDMM through an isolation device in the SR channels. Therefore, there are no failure modes associated with SDMM that could affect the ability of the SR channels to perform their intended function.
- 4. Please provide the setpoint calculation or the following additional information pertaining to the setpoint calculation:
a) Limiting trip setpoint for Source Range (SR) and Intermediate Range (IR).
PSEG Response:
The Setpoint Analysis (Enclosure 1 ) describes calculation of the Source Range and Intermediate Range trip setpoints and allowable values. The SR and IR setpoints are maintained at the existing Technical Specification setpoint values. The SR and IR trip functions do not have an analytical limit since they are not credited in the Salem accident analysis as discussed in response to Question 7.
b) Drift over the calibration period and the length of the calibration interval.
PSEG Response:
The equipment vendor (Thermo Fisher) provided a 24 month rack drift of +/-1 . 0 % of span. The data provided by Thermo Fisher is included in the Setpoint Analysis (Attachment 1 of Enclosure 1 ). The current calibration interval for the Source Range and Intermediate Range systems is 1 8-months in accordance with the Surveillance Frequency Control Program.
c) Please provide the basis for changing the allowable values for intermediate range neutron flux from :::; 30% to :::; 38.5% of rated thermal power and for the source range 5 5 neutron flux allowable value from :::; 1 . 3 x 1 0 to :::; 1 . 44 x 10 counts per second.
PSEG Response:
° 6 1 6 The source range indication scale changes from 10 - 1 0 cps to 1o- - 1 0 cps increasing the span from six to seven decades. The intermediate range indication scale 11 3 6 changes from 1 0" - 1 0" amps (equivalent to 10" to 1 20% Rated Thermal Power) to 1 o 200% Rated Thermal Power.
The total rack uncertainty increases as span increases, requiring a change in the allowable value. As described in the Setpoint Analysis (Enclosure 1 ), the allowable values for the Source Range and Intermediate Range neutron flux were changed to account for an increased rack uncertainty. The increased rack uncertainty is the product of the increased instrument span.
LR-N1 5-01 24
- 5. Item B on top of page 5 of attachment 1 to the license amendment request states, "Due to the changes in the IR detector output and units, an assessment was completed to verify adequate coordination between the P-6 setpoint and the SR neutron flux reactor trip setpoint for the Thermo Scientific instrumentation. " Please provide a summary of this assessment for NRC staff review.
PSEG Response:
Section 3.7.1 . 3 of Enclosure 1 provides a discussion of the P-6 Interlock. Figure 1 of Enclosure 1 provides a graphical representation of the margin between the P-6 setpoint and the SR neutron flux reactor trip setpoint.
- 6. The license amendment request does not include all the Applicable Regulatory Requirements/Criteria. Please address the requirements and criteria listed below. Salem Updated Final Safety Analysis (UFSAR) Section 3. 1. 1 , "Conformance with AEC [Atomic Energy Commission] General Design Criteria (July 1 971 )," states that the Salem Plant design conforms with the intent of the General Design Criteria (GDC) for nuclear power plants with the exceptions listed.
a) Please address how Salem meets the intent of GDC 1 , GDC 4 (with exceptions listed in UFSAR Section 3.1.1 ), GDC 13, and GDC 19, as it applies to this change request.
PSEG Response:
Salem's conformance with the AEC General Design Criteria (July 1971 ) is described in UFSAR Section 3. 1 . 3.
GDC 1 requires structures, systems, and components important to safety to be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The replacement SR and IR instrumentation is safety related. The design, procurement, installation, and operation of the replacement instrumentation is subject to the requirements of the PSEG quality assurance program.
GDC 4 requires structures, systems, and components important to safety to be designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents. The replacement instrumentation is designed to function in the environmental conditions associated with normal operation, maintenance, testing, and the postulated accidents for which it is credited.
GDC 1 3 requires instrumentation to be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges. Following implementation of the proposed changes, Salem will continue to meet the intent of GDC 13.
GDC 1 9 requires that a control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a LR-N1 5-01 24 safe condition under accident conditions, including loss-of-coolant accidents. Equipment at appropriate locations outside the control room shall be provided (1 ) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures. Following implementation of the proposed changes, Salem will continue to meet GDC 1 9.
b) Please address 1 0 CFR 50.36, "Technical Specifications," and 1 0 CFR 50.49.
PSEG Response:
The reactor trip system instrumentation satisfies Criterion 3 of 1 0 CFR 50.36(c)(2)(ii).
As discussed in response to Question 1 , the SR and IR were exempted from 1 0 CFR 50.49.
c) In addition, RG 1 .1 05, RG 1 .1 80, and BTP 7-1 9, should be addressed, as applicable.
PSEG Response:
RG 1 . 1 05, Instrument Setpoints For Safety-Related Systems- Revision 2 of RG 1 . 1 05 endorsed ISA S67.04-1 982 for use in establishing and maintaining setpoints in safety related systems. The methodology used to calculate the Salem trip setpoints is consistent with ISA-S67.04-1 982.
