Text

Kansas State Univ., Manhattan - 2018 Annual Operating Report
	ML19337B054
Person / Time
Site:	Kansas State University
Issue date:	11/28/2019
From:	Cebula A; Kansas State University, Manhattan
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML19337B054 (47)
	v • d • e

Alan Cebula, Ph.D.

Nuclear Reactor Facility Manager 3002 Rathbone Hall Kansas State University Manhattan, KS 66506 USNRC Attn: Document Control Desk Washingto~ DC 20555-0001 28 November 2019

Subject:

2018 Annual Operating Report for the Kansas State University TRIGA Mark II Nuclear Reactor (Facility Lic~nse # R-88, Facility Docket# 50-188)

To Whom It May Concern:

This document serves as the annual operating report for the Kansas State University, (KSU) nuclear reactor. The report is divided into paragraphs addressing specific items listed as requirements in the Technical Specifications 6.11 e.

Sincerely, Alan Cebula, Ph.D.

Nuclear Reactor Facility Manager Kansas State Uni:versity Attachments:

1. Kansas State University TRI GA Mark II Reactor Annual Report, CY 2018.

2. 10CFR50.59 Screening.FoilIL5 Cc: Linh Tran, Project Manager, NRC Craig Bassett, Inspector, NRC

AITACHMENT 1 KANSAS STATE UNIVERSITY TRIGA MARK II REACTOR ANNUAL REPORT Kansas State University TRIGA Mark II Reactor Annual Report, CY 2018 Introduction The Kansas State University Nuclear Reactor Technical Specifications (TS) require a routine written report to be transmitted to the US Nuclear Regulatory Commission within 60 days after completion of the first calendar year of operating, and at intervals not to exceed twelve months thereafter, providing the following information:

TS.6.11.e. l -

A brief narrative summary of operating experience (including experiments performed), changes in facility design, performance characteristics, and operating procedures related to reactor safety occurring during the reporting period; and results of surveillance tests and inspections.

TS.6.11.e.2 -

A tabulation showing the energy generated by the reactor (in megawatt-hours).

TS.6.11.e.3 -

The number of emergency shutdowns and inadvertent scrams, including the reason thereof and corrective action, if any, taken.

TS.6.11.e.4 -

Discussion of the major maintenance operations performed during the period, including the effects, if any, on the safe operation of the reactor, and the reasons for any corrective maintenance required.

TS.6.11.e.5 -

A summary of each change to the facility or procedures, tests, and experiments carried out under the conditions of 10.CFR.50.59.

TS.6.11.e.6 -

A summary of the nature and amount of radioactive eflluents released or discharged to the environs beyond the effective control of the licensee as measured at or before the point of such release or discharge.

TS.6.11.e. 7 -

A description of any environmental surveys performed outside the facility.

TS.6.11.e.8 -

A summary of radiation exposures received by facility personnel and visitors, including the dates and time of significant exposure, and a brief summary of the results of radiation and contamination surveys performed within the facility.

This information is transmitted in this report, in sections separated by TS clause. This report covers January 2018 -December 2018.

Page 1 of9

ATTACHMENT 1 KANSAS STATE UNIVERSITY 1RIGA MARK II REACTOR ANNUAL REPORT TS.6.11.e.1 - A brief narrative summary of operating experience (including experiments performed), changes in facility design, performance characteristics~ and operating procedures related to reactor safety occurring during the reporting period; and results of surveillance tests and inspections.

The KSU reactor operated for its usual purposes in CY2018. Two reactor operation laboratory classes and a reactor theory laboratory class were directly supporte4 along with approximately 9 other courses with occasional use of the reactor. Through various outreach activities, classes, and research experiments, the facility hosted 1272 visitors.

Compared to CY2017, the number of visitors to the facility decreased by 22 percent Operations were significantly reduced for about a quarter of the year to address maintenance issues.

A majority of research experiments involved neutron activation analysis (NAA) and neutron detector testing utilizing in-core and beamline facilities. Other research activities included neutron radiography and gamma irradiation. Five pulses were performed during the first half ofCY2018. A new experiment to measure the distribution of fission products in a fuel element by gamma spectroscopy was approved.

