NMP1L3064, Supplemental Response to Request for Additional Information - Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements....

Supplemental Response to Request for Additional Information - Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements....

ML15356A110
Person / Time
Site:	Nine Mile Point
Issue date:	12/22/2015
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF6061, NMP1L3064
Download: ML15356A110 (9)

Text

Exelon Generation 200 Exelon Way Kennett Square. PA 19348 www.exeloncorp.com 10 CFR 50.90 NMP1L3064 December 22, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Nine Mile Point Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-63 NRC Docket No. 50-220

Subject:

Supplemental Response to Request for Additional Information - "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)"

References:

1. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, "Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3},"

dated May 12, 2015.

2. Letter from Alexander N. Chereskin (Project Manager, U.S. Nuclear Regulatory Commission) to Bryan C. Hanson (Exelon) "Nine Mile Point Nuclear Station, Unit 1 - Request for Additional Information Regarding Adoption of Technical Specification Task Force Traveler 425 (CAC No.

MF6061)," dated November 9, 2015.

3. Letter from J. Barstow (Exelon Generation Company, LLC) to U.S.

Nuclear Regulatory Commission, "Response to Request for Additional Information -Application for Technical Specification Change Regarding Risk-Informed Justification for the Relocation of Specific Surveillance Frequency Requirements to a Licensee Controlled Program (Adoption of TSTF-425, Revision 3)," dated December 3, 2015.

By letter dated November 9, 2015 (Reference 2), the Nuclear Regulatory Commission (NRC) issued a Request for Additional Information (RAI) relative to Exelon's License Amendment Request (LAR) dated May 12, 2015 (Reference 1).

On December 3, 2015 (Reference 3), Exelon responded to the NRC RAI.

On December 15, 2015, a clarification call was held between NRC and Exelon personnel relative to Exelon's response as documented in Reference 3.

U.S. Nuclear Regulatory Commission Supplemental Response to Request for Additional Information Docket No. 50-220 December 22, 2015 Page 2 contains Exelon's revised response to Standard Technical Specifications Branch (STSB) RAl-4 and RAl-8. The NRC RAls as original documented in Reference 2 are re-stated followed by Exelon's revised response, which supersedes and replaces Exelon's original responses to STSB RAl-4 and RAl-8. to this letter contains revised Technical Specifications and Bases pages associated with Exelon's original response to STSB RAI Nos. -1, -3, and -6, and the current response to RAl-4.

Exelon has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference 1.

The additional information provided in this response does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. Furthermore, the additional information provided in this response 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.

There are no commitments contained in this response.

If you should have any questions regarding this submittal, please contact Enrique Villar at 610-765-5736.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 22rd day of December 2015.

~0;- 4-,Jrv= _j ~

James Barstow Director - Licensing & Regulatory Affairs

-u .

Exelon Generation Company, LLC Attachments: 1. Supplemental Response to Request for Additional Information

2. Revised Technical Specifications Pages cc: USNRC Regional Administrator, Region I w/attachments USNRC Senior Resident Inspector, NMP "

USNRC Project Manager, NRR, NMP "

A L. Peterson, NYSERDA II

License Amendment Request Attachment 1 Supplemental Response to Request for Additional Information Page 1 of 2 Docket No. 50-220 Technical Specifications Branch (STSB) RAl-4 In Attachment-3, "Proposed Technical Specification and Bases Page Changes," of the LAR submittal, the licensee requests to incorporate control of the SR 4.2. 7 .d frequency (page 109) into the SFCP. The current verbiage of this SR indicates that its performance is event driven. Specifically, it is performed "At least once per plant cold shutdown;" therefore, the event driving the SR performance is a cold shutdown . The licensee states that this LAR submission is in accordance with TSTF-425, Revision 3, which explicitly excludes purely event driven SRs from being eligible for incorporation into the SFCP.

Based on the above discussion, provide a justification for deviating from the TSTF-425 requirements or, if necessary, modify the submittal as appropriate.

Exelon Response to STSB RAl-4 Exelon will not be relocating Surveillance Requirement (SR) 4.2.7.d as part of this submittal. contains the mark-up TS page.

STSB RAl-8 In Attachment-3 of the LAR submittal, the licensee proposed to incorporate control of the following SR frequencies into the SFCP:

• Frequencies associated with individual SRs 4.1.3.e, 4.1.4.a, 4.2.5.b(1), 4.3.2.b, 4 .3.6.c(2), 4.6.3.a, 4.6.12.b, 4.6.13.b and

• SR frequencies listed in the following tables: 4.6.2b (parameters: 2, 6, 7, 8), 4.6.2g (parameters: 6, 7), 4.6.2i (parameters: a, b), 4.6.11 (parameters: 3, 4, 5, 7, 8), 4.6.13-1 (parameters: Reactor Water Temperature, Torus Water Temperature, Emergency Condenser Water Level, Drywell Temperature, and "All Rods In" Light).