RG 1 . 1 80, Revision 1 . "Guidelines For Evaluating Electromagnetic And Radio Frequency Interference In Safety-Related Instrumentation And Control Systems"- As discussed in response to Question 2, testing, qualification, and analysis of the replacement NIS is in accordance with EPRI TR-1 02323, "Guidelines for Electromagnetic Compatibility Testing of Power Plant Equipment. "
BTP 7-1 9, "Guidance for Evaluation of Diversity and Defense-in-Depth in Digital Computer-Based Instrumentation and Control Systems, "- As discussed in the response to Question 3, the only microprocessor based device associated with the replacement NIS is the Shutdown Margin Monitor (SDMM). There are no failure modes associated with SDMM that could affect the ability of the SR channels to perform their intended function.
- 7. Page 3 of Attachment 1 indicated that "No credit is taken for the reactor trips associated with either the SR or IR channels in the accident analyses described in Chapter 1 5 of the Salem UFSAR." However, SR readings and/or its associated reactor trip signals are credited for some accident analysis at some plants. Please provide information addressing why UFSAR Chapter 1 5 analysis for each of the events (including a Boron dilution event) will not be affected by the proposed TS changes. A list of reactor trips credited for each event confirming that the SR and IR instrumentation have no role in the transient analyses would help address this.
LR-N 1 5-01 24 PSEG Response:
As discussed in the Salem TS Section 2.0 bases:
The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup. These trips provide redundant protection to the low setpoint trip of the Power Range, Neutron Flux channels. The Source Range Channels will initiate a 5
reactor trip at about 1 0+ counts per second unless manually blocked when P-6 becomes active. The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-1 0 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.
No SR or IR trip functions are listed in UFSAR Table 1 5.1 -3 which provides the limiting trip setpoints assumed in the Chapter 1 5 accident analyses. The table below lists the reactor trips credited for each event.
UFSAR Accident Reactor Trip Initiator Section 1 5.2.1 Uncontrolled RCCA Bank Power Range High Neutron Flux (low setting)
Withdrawal from a Subcritical Condition 1 5. 2.2 Uncontrolled Rod Cluster Power Range High Neutron Flux (1 1 8% of Control Assembly Bank Nominal Full Power)
Withdrawal at Power Overtemperature LlT Overpower LlT 1 5.2. 3 Rod Cluster Control Assembly Reactor trip is not credited Misalignment 1 5. 2.4 Uncontrolled Boron Dilution Operator Action based upon Audible Count Rate*
1 5. 2. 5 Partial Loss of Forced Reactor Low Flow in Any Loop ( >36% Power)
Coolant Low Flow in Any Two Loops (Between 1 1 %
and 36% Power) 1 5. 2. 6 Deleted from the UFSAR N/A 1 5.2.7 Loss of External Electrical Turbine Trip Load and/or Turbine Trip High Pressurizer Pressure High Pressurizer Water Level Low-Low Steam Generator Water Level Overpower LlT Overtemperature LlT No Direct or Immediate Trip Credited For a Loss of External Electrical Load without Turbine Trip.
1 5.2.8 Loss of Normal Feedwater Low-Low Steam Generator Water Level 1 5.2. 9 Loss of Offsite Power to the Low-Low Steam Generator Water Level Station Auxiliaries LR-N15-0124 UFSAR Accident Reactor Trip Initiator Section
- 15. 2. 10 Excessive Heat Removal Due Power Range High Neutron Flux to Feedwater System Overtemperature nT Malfunctions Overpower nT 15.2. 11 Excessive Load Increase Power Range High Neutron Flux Incident Overtemperature nT Overpower nT
- 15. 2.12 Accidental Depressurization of Pressurizer Low Pressure the Reactor Coolant System Overtemperature nT.
- 15. 2. 13 Accidental Depressurization of Overpower Reactor Trips (Neutron Flux and the Main Steam System nT)
- 15. 3. 1 Loss of Reactor Coolant from Pressurizer Low Pressure Small Ruptured Pipes or from Cracks in Large Pipes which Actuates the ECCS 15.3.2 Minor Secondary Pipe Breaks No Reactor Trip
- 15. 3. 3 Inadvertent loading of a Fuel No Reactor Trip Assembly into an Improper Position
- 15. 3. 4 Complete Loss of Forced Reactor Coolant Pump Bus Undervoltage Reactor Coolant Flow Low Primary Coolant Loop Flow ( >36%
Power)
Low Flow in Any Two Loops (Between 11%
and 36% Power
- 15. 3.5 Single Rod Cluster Control Overtemperature nT Assembly Withdrawal at Full Power 15.3.6 Accidental Release of Waste No Reactor Trip Gases
- 15. 3.7 Accidental Release of No Reactor Trip Radioactive Liquids 15.4.1 Major Reactor Coolant Pressurizer Low Pressure System Pipe Ruptures (Loss-of-Coolant Accident}_
- 15. 4.2 Major Secondary System Pipe The overpower reactor trips (neutron flux and Rupture nT) and the reactor trip occurring in conjunction with receipt of the safety injection signal.