A maintenance outage occurred from early July to the end of September. Thorough inspection and repair of the rod control system was conducted during the outage. Long-term issues involving the control rod drives and console were alleviated following the maintenance period. Revised operating procedures were incorporated following the outage to include periodic drive function testing. In addition to typical component maintenance, other facility changes during CY2018 included placing radiation monitoring systems on uninterruptable power supply systems.

The NRC routine annual inspection was completed from August 14-16, 2018. No violations or inspector follow-up items were reported. (See Inspection Report No. 50-188/2018-201 ).

Water ingress into the beam port facilities is still being monitored. Ingress is observed to be minor and intermittent throughout the year. A repair plan for the. water ingress is still under evaluation.

Page 2 of9

ATTACHMENT 1 KANSAS STATE UNIVERSITY TRIGA MARK II REACTOR ANNUAL REPORT TS.6.11.e.2 - A tabulation showing the energy generated by the reactor (in megawatt-hours).

The monthly total energy generated by the KSU reactor is recorded in Table 1. The same data is shown as a bar chart in Figure 1. The total MWh of operation decreased from the prior year, from 36.2 MWh to 25.0 MWh.

Table 1 - Energy generated by the KSU Triga Mark II reactor by month for CY 2018.

Month MWh Burnup January 1.41 February 2.09 March 7.13 April 5.44 May 1.40 June 1.89 July 0.20 August 0.00 September 0.00 October 1.26 November 4.02 December 0.18 TOTAL 25.03 MWh Burnup December (]

November October r I

September

August

July LI June May 1

April I:

j March February I January !:

I l

8 MONTHLYMWH Figure 1 - Energy generated by the KSU Triga Mark II reactor by month for CY 2018.

The reactor operated for a total of 313 hours0.00362 days <br />0.0869 hours <br />5.175265e-4 weeks <br />1.190965e-4 months <br /> during 2018, at an average power of 80 kW.

Table 2 lists the nwnber of hours operated and Figure 2 shows the percentage of Page 3 of9

ATIACHMENT 1 KANSAS STA TE UNIVERSITY TRIGA MARK II REACTOR ANNUAL REPORT operation for various purposes, i.e., research support, training, education, etc. Training percentage seems low because operator training was often performed when the reactor was being operated for another purpose, such as research support, classes, or maintenance. The plot demonstrates that the reactor is operated in accordance with our stated primary functions: education, research support, operator training, and demonstration (e.g., tours). Compared to CY2017, research and tour operations were reduced while class and maintenance remained relatively constant. The extended maintenance outage contributed to the reduction in operating hours for research and tours by at least one-half. Since the maintenance period occurred mostly outside of the academic schedule, class operations were un-affected.

Table 2 - Operating hours grouped by purpose at the KSU TRIGA Mark II reactor for CY 2018.

Operating Time (hr]

Purpose Research 71 Tours 25 Classes 111 Maintenance 86 Training 10 Testing 10 TOTAL 313 Reactor Operations Hours by Purlpose

~ Research

Tours Classes

~ Maintenance

= Training

Testing Figure 2 - KSU operations distribution, CY2018, based on purpose of operation.

Page 4 of9

ATIACHMENT 1 KANSAS STATE UNIVERSITY TRIGA MARK II REACTOR ANNUAL REPORT TS.6.11.e.3 - The number of emergency shutdowns and inadvertent scrams, including the reason thereof and corrective action, if any, taken.

For CY 2018, there was a total of20 inadvertent SCRAMS. Table 3 summarizes the inadvertent SCRAMS for CY 2018 at the KSU reactor. No emergency shutdowns occurred during the time period reported. Table 3 does not include single dropped rods.

Occasionally, a single rod would drop, but not due to a reactor trip. Single rod drops were corrected following refurbishing the shim rod drive.

Note that the period scram due to electrical noise occurred when the other rods were bottomed. In other words, the reactor is at very low power when the scram occurred.