LR-N1 5-01 24 UFSAR Accident Reactor Trip Initiator Section 1 5.4.3 Major Rupture of a Main a. High pressurizer pressure, Feedwater Line b. Overtemperature b.T,
- c. Low-low steam generator water level in any steam generator,
- d. Safety injection signals from any of the following:
1 . High steam flow coincident with low steam line pressure
- 2. High containment pressure
- 3. High steam line differential pressure
- 4. Low pressurizer pressure
- 5. High steam flow coincident with low-low T-avg 1 5. 4. 4 Steam Generator Tube Pressurizer Low Pressure Rupture Overtemperature b.T 1 5. 4.5 Single Reactor Coolant Pump Low Reactor Coolant Flow Locked Rotor and Reactor Coolant Pump Shaft Break 1 5. 4. 6 Fuel Handling Accident No Reactor Trip 1 5.4.7 Rupture of a Control Rod Power Range High Neutron Flux Drive Mechanism Housing Power Range Neutron Flux High Positive Rate (Rod Cluster Control Assembly Ejection)
- For the uncontrolled boron dilution event, UFSAR section 1 5.2.4 notes that the operator has prompt and definite indication of any boron dilution from the audible count rate instrumentation. The Source Range and Intermediate Range upgrades do not affect the audible count rate instrumentation. The UFSAR also states that high count rate is alarmed in the reactor containment and the Control Room. The UFSAR does not provide for a High Count Rate alarm setpoint.
PSEG previously submitted a license amendment request (Reference 1 ) requiring isolation of unborated water sources and to allow use of Gamma-Metrics Post-Accident Neutron Monitors during Mode 6 (Refueling). The proposed changes in the Reference 1 LAR would permit the replacement of the SR channels in Mode 6; however, approval of this NIS SR and IR replacement LAR is not contingent on approval of the Reference 1 LAR. Absent approval of the Reference 1 LAR, the SR instrumentation replacement could only be performed In Mode 6 with core alterations suspended or with the Reactor defueled. The replacement of the Westinghouse source range detectors with the Thermo-Scientific source range detectors will continue to satisfy the current TS Mode 6 instrumentation requirements.
References
- 1. PSEG letter to NRC, " License Amendment Request to Isolate Unborated Water Sources and Use Gamma-Metrics Post-Accident Neutron Monitors during Mode 6 (Refueling), ' "
dated March 9, 201 5 (ADAMS Accession No. ML15068A359)
LR-N15-01 24 Enclosure 1 Analysis of SC-NIS001 -01 and SC-NIS002-01
Analysis of SC-NISOO 1-01 and SC-NIS002-0 1 Analysis in Support of LAR S15-002 - SR/IR Replacement PSEG Review by:
(print name)
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. _____
- Date: o4u--
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1.0 Introduction The new Thermo Fisher Scientific signal processors, herein designated as Rack, are designed to replace the existing source and intermediate range drawers of the Nuclear Instrumentation System (NIS) . The processor will combine the functions of the existing Source Range Neutron Flux (SR) and Intermediate Range Neutron Flux (IR) drawers into a single drawer. Therefore, where required, calculations are performed to reflect the new instrumentation specifications performance including: (1) Total Loop Uncertainty (TLU),
(2) As Left Tolerance (ALT) and As Found Tolerance (AFT)/Allowable Value (AV), and its effect in (3) Nominal Trip Set Points (NTSP) .
The calculations use a statistical combination method (squae root of the sum of the squares) in references 4. 2 and 4.5 for independent random terms, and algebraic combination for dependent terms and bias effects, to determine the uncertainties. The calculations in percent of span is percent of Equivalent Linear Full Scale (% ELFS, %
span).
2.0 Source Range ,
\,_
The source range indication scale will change from W-1a yP$ to 1a(1a6 cps, that is, the source range scale changes from six to seven decades.' * ' * * * *
- 2. 1 Design Inputs Technical Specifications Surveillance Requirements Trip Setpoint Allowable Value Source Range ::; 1.a x 1as counts ::; 1.3 x 1 as counts Neutron Flux Rack Design Inputs (Ref. 4.3, Attachment 1):
Rack Accuracy (RCA source) = +/- 2. aa % span Rack Measurement &Test Equipment (RMTE_source) = +/- a . a 5 % span Rack Comparator Accuracy (RCSA_source) = +/- a 25% span Rack Temperature Effects (RTE source) = +/- a.aa% span (RCA includes temperature effects)
Rack Drift (RD _source) = +/- 1.aa% span S pan Scale Inputs Span (Number of Decades) = a.1- 1a6 cps (7 decades)
Process D esign Inputs Process Measuring Uncertainty (PMA_source) (Note) = +/- 1a.aa% span Note: Per reference 4. 2, Section 7.1, Process Measurement Uncertainty consists of water density variations in the downcomer due to Reactor Coolant temperature Page 1 of 18 3/11/2015
variations, radial power redistributions, detector Drift, Sensor Calibration Tolerances and Sensor Temperature Effects. These are combined such that the uncertainty is random and independent and is equal to+/- 10% span.
2.2 Rack Uncertainty (Rack_source)
Source Range Neutron Flux Rack Uncertainty is the statistical combination of the rack's random independent accuracy terms:
Rack_source= (RCA_source2 + RMTE_source2 + RCSA_source2 + RTE_source2+ RD_source2) 112 Rack_source= +/- 2.25% span 2.3 Total Loop Uncertainty (TLU_source)
The Rack Total Loop Uncertainty (TLU) is determined statistically combining rack uncertainty and the process measurement effects.
TLU_source= +/- (Rack_source2 + PMA_source2) 112 TLU_source= +/- 10.25% span 2.4 As Found Tolerance The As Found Tolerance, also termed Allowable Value, is determined on instrumentation based performance, TSTF-493, which is defined by the combination of the rack uncertainty terms comprising Accuracy, MTE, and Drift terms.
AFT_Rack Source= +/- (RCA Source2 + RMTE Source2 + RCSA Source2 + RD Source2 ) 112 AFT_Rack_source= +/- (2.002 + 0.052 + 0.252 + 1.002) 112 AFT_Rack_source= +/- 2.25% span 2.5 As Left Tolerance The As Left Tolerance, is determined on instrumentation based performance, TSTF-493, and is defined by the combination of the Accuracy and MTE terms.
ALT_Rack_source= +/- (RCA_sourc/ + RMTE_source2 + RCSA_source2 ) 112 ALT_Rack_source= +/- (2.002 + 0.052 + 0.252 ) 112 ALT_Rack _source= +/- 2.02% span Page 2 of 18 3/11/2015
- 2. 6 Setpoint Analysis In this section a review of the replacement system accuracies in cps, is performed per below equation.
Setpoint (cps) = Process Value (cps) X 1 o(CU %/iOO% X b) where:
a= base decade operation (-1 )
b= number of decades of operation (7)
CU = channel uncertainty (in % span)
The Technical Specifications setpoints are set in process values, or counts per second for the logarithmic base 1 0 scale; therefore, the calculated values are applied to the trip setpoint and converted into counts per second and percent of linear scale for a numerically familiar verification. The conversion is made as per below equation:
Percent Linear Scale= 1 00 x (log1o(cps) - a) 17
- 2. 6 . 1 Instrumentation Setpoints As Found Tolerance ITS Allowable Value The determination of the As Found Tolerance, or Allowable Value, for the Technical Specifications setpoint by applying the AFT_Rack_source uncertainty is:
[( . % / 00%)x7]
+ AFT= 1 X 1 05 X 1 0 225 1
+ AFT= 1 . 44 x 1 05 cps
- AFT= 1 X 105 X 1 0 [( -2.25 % /100%)x7]
- AFT = 6 . 96 x 1 04 cps In percent of linear scale Ivolts:
+ AFT= [log101 4 4 x 1 05 - (-1 )] 17
+ AFT= 8 7. 96% or 8 . 796 volts (note) 5 '
- AFT= [log10 6.96 x 1 0 - (-1)] I7
- AFT= 8 3. 46% or 8. 346 volts (note)
Note: Per figure 1 , the 0%-1 00% linear scale is equivalent to 0-1 0 volts.
Page 3 of 18 3/11/2015
Where the current TS Allowable Value of 1.3 x 105 cps in percent of linear scale is:
TS_AV_current = log1o[1.30 X 105 - (-1)] I 7 TS_AV_current = 87. 34%
The calculated value, is slightly larger than the TS value by a difference of 0.62%
(87.96% - 87.34%). See section 2.7 for conclusions.
2.6. 2 As Left Tolerance The determination of the As Left Tolerance is made by applying the ALT_Rack_source uncertainty:
+ ALT = 1 X 105 X 10 [(2.02%/100%)x7]
+ ALT = 1.38 x 105 cps
[(-2.02% /iOO%) X?]
- ALT = 1 X 105 X 10
- ALT = 7. 22 x 104 cps The positive ALT in percent linear scale:
+ ALT= log1o[1. 38 x 1 05- (-1)] I 7
+ ALT = 8 7.73%
- 2. 6 . 3 Analytical Limit There is not a defined Analytical Limit (AL); therefore, the Analytical Limit is set as the 100% scale, upper range value of the system, or 1 x 106 cps, and it is verified that the setpoint(s) does not fall outside of the system range.
Calculated Setpoint = AL - TLU_source and Margin:
Margin = Calculated Setpoint - TS Setpoint The setpoint calculated in Percent Scale is:
Calculated_Setpoint = 1 00% span - 10.25% span Calculated_Setpoint = 89.75% span Margin between TS Setpoint and Calculated Setpoint:
Margin = Calculated Setpoint- Setpoint where current TS setpoint of 1 x 105 cps in percent linear scale is:
Page 4 of 18 3/11/2015
TS_SPT_c urrent = log1o[(1 X 1 05 - (-1 )] / 7 TS_SPT_c urrent = 8 5.71 %
Thus the margin is:
Margin= 8 9.75% - 85. 71 %
Margin = + 4.04% span There is a positive margin from the TS setpoint to the calculated setpoint, and to the Analytical Limit; therefore, the current TS setpoint is acceptable.