During the maintenance outage from 7 /11/18 to 9/28/18, the NL W-1000 channel was investigated for causes of electrical noise. The P A-1000 preamp for the NL W-1000 was checked and a ground connection was secured. No other period scrams due to electrical noise were observed following the maintenance outage.

Table 3 - Inadvertent SCRAMS and Emergency Shutdowns.

Date ActiQn (Qmments 2/14/18 Period Scram Caused by electrical noise during checkout 2/15/18 Period Scram Electrical noise during checkout 3/13/18 Period Scram Electrical noise while driving in cylinder following manual shutdown 3/16/18 Period Scram Operator Error 3/30/18 Peliod Scram Electrical noise from rod motion in source range 3/30/18 Period Scram Electrical noise from rod motion in source range 3/30/18 Period Scram Electrical noise from rod motion in source range 3/31/18 Period Scram Electrical noise from rod motion in source range 4/20/18 SCRAM NPP-1000 Test button accidently depressed 4/24/18 Peliod Scram Electrical noise from rod motion in source range 4/25/18 Period Scram Operator Error 4/25/18 Period Scram Operator Error 4/26/18 Period Scram Operator Error 5/1/18 SCRAM Mode selector switch moved passed Steady State from Auto 5/2/18 SCRAM Console power loss 5/4/18 Period Scram Electrical noise from rod motion in source range 6/6/18 SCRAM Occurred while moving from Auto to Steady State 6nt18 SCRAM All SCRAMs indicated, cause unknow to operator.

Likely power loss 7/3/18 Period Scram Electrical noise from rod motion in source range Manual and Fuel Fuel temps indicated 19 and 18 C during SCRAM.

10/26/18 Temp Spurious SCRAM. Possible r~lay issue suspected.

Relay K4 reseated.

Page 5 of9

ATTACHMENT 1 KANSAS STATE UNIVERSITY TRIGA MARK II REACTOR ANNUAL REPORT TS.6.11.e.4 - Discussion of the major maintenance operations perfonned during the period, including the effects, if any, on the safe operation of the reactor, and the reasons for any corrective maintenance required.

Various system maintenance was performed throughout CY2018 for part failure due to normal wear and tear. An extended outage from 7 /11/18 to 9/28/18 was taken to troubleshoot and repair issues with control rod drives. The following is a summary of all major maintenance activities during CY2018:

AC power line conditioner installed.

Shim rod ON indicator lamp replaced Safety rod position potentiometer replaced.

NPP-1000 current limiting resistor replaced.

Control room area radiation monitor replaced due to detector failure.

Secondary cooling wye strainer elbow replaced due to pinhole leak.

North cooling tower fan belt slipped off power transmission system. Fan belt replaced.

Heating Ventilation and Air Conditioning supply line valve replaced.

Fuel tool actuating line repaired.

Area Radiation Monitoring system uninterruptable power supply (UPS) battery failed. UPS replaced.

Continuous Air Monitor power routed through UPS.

Rod drive control system troubleshoot o

Current limiting resistor for rod control indicators resoldered.

o Contact resistance in console microswitches resulted in rod drive drift. All control rod drive console microswitches replaced.

o Regulating rod drive drift from bias resistance too high. Regulating rod drive fixed bias resistor bypassed. Bias resistance provided by potentiometer.

o All control rod drives cleaned, refurbished, and adjusted.

o Safety rod drive motion issues. Slide bearings replaced.

o Regulating rod drive pausing while withdrawing. Regulating rod drive gearbox replaced.

o Transient rod drive limit switch LS3 replaced.

NL W-1000 preamp P A-1000 grounding secured.

Exhaust plenum power routed through UPS.

Page 6 of9

AITACHMENT 1 KANSAS STATE UNNERSITY 1RIGA MARK II REACTOR ANNUAL REPORT TS.6.11.e.5 - A summary of each change to the facility or procedures, tests, and experiments carried out under the conditions of 10.CFR.50.59.

Tue following changes were carried out under 10CFR50.59:

Added Experiment 55 - Fuel Element Gamma Spectroscopy Using Fuel Movement Device.

Area Radiation Monitoring system uninterruptable power supply replaced.

Continuous Air Monitor connected to uninterruptable power supply.

Rod drive control microswitches replaced.

Regulating rod drive R903 bias resistor bypassed.

Procedure 8, Reactivity Balance revised.

Procedure 12, Instrument Checkout revised.

Procedure 15, Steady State Operations revised.

Exhaust Plenum Monitor connected to uninterruptable power supply.

Tue screening forms for these changes are attached.

TS.6.11.e.6 - A summary of the nature and amount of radioactive effluents released or discharged to the environs beyond the effective control of the licensee as measured at or before the point of such release or discharge.

Per procedure, the concentration and total activity were calculated prior to discharge, showing both to be below the limits in 1 OCFR20. Table 4 summarizes the average concentration and total activity released.

Table 4 - Summary of radioactive effluent (water)

Avg.

Total Total Activity Isotope Concentration Volume (Ci/ ml)

(ml)

Released (Ci)

Alpha-6.14E-15 1.01+07 6.20E-08 emitters Beta-l.98E-11 l.01E+07 2.00E-04 emitters Tue only other discharge beyond the facility boundary was HV AC condensate discharge to the sanitary sewer. Since the Kansas State University average water usage is 750,000 gallons per day, it is nearly impossible to exceed 1 OCFR20 limits for effluent concentration at the KSU reactor. HV AC condensate water is never circulated through or near the reactor core and historically radiation levels in HV AC condensate are at or near background levels.

Page 7 of9

ATIACHMENT 1 KANSAS STATE UNIVERSITY TRI GA MARK II REACTOR ANNUAL REPORT TS.6.11.e. 7 - A description of any environmental surveys performed outside the facility.

Radiation surveys are performed within and around the facility to verify that radiation levels remain safe when at full-power operation. These surveys indicate that the dose rate (gamma and neutron) at the reactor dome does not exceed the hourly dose limit to members of the public of 2 mrem / h, as set forth in 1 OCFR20, which indicates that the outside dose cannot exceed this limit.

TS.6.11.e.8 - A summary of radiation exposures received by facility personnel and visitors, including the dates and time of significant exposure, and a brief summary of the results of radiation and contamination surveys performed within the facility.

Overall, no staff exceeded 100 mrem for CY2018. Table 5 shows the distribution of workers receiving given amounts of dose. The average deep dose equivalent was 17.4 7 mrem with a maximum of 54 mrem. The lens dose equivalent had a similar average of 18.53 mrem and the maximum for an individual of 59 mrem. Shallow dose equivalent average was 20.07 mrem with a maximum of 59 mrem. Extremity monitoring had an average of 32.07 mrem and a maximum of 82 mrem.

Table 5 - Summary of total occupational dose received by KSU reactor workers from 1/1/2018 -

12/31/2018.

DDE LOE SDE Max mrem Extremity (0, 10) 4 4

4 4

(10, 20) 4 4

2 0

(20, 30) 5 5

7 2

(30, 40) 1 1

1 2

(40, SO]

0 0

0 5

(50,100]

1 1

1 2

(100,150]

0 0

(150,200]

0 0

Page 8 of9

ATTACHMENT 1 KANSAS STATE UNIVERSITY TRIGA MARK II REACTOR ANNUAL REPORT Visitor dose at the KSU TRIGA reactor facility is measured using self-reading pocket ion chamber dosimeters, with an indication range from 0-200 mR. Self-indicated pocket dosimeter readings suffer from imprecision due to parallax error, sometimes resulting in negative values or readings above the true value.

2018 Visitor Dose Records 900 823 800 700 Ul "C...

8 600 Q,J ix:...

~ 500 Ul >

0 400 Q,I "E 300

i z 200 100 6

9 5

12

~ ~

0

<=0 (0,1]

(1,2]

(2,5]

(5,10]

(10,20]

{20,50]

>50 Exposure [mR]

Figure 3 - Visitor exposure records from CY 2018.