2.7 Conclusions The current allowable value of S 1.30 x 105 cps should be updated to the new rack performance by the calculated AFT of S 1.44 x 105 cps.
3.0 Intermediate Range Neutron Flux Instrument Uncertainty The Intermediate Range scale will change from 1 x 1 o-11 to 1 x -1 o- 3 amperes (equivalent to 1 x 1 0-6to 1 20% RTP power, Figure 2) to 1 x 1 0-8 to 1 x 1 02*30103% (200%) RTP power, the change increases the intermediate range. Per observation of the systems scales correlation, Figure 1 , the Intermediate Range has 5 decades overlap with the Source Range system scale.
- 3. 1 Design Inputs Technical Specifications Surveillance Requirements Table 2. 2-1 Trip Setpoint Allowable Value Intermediate Range s 25% of Rated s 30% of Rated Neutron Flux Thermal Power Thermal Power Table 3.3-1 Condition and Setpoint Function P-6 With 2 of 2 Intermediate Range P-6 prevents or defeats Range Neutron Flux Channels the manual block of
<6 x 1 o-11 amps (*) source range reactor trip
(*) 6 x 1 o-11 amps in the intermediate range monitor is the P-6 reset allowable value for the bistable reset value of7 x 1o-11 amps Rack Design Inputs (Ref. 4.3, Attachment 1):
Rack Accuracy (RCA_Inter_Range) =+/- 1 .50% span Rack Measurement & Test Equipment (RMTE_Inter_Range) =+/- 0.05 % span Rack Comparator Accuracy (RCSA_inter_Range) =+/- 0.25% span Rack Temperature Effects (RTE_lnter_Range) =+/- 0.00% span Page 5 of 18 3/ 11/2015
(RCA includes temperature effects)
Rack Drift (RD_lnter_Range) = +/- 1.00% span Span Scale Inputs Span (Number of Decades) = 1 o-s - 1023 (200) %
(10.3 decades)
Process Design Inputs Process Measuring Uncertainty (PMA_lnter_Range) (Note) = +/-5.558% RTP random
-8.240% RTP bias Note:
Per reference 4.5, Section 5.2.1, the four PMA terms calculated/allowed for the Intermediate Range channels are:
- 1) Power Calorimetric+/- 2.0% RTP
- 2) Downcomer Temperature+/- 3.3% RTP
- 3) Radial Power Distribution+/- 4.0% RTP
While the Rod shadowing PMA was determined at 35% it will be applied to the 25%
setpoint to determine the magnitude of the effect in the calibrated setpoint value.
Terms (1), (2) and (3) are considered independent of each other and (4) is considered as an independent parameter. The total uncertainty for these four parameters is the SRSS of their magnitudes for the three random components and adding the bias term, or:
(2.0% 2 + 3.3% 2 + 4.0% 2 f2 - 8.24% = +/- 5.558% RTP (random) I- 8 .240% RTP (bias) 3.2 Rack Uncertainty (Rack_tnter_Range)
Intermediate Range Neutron Flux Rack Uncertainty is the statistical combination of the rack's random independent accuracy terms:
Rack-Inter-Range = (RCA Inter Range2 + RMTE-Inter Range2 + RCSA-Inter-Range2 + RTE-Inter Range2+
+ RD-nter_Range2) 112 Rack_lnter_Range= +/- 1.8 2% span 3.3 Process Measurement Uncertainty (PMA_Inter_Range)
The Technical Specifications setpoints are set in % RTP logarithmic base 10 scale; therefore, the calculated values are applied to the trip setpoint and converted into percent of linear scale below for a numerically familiar verification. The conversion is made as per below equation:
Percent Linear Scale= [ log10(% RTP- (-8 )] I 10.3 Page 6 of 18 3/11/2015
25% RTP setpoint converted to percent span linear scale:
S etp oint_25%_%_Span= [ log1o 25.00- (- 8) ] I 1 0. 3 Setpoint_25%_%_Span= 91 . 24%
The PMA in linear scale is calculated by adding and subtracting the PMA in % RTP from the 25% RTP setpoint; therefore:
PMA_inter_Range =+/- 5. 558% RTP random I- 8. 240% RTP bias Setpoint plus + PMA Random:
+ Setpoint_25%_%_PMA_Random = 25% RTP + 5.558 % = 30. 558% RTP In percent linear scale:
+Setpoint_25%_%_PMA_Random_Span = [ log1o 30.558 - ( -8 ) ] I 1 0. 3
+ Setpoint_25%_%_PMA_Random_Span = 92.09%
Setpoint minus - PMA Random:
- Setpoint_2s%_%_Span_Random = 25% RTP - 5. 558% = 1 9.442%
In percent linear scale:
- Setpoint_25%_%_PMA_Random_Span = [ log1o 1 9.442 - (-8) ] I 1 0. 3
- Setpoint_25%_%_PMA_Random_Span = 90. 1 8%
And the random PMA effect on the setpoint value in linear scale is:
PMA/R+ = 92. 09% - 91 . 24% = + 0.85%
PMA/R=90.18%- 91 .24% =-1.06%
Setpoint plus - PMA Bias:
Setpoint_2s%_%_PMA_Bias = 25% RTP- 8 . 240% = 1 6.760% RTP In percent linear scale:
Setpoint_25%Yo_PMA_Random_Bias = [ log1o 1 6.760 - (- 8) ] l 1 0. 3
+Setpoint_25%_%_PMA_Random_Span = 8 9. 56%
And the bias PMA effect on the setpoint value in linear scale is:
PMA/B- = 89. 56% - 91 . 24% =-1 . 68%
Page 7 of 18 3/11/2015
3.4 Total Loop Uncertainty (TL U_inter_Range)
The Rack Total Loop Uncertainty (TLU ) is determined combining statistically the rack random uncertainty with the process random measurement effects, and resultant random effects combination is algebraically combined with the process bias effects. The combination of the rack uncertainties with the PMAs is performed in percent of linear scale, therefore, the rack uncertainty effect in the setpoint is first converted to percent of linear scale.