All monthly radiation surveys and contamination surveys conducted at the facility in 2018 were nominal. Overfill of the Bulk Shield Tank (BST) occurred during a routine refill. The BST is filled with de-ionized water and used for storing materials for decay.

Monthly surveys ofBSTwater show less than or equal to background levels. Water samples and wipe tests of areas affected by the overfill showed no areas above background.

This concludes the 2018 Annual Report for the Kansas State University TRIGA Mark II Nuclear Reactor.

Page 9 of9

Date: 5/29/18

Title:

Experiment 55 - Fuel Element Gamma Spectroscopy Using Fuel Movement Device Performer: Max Nager

==

Description:==

Experiment 55 has proposed a new method for fuel movement. Fuel will be removed from the tank using previously approved facility procedures, but subsequently transferred to a Fuel Movement Device (FMD) that holds it in place and rotates it while gamma spectroscopy is performed. The description of the FMD is outlined in both Experiment 55 and the experimenter proposal.

SCREENING - The following guidance provides criteria to screen the proposed change from further assessing the need for NRC review.

SSC Affected SSC Design Function Failure Mode(s) Accident Scenarfo(s)

Fuel Fuel Movement Fuel Movement Fuel Element Failure in Movement Device Failure Air Safetv Analysis and Accident Response/Mitigation YES NO Decrease SSC design function reliability when failure would initiate X

accident Decrease SSC design function reliability when failure would affect X

accident mitigation Reduce redundancy, reliability, or defense in depth X

Add or delete an automatic or manual design function of an SSC X

Human Interface YES NO Convert an automatic feature to manual or vice versa X

Adversely affect ability to perform required actions X

Adversely affect time response of required actions X

Interface Outside of 'the Proposed Chanae YES NO Degrade seismic or environmental qualification X

Affect method of evaluation used to establish design basis or safety X

analvsis Introduce unwanted or previously unreviewed system or material X

interaction

{Not described in SAR) indirect effects on electrical distribution X

{Not described in SAR) indirect effects on structural integrity X,

{Not described in SAR) indirect effects on environmental conditions X

(Not described in SAR) indirect effects on other SAR design functions X

EVALUATION - If the change does affects (1) a design function of SSC, (2) a method of performing or controlling design function, or (3) evaluation for demonstrating the design function will be accomplished, as indicated by one or more YES answers in the "Screening" section, complete the applicable tables below.

Does the change result in more than a minimal increase in the frequency YES NO of occurrence of an accident previously evaluated in the final SAR (as X

updated)?

Accident Potential Impact on Accident Freouencv Reactivity Addition NIA LOCA NIA Fuel Handling NIA Does the change result in more than a minimal increase in the likelihood YES NO of occurrence of a malfunction of an SSC important to safety previously X

evaluated in the final SAR (as updated)?

Affected SSC Potential Impact on Likelihood of Malfunction Continuous Air NIA Monitor Does the change result in more than a minimal increase in the YES NO consequences of an accident previously evaluated in the final SAR (as X

updated)?

Accident Potential Impact on Accident Conseouences Reactivity Addition NIA LOCA NIA Fuel Handling NIA Does the change result in more than a minimal increase in the YES NO consequences of a malfunction of an SSC important to safety previously X

evaluated in the final SAR (as updated)?

Affected SSC Potential Impact on ConseQuences of Malfunction NIA NIA

EVALUATION - continued Does the change create a possibility for an accident of a different type YES NO than previously evaluated in the final SAR as u dated ?

x Accident Description (Including Ukel/hood and Consequences)

NIA Does the change create a possibility for a malfunction of an SSC YES NO important to safety with a different result than any previously evaluated X

in the final SAR (as updated)?

Accident Affected SSC Result Reactivity NIA NIA Addition LOCA NIA NIA Fuel Handling NIA NIA Other NIA NIA Does the change result in exceedance or alteration of a design basis YES NO limit for a fission roduct barrier as described in the SAR as updated ?

x Cateaory Reference/Text Value Design Basis Limit NIA NIA Analysis NIA NIA Approach to Limit NIA NIA Does the change result in departure from a method of evaluation YES NO described in the final SAR (as updated) used to establish design bases X

or in the safetv analysis?

Cateaory Reference/Text Value Design Basis NIA NIA New Analysis NIA NIA Comparison NIA NIA

Comments: The FMD utilizes an electronically actuated apparatus to grip and move the fuel element. The Fuel Element Failure in Air MHA does not credit any type of fuel movement device, including the fuel tool used for inspections. The main mode of failure is accidental release of the element by the FMD. Even if the roughly two and a half foot drop resulted in element failure, the consequences would be within the scope of the highly conservative MHA. The other, hypothesized mode of failure is the FMD accidentally exerting its full weight on the element. In that scenario analyses have shown a buckling factor of safety greater than 50. Any additional fuel movement increases the likelihood of occurrence of the MHA. However, this device does not pose any more likelihood of increasing the occurrence of the MHA than the fuel handling tool.

APPROVAL -

According to Technical Specifications, Section 6.2(b)4, the Reactor Safeguards Committee is responsible for detennining "whether changes in the facility as described in the safety analysis report (as updated), changes in the procedures as described in the final safety analysis report (as updated), and the conduct of tests or experiments not described in the safety analysis report (as updated) may be accomplished in accordance with 10 CFR 50.59 without obtaining prior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90."

Date of RSC approval:

~ :{

tLAJL, )._o t <g Method of RSC approval: £ ~ ~1 v err -e.

Attach appropriate records of RSC approval (e.g., email ballots or meeting minutes) to this fonn.

Date: 7 /9/18

Title:

ARM UPC Replacement Performer: Max Nager

==

Description:==

The Area Radiation Monitoring system (ARM) is powered by an APC SMART-UPS uninterruptible power supply. It is desired to replace the APC SMART-UPS with the APC Back-UPS Pro. The proposed alternative matches the current system in function.

SCREENING-The following guidance provides criteria to screen the proposed change from further assessing the need for NRC review.

SSC Affected SSC Design Function Failure Mode(s) Accident Scenario(&)

Area Radiation Radiation Monitoring Power loss LOCA Monitoring system Sa'fety Analysis and Accident Response/Mitigation YES NO Decrease SSC design function reliability when failure would initiaE X

accident Decrease SSC design function reliability when failure would affect X

accident mitigation Reduce redundancy, reliability, or defense in depth X

Add or delete an automatic or manual desian function of an SSC X

Human Interface YES NO Convert an automatic feature to manual or vice versa X

Adverselv affectabilitv to oerform required actions X

Adversely affect time response of required actions X

Interface Outside of the Proposed Chanae YES NO Dearade seismic or environmental aualification X

Affect method of evaluation used to establish design basis or safety X

analysis Introduce unwanted or previously unreviewed system or material X

interaction (Not described in SAR) indirect effects on electrical distribution X

(Not described in SAR) indirect effects on structural lnteoritv X

(Not described In SAR) Indirect effects on environmental conditions X

(Not described in SAR) Indirect effects on other SAR desian functions X

EVALUATION - If the change does affects (1) a design function of SSC, (2) a method of performing or controlling design function, or (3) evaluation for demonstrating the design function will be accomplished, as indicated by one or more YES answers in the "Screening" section, complete the applicable tables below.

Does the change result in more than a minimal increase In the frequency YES NO of occurrence of an accident previously evaluated in the final SAR (as X

updated)?

Accident Po'lential lmoact on Accident Freauency Reactivity Addition NIA LOCA NIA Fuel Handling NIA Does the change result in rnorethan a minimal increase in the likelihood YES NO of occurrence of a malfunction of an SSC important to safety previously X

evaluated In the final SAR (as uodated)?

Affected SSC Potential lmoact on Likelihood of Ma/function Radiation NIA Monitoring Does the change result in more than a minimal increase in the YES NO consequences of an accident previously evaluated in the final SAR (as X

updated)?

Accident Potential Impact on Accident Consequences Reactivity Addition NIA LOCA NIA Fuel Handling NIA Does the change result in more than a minimal increase in the YES NO consequences of a malfunction of an SSC important to safety previously X

evaluated in the final SAR (as updated)?