Setpoint plus + Rack U ncertainty Random:
Setpoint in linear scale from section 3.3 is:
Setpoint_ %_%_Span = 91. 24%
25 PMAs effects in linear scale from section 3.3 are:
PMA/R+ = + 0.85%
PMA/R- = -1.06%
PMA/B-= -1.68%
Total rack uncertainty from section 3.2 is:
Rack_ lnter_Range= +/- 1. 82% span The total loop uncertainty is:
+ TLU-Inter-Range = + (Rack-Inter-Range2 + PMA/R+ 2) 112
+ TLU_lnter_Range = + 2.01% span
- TLU_lnter_Range = - [ Rack_ lnter_Range 2 + PMA/R 2] y, + PMA/B-
- TLU _lnter_Range= - [1. 822 +0.852] y, - 1.68
- TLU _lnter_Range = - 3.69% span Page 8 of 18 3/11/2015
3.5 As Found Tolerance The As Found Tolerance, also termed Allowable Value, is determined on instrumentation based performance, TSTF-493, which is defined by the combination of the Rack Uncertainty Terms comprising Accuracy, MTE, and Drift terms.
2 2 AFT_Rack-Inter-Range= +/- (RCA-Inter Rang/ + RMTE-Inter Range + RCSA-Inter Range +
- 2
+RD_lnter_Range )1 /2 - -
AFT_Rack_lnter_Range= +/- (1.502 + 0. 052 + 0.252 + 1.002 ) 112 AFT_Rack_lnter_Range= +/- 1.82% span
- 3. 6 As Left Tolerance The As Left Tolerance, is determined on instrumentation based performance, TSTF-493, and is defined by the combination of the Accuracy and MTE terms.
2 2 2 12 AL_Rack_lnter_Range= +/- (RCA_Inter_Range + RMTE_Inter_Range +RCSA_Inter_Range )1 AL_Rack_lnter_Range= +/- (1 .502 + 0. 052 + 0. 252) 112 AL_Rack_lnter_Range= +/- 1.52% span
- 3. 7 Setpoint Analysis In this section a review of the replacement system accuracies in percent RTP logarithmic, is performed per below equation.
oo Setpoint (% RTP) =Process Value ( %) x 10(CU%t1 %xb) where:
a= base decade operation (-8) b= number of decades of operation (1 0.3)
CU = channel uncertainty (in % span)
The Technical Specifications setpoints are set in % RTP logarithmic base 10 scale; therefore, the calculated values are applied to the trip setpoint and converted into percent of linear scale below for a numerically familiar verification. The conversion is made as per below equation:
Percent Linear Scale = 1 00 x log10[% R TP - (-8)] I 10. 3 3.7.1 Instrumentation Setpoints
- 3. 7 .1 .1 Trip Setpoint
- a. As Found Tolerance ITS Allowable Value Page 9 of18 3/11/2015
The determination of the As Found Tolerance, or Allowable Value, for the Technical Specifications setpoint by applying the AFT_Rack_rnter_Range uncertainty is:
+ AFT= 25% X 10 [1.82%/100%) X 10.3]
+ AFT= 38. 5% RTP o%)x10.3J
_AFT= 25% X 10 [-1.82%t1o
-AFT= 16 . 2% RTP Where the current TS Allowable Value is 30% percent of logarithmic scale. The calculated value is larger than the TS value by a difference of 8.5%.
The positive AFT in percent linear scale:
+ AFT= [ log1o 38.5 - (-8) ] I 10.3
+ AFT= 93. 063%
There is margin of 106.937% (200% - 93. 063%) to the upper range of the scale of 200%;
therefore, the calculated value is found acceptable. However, analysis of the setpoint with channel total uncertainty included in respect to the process Analytical Limit is made in section 3. 7.1. 2.