Affected SSC Potential lmoact on Consequences of Ma/function NIA NIA

EVALUATION - continued Does the change create a possibility for an accident of a different type YES NO than reviousl evaluated in the final SAR as u dated ?

x uences NIA Does the change create a possibility for a malfunction of an SSC YES NO important to safety with a different result than any previouslyevalua1Bd X

in the final SAR (as updated)?

Accident Affected SSC Result Reactivity NIA NIA Addition LOCA NIA NIA Fuel Handling NIA NIA Other NIA NIA Does the change result in exceedance or alteration of a design basis YES NO limit for a fission roduct barrier as described in the SAR as u dated ?

x Cateaory Reference/Text Value Design Basis Limit NIA NIA Analysis NIA NIA Approach to Limit NIA NIA Does the change result in departure from a method of evaluation YES NO described in the final SAR (as updated) used to establish design bases X

or in the safetv analvsis?

Ca'leaorv Reference/Text Value Design Basis NIA NIA New Analysis NIA NIA Comparison NIA NIA

Comments: The SMART UPS has an output of 950 watts/1400 VA while the Back Ups has an output of 900 watts/1500 VA. The SMART-Ups is not described in the SAR so a replacement with another UPS of nearly the same output does not deviate from the safety basis.

APPROVAL -

According to Technical Specifications, Section 62(b)4, the Reactor Safeguards Committee is responsible for determining "whetherchanges in the facility as described in the safety analysis report (as updated), changes in the procedures as described in the final safety analysis report (as updated), and the conduct of tests or experiments not described in the safety analysis report (as updated) may be accomplished in accordance with 10 CFR 50.59withoutobtaining priorNRC approval via license amendmentpursuantto 10 CFR Sec. 50.90."

Date of RSC approval: ;/,o/R Method of RSC approval: e.~~I 8-llo Attach appropriate records of RSC approval (e.g., email ballots or meeting minutes) to this form.

Date: 8/23/18

Title:

Connection of CAM to Uninterruptible Power Supply Performer: Max Nager

==

Description:==

The Continuous Air Monitor (CAM) consists of a Thermo Fisher AMS4 Iodine unit. The CAM is currently connected directly to mains power, but a change is desired in the form of powering it from an Uninterruptible Power Supply (U~S). The UPS will be powered from mains. The CAM is required to be OPERATING per TS 3.3.4(e) and provides indication of a potential instance of fuel element failure in air.

SCREENING - The following guidance provides criteria to screen the proposed change from further assessing the need for NRC review.

SSC Affected SSC Design Function Failure Mode(s)

Accident Scenario(s)

Continuous Air Radiation Monitoring Detector Failure Fuel Element Failure Monitor in Air Safetv Analysis and Accident Response/Mitigation YES NO Decrease SSC design function reliability when failure would initiate X

accident Decrease SSC design function reliability when failure would affect X

accident mitigation Reduce redundancv, reliability, or defense in depth X

Add or delete an automatic or manual desiqn function of an SSC X

Human Interface YES NO Convert an automatic feature to manual or vice versa X

Adversely affect ability to perform required actions X

Adversely affect time response of required actions X

Interface Outside of the Proposed Chanae YES NO Deqrade seismic or environmental qualification X

Affect method of evaluation used to establish design basis or safety X

analysis Introduce unwanted or previously unreviewed system or material X

interaction (Not described in SAR) indirect effects on electrical distribution X

(Not described in SAR) indirect effects on structural integrity X

(Not described in SAR) indirect effects on environmental conditions X

(Not described in SAR) indirect effects on other SAR desiqn functions X

EVALUATION - If the change does affects (1) a design function of SSC, (2) a method of performing or controlling design function, or (3) evaluation for demonstrating the design function will be accomplished, as indicated by one or more YES answers in the "Screening" section, complete the applicable tables below.

Does the change result in more than a minimal increase in the frequency YES NO of occurrence of an accident previously evaluated in the final SAR (as X

updated)?