- b. As Left Tolerance The determination of the As Left Tolerance is made by applying the AL_Rack_r nter_Range uncertainty:
+ ALT= 25% X 10 [(1.52%/100%) X 10.3]
+ ALT= 35.9% RTP
-AL T = 25% X 10 [(-1.52%/100%) X 10.3]
-ALT = 17.4% RTP The positive ALT in percent linear scale:
+ ALT= [ log1o35.9- (-8) ] I 10.3
+ ALT= 92.8%
- 3. 7 . 1. 2 Trip Setpoint- Analytical Limit There is not anAnalytical Limit; therefore, an Analytical Limit is set at 75% RTP and verified that the setpoint falls within an established range. Below is analyzed the trip setpoint with the new rack uncertainties to the selectedAnalytical Limit.
TheAnalytical Limit in percent linear scale:
AL_ rnter_Range = [ log1o 75 - (-8) ] I 10. 3 Page 10 of 18 3/11/2015
AL_Inter_Range = 95.9%
And the Margin from the Trip Setpoint to the Analytical Limit is:
Margin=AL- [ Trip Setpoint + ( TL U) ]
Where the setpoint and TLU from sections 3. 3 and 3.4 are:
Setpoint_ 5 %_%_Span= 91.24%
2
- TLU_lnter_Range=- 3.69% span Margin =A - (Setpoint + TLU)
Margin = 95.9% - 91. 24% - 3. 69%
Margin =+ 0. 97%
There is a positive margin from the Analytical Limit in the linear scale from the 75% RTP value; therefore, the setpoint is found acceptable.
Conclusions:
Current TS allowable value should be replaced by the calculated AFT of S 38.5%
RTP.
3.7 . 1.3 P-6 Interlock The P-6 Permissive enables the Source Range Reactor trip setpoint to be blocked on increasing power and disables this block on decreasing reactor power. Permissive setpoint is 1o-10 amperes (approximately 1o-s % RTP) and the reset is 7x1o-11 amperes (approximately 7 . 3x1o-6% RTP) (Reference 4.5). The reset has its allowable value in the TS of <6 x 1o-11 amperes; therefore, the setpoint of interest is the reset with current allowable value in the Technical Specifications. The specified values are nominal settings which do not have Analytical or Process Limits. Per observation of the systems scales correlation, Figure *1, the setpoint value allows extends 2 decades into the Source Range system scale. The setpoint is a permissive setpoint and is not used in any safety analyses; therefore, there is not an analysis made against any Analytical Limit.
- a. As Found Tolerance ITS Allowable Value The determination of the As Found Tolerance, or Allowable Value, for the P-6 Reset setpoint by applying the AFT_Rack_lnter_Range uncertainty is:
+AFT = 7.3 x 1 o-6 % x 1o [(1.82% /100%) x 10.3]
+ AFT = 1.1 x 1Oc5 % RTP
- AFT = 7.3 x 1o-6 % x 1o [(-1.82% /100%)x 10.3]
- AFT= 4. 7 x 10-6% RTP Page 11 of18 3/11/2015
The positive AFT in percent linear scale I volts:
+AFT= [ log 1o 1. 1 x 10 5 (-8)] I 10.3
+AFT= 29.5% or 3.041 volts (note)
The P-6 reset +AFT in percent linear scale I volts:
+AFT= [ log1o 7.3 x 10 (-8)] I 10. 3
+AFT= 27.8% or 2.863 volts (note)
Note: Per figure 1, the 0%-100% linear scale is equivalent to 0-10.3 volts.
- b. As Left Tolerance The determination of the As Left Tolerance is made by applying the AL_Rack_source uncertainty:
+ALT= 7.3 10-6% 10
[(1.52%/100%)x10.3]
X X
- ALT = 7.3 X 1o-6% X 10 [(-1.52%/100%)x10.3]
- AL T = 5.09 x 1o-6% RTP The positive ALT in percent linear scale:
+ALT= = [ log10 1.05 x 10 (-8)] I10.3
+ALT = = 29. 3%
- c. Analytical Limit- Setpoints Separation There is not a defined Analytical Limit (AL) for the P- 6 Permissive reset setpoint.
However, an adequate separation between the Source Range Setpoint decreasing value and the P-6 Reset increasing value exists as presented in Figure 1.
- d. Conclusions 11 The current allowable value of< 6 x10- amps should be updated to the new rack 6
performance by the calculated AFT of< 4.7 x 10- RTP.
Page 12 of 18 3/11/2015
Figure 1 NUCLEARINSTRUMENTRANGES OFTHEEXCO'RENEUTRONFLUXMONITORINGSYSTEMS Detector Power Source Voltage Isolated Intermediate Voltage Isolated Neutron Level Range Level Current Range Level Current Flux (nv) % Power (CPS) (VDC) (mA) (% Power) (VOC) (mA) 20 10 19.52 10 9 18.4 10 16.8 7 5 Decades Overlap 10 13.6 10 .Q
,j 10 CPS 6 1.2 10 --- - *
. 4 10A 10 104 8.8 10 .s 6.96x104 2 (8.346 vo!ts) 72 10
-6 10
- 1 5.6 10 4
10 10 ...