Accident Potential Impact on Accident Frequency Reactivity Addition LOCA Fuel Handling Does the change result in more than a minimal increase in the likelihood YES NO of occurrence of a malfunction of an SSC important to safety previously X

evaluated in the final SAR (as updated)?

Affected SSC Potential Impact on Likelihood of Malfunction Continuous Air There would be a decrease in the likelihood of malfunction.

Monitor Does the change result in more than a minimal increase in the YES NO consequences of an accident previously evaluated in the final SAR (as X

updated)?

Accident Potential Impact on Accident Consequences Reactivity Addition LOCA Fuel Handling Does the change result in more than a minimal increase in the YES NO consequences of a malfunction of an SSC important to safety previously X

evaluated In the final SAR (as updated)?

Affected SSC Potential Impact on Consequences of Ma/function

EVALUATION - continued Does the change create a possibility for an accident of a different type YES NO than previous! evaluated in the final SAR as updated ?

x Accident Description (Including Likelihood and Consequences)

Does the change create a possibility for a malfunction of an SSC YES NO important to safety with a different result than any previously evaluated X

in the final SAR (as updated)?

Accident Affected SSC Result Reactivity Addition LOCA Fuel Handling Other Does the change result in exceedance or alteration of a design basis YES NO limit for a fission product barrier as described in the SAR as updated)?

x Cateaory Reference/Text Value Design Basis Limit Analysis Approach to Limit Does the change result in departure from a method of evaluation YES NO described in the final SAR (as updated) used to establish design bases X

or in the safety analysis?

Cateaory Reference/Text Value Design Basis New Analysis Comparison

Comments: The CAM provides indication of fuel element failure, but is not credited in the fuel element failure MHA. Repeated power outages at the facility have resulted in malfunction of the CAM, rendering it INOPERABLE for as long as several weeks at a time. Powering it from a UPS will reduce the likelihQod of a sudden power loss to the CAM, along with an associated malfunction. Furthermore, it will improve the safety function since the CAM will be capable of monitoring for short periods following a facility power outage.

APPROVAL -

According to Technical Specifications, Section 6.2(b)4, the Reactor Safeguards Committee is responsible for determining "whether changes In the facility as described in the safety analysis report (as updated), changes in the procedures as described in the final safety analysis report (as updated), and the conduct of tests or experiments not described in the safety analysis report (as updated) may be accomplished in accordance with 10 CFR 50.59 without obtaining prior NRC approval via license amendment pursuant to 10 CFR Sec. 50.90."

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Date of RSC approval: ~/1q{fl.

~

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Method of RSC approval: £,._.\\I Ba.l\\o+-

Attach appropriate records of RSC approval (e.g., email ballots or meeting minutes) to this form.

Date: 9/25/18

Title:

Replacement of Console 1 SM299-T2 Micro Switch to Honeywell 11 SM 1-T2 Micro Switch Performer: Max Nager

==

Description:==

The reactor console has a set of 15 Twist-Lite Series 12*, Push-Button lamp indicators, referred to as switches S3-S17. S3-S7 correspond to the rod drive magnet/contact buttons, S8-S12 to the "UP" buttons, and S13-S17 to the "ON" buttons.

Each switch contains a sub-assembly of three 1 SM299-T2 Micro switches. Current 1 SM299-T2 Micro switches have developed contact resistance and therefore require replacement.

SCREENING - The following guidance provides criteria to screen the proposed change from further assessing the need for NRC review.

SSC Affected SSC Deslan Function Failure Mode(s)

Accident Scenario(s)

Manual Scram Scram Switch Failure None Safety Analysis and Accident Response/Mitigation YES NO Decrease SSC design function reliability when failure would initiate X

accident Decrease SSC design function reliability when failure would affect X

accident mitioation Reduce redundancv reliabilitv, or defense in depth X

Add or delete an automatic or manual design function of an SSC X

Human Interface YES NO Convert an automatic feature to manual or vice versa X

Adversely affect ability to perform required actions X

Adversely affect time response of required actions X

Interface Outside of the Proposed Change YES NO Degrade seismic or environmental qualification X