10°
- 10 10 0.1 CPS Page 13 ofl8 3/11/2015
Figure 2 Current Systems Scaling Relationship A :.. 11:r M .ii..
p I!
R
% E p :s 0
w 5
R u'
1 c -
1U' Page 14 of 18 3/11/2015
i 4.0
References:
4.1 Purchasing Specification S-C-DE-NIS-0210, Excore Nuclear Instrumentation System (NIS) 4.2 Calculation SC-NIS001-01, Rev. 3, Salem Unit 1/2 Reactor Flux Source Range Instrumentation 4.3 Thermo Fisher instrument accuracies, Attachment 1 4.5 SC-NIS002-01, Rev. 1, Salem Unit 1/2 Reactor Flux Intermediate Range Instrumentation, NI35B, N136B
- 4. 6 TSTF-493, Rev. 4, Technical Specifications Task Force 4.7 Nuclear Instrument Ranges of the ex-core Neutron Flux Monitoring Systems, Thermo Fisher- Figure 1 4.8 Current Systems Scaling Relationship, PSEG's Excore Nuclear Instrumentation System, ILT-11-01, Paul Abbott July 11th and 12th 2011. , Figure 2 Page15of18 3/11/2015
Attachment 1
'MwdQi k!:;;ardnptl*nw' January 15., 2015 PSEG N:t.rolaar LlC Sailem Slte Attn: Saaed Savar Alloway Cre![l'k. N:eck Road H.anccks 'BJidge, NJ 08038 USA
Subject:
Sl!iltpoint Melhoool:ogy Tabie 1 Source Ransa Values Tabla 2"' Intermediate Range Values
- SaeMxl, Atlac:hed P'tsaei find two tab'l:es that provide information that can be to upgfiilde your wetl:l'lghouse SGtpolnt Methodology Crucutatio:I1S tO reflect the morn mod;am !il[ectronlos that will be incll.lded in your new Tham1o Flsh:er Scl:entif!c Source and lnt{i!rmectnate Rang Neutron Flux Monitoring Systems, Eliiictrlcal Englnermr
($6l88:?12:i phone L4ll.vLpa!'G:slt them1Q,cc:tm
$>m0ft.ta-.CA !A>!!; WHN!ll
\tiOI:;i! Hlrnf14*)!l' h;t Page 16 of 18 3/11/2015
{PMA)
PR!M\ARY ELEMENT ACCURACY
{I ocluded i!n RCA)
SENSOR CALIBRATION (SCA} 0%0FSPAN
{lnciud lin RCA)
SENSOR MEASU1R.EMENT & T'EST O% OF SiP.AN
'SENSOR TEMPERATURE iEF!FECTS 0%0fS;?J\N
.S!EN:SQR PRESSURE EFFECTS 0%0fSJPAN
..... ..... O"Jclygt?fijp, .8J;f\1 .....
SENSOR: DRIFT {SO) 0%0FSPAN ENVJRONMENT:AL ALI.OWANCE:S RACK C.At !BRAI ION Ri\C.K ACCURACY *I:Rcft,) +.!* 2J1% 0 F SPAN M EIU?IJ81;1i:: NT & TEST .......... . . .....fRMTt ..... l+l" CkOfi%, OF SPAN (RCSA)
AACK'TEMPEAATUR !FFECTS (RTE)
{RCA nclud;ss temper1:ltuwe effects:)
F'ACK ORI\FT (24 MONTH$) (RD}
"' In pw*ent of :span (0. '1 to 1 OE6 CPS)
- Sam termined val'e Page 17 of 18 3/11/2015
T.ABLE 2 PARAMETE:R
- ..:::*J PROCESS MEASUREMENT ACCURAOY PRlM'ARY Ei:.iEMENT ACCURACY (PEA)
{Included fn RCA) lSEN:SOR CAll BRATION (SGA)t 0%0FSPAN (lncludiFldhn RCA)
SESOR MEASUREMENT & rESl (SMTE) r=.:to OF SPAN
.. {lnc!lJd>ed In RCA)
SENSOR DRI.FT 0%0F SPAN ENVI:IRONMENT.Al AlLO\lVANCES 0%0F SPAN 1RACK CAUSPJ\llON RACK .A.CCU RACY {RCA}
(RCA InCludes ternrature effects)
MEASUFtEMENl & TEST
, COMPARATOR (RCSA)
AACK TEMPEAATU RE EFFECTS {RTE} 0%0F SPAN 03G&incluc.iE?Jrn:P9lrf3iWffts}.
1RACK DRIFT'(24 MONTHS) {RD)
"" AU p:a;ran1eters :tre :z;hO'Wf! In percent of' >prl (l\ogarithmc $ca:le 1 e;.eil<,.)) to 2Qto)
- SJem d\t'ierrnrnn:id valllJ.
Page 18 of18 3/11/2015