These SR frequencies contain verbiage (e.g., refueling outage, major refueling outage, refueling cycle), which indicates that SR performance is based on a refueling outage event. In the licensee's submission it was unclear whether or not all of these SRs were frequency based or, possibly, purely event driven. For example, if performance of an SR is mandated every time the unit is transitioned to a plant shutdown condition during an operating cycle (i.e., not during a major refueling outage), then this SR would be considered purely event driven. The licensee states that this LAR submission is in accordance with TSTF-425, Revision 3, which explicitly excludes purely event driven SRs from being eligible for incorporation into the SFCP.

Based on the above discussion, address each of the aforementioned SR frequencies and indicate whether they are frequency based or purely event driven. If the SRs are purely event driven, justify their inclusion into the SFCP or if necessary, modify the submittal as appropriate.

License Amendment Request Attachment 1 Supplemental Response to Request for Additional Information Page 2 of 2 Docket No. 50-220 Exelon Response to STSB RAl-8 The *applicable portion of Technical Specification Task Force -425 (TSTF-425) for Nine Mile Point -1 (NMP-1) is based on the Boiling Water Reactor 4 Standard Technical Specifications (BWR4/STS), which defines activities to be performed during a refueling outage in terms of a specific numerical frequency interval (e.g., 18 months, 24 months depending on the length of the operating cycle). Because NMP-1 has custom Technical Specifications (TS), it does not use the specific numerical value to define these surveillance frequencies. The NMP-1 TS defines these surveillance frequencies utilizing the phrases refueling outage, major refueling outage, operating cycle, or refueling cycle.

NMP-1 TS defines the terms Major Refueling Outage and Operating Cycle as:

"For the purpose of designating frequency of testing and surveillance, a major refueling outage shall mean a regularly scheduled refueling outage (emphasis added); however, where such outages occur within 8 months of the end of the previous refueling outage, the test or surveillance need not be performed until the next regularly scheduled outage."

"An operating cycle is that portion of Station operation between reactor startups following each major refueling outage."

Nine Mile Point Unit 1 Final Safety Analysis Report (Updated) (UFSAR),Section IV, Reactor, states that, "The reactor is currently operating on a 24-month refueling cycle. Approximately 33 percent of the core is changed out each refueling." Furthermore, page X-30 of Section X .H.1 states, "Normal refueling conditions are based on refueling the reactor every 24 months."

From the TS definitions and the UFSAR statements, it is clear that the terms "operating cycle" and "refueling cycle" are equivalent terms. Similarly, the terms "major refueling outage" and "refueling outages" are equivalent to each other, with a frequency of 24-months.

Additionally, from these definitions and statements, it can be concluded that the performance of SRs containing the verbiage (e.g., refueling outage, major refueling outage, operating cycle, or refueling cycle), are based on a "regularly scheduled" interval and therefore are not event driven. Consequently, relocating the surveillance frequencies listed in NRC Question STSB RAl-8 to the Surveillance Frequency Control Program (SFCP) is in compliance with the TSTF-425 and the marked up pages of the Boiling Water Reactor 4 Standard Technical Specifications (BWR4/STS).

Exelon has previously utilized this interpretation successfully in obtaining approval of similar license amendments for Oyster Creek and Three Mile Island which, like NMP-1, have custom TS.

ATTACHMENT 2 SUPPLEMENTAL RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DOCKET NO. 50-220 REVISED TECHNICAL SPECIFICATIONS PAGES

NOTES FOR TABLES 3.6.2a and 4.6.2a (a) May be bypassed when necessary for containment inerting.

(b) May be bypassed in the refuel and shutdown positions of the reactor mode switch with a keylock switch.

(c) May be bypassed in the refuel and startup positions of the reactor mode switch when reactor pressure is less than 600 psi, or for the purpose of performing reactor coolant system pressure testing and/or control rod scram time testing with the reactor mode switch in the refuel position.

(d) No more than one of the four IRM inputs to each trip system shall be bypassed.

(e) No more than two C or D level LPRM inputs to an APRM shall be bypassed and only four LPRM inputs to an APRM shall be bypassed in order for the APRM to be considered operable. No more than one of the four APRM inputs to each trip system shall be bypassed provided that the APRM in the other instrument channel in the same core quadrant is not bypassed. A Traversing In-Core Probe (TIP) chamber may be used as a substitute APRM input if the TIP is positioned in close proximity to the failed LPRM it is replacing.

(f) Verify SRM/IRM channels overlap during startup after the mode switch has been placed in startup. Verify IRM/APRM channels overlap at least 1/2 decade during entry into startup from run (normal shutdown) if not performed within the previous 7 days.

(g) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before startup, if not performed within the previous 7 days. Not required to be performed during shutdown until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering startup from run.

(h) Each of the four isolation valves has two limit switches. Each limit switch provides input to one of two instrument channels in a single trip system.

(i) May be bypassed when reactor power level is below 45%.

0) Trip upon loss of oil pressure to the acceleration relay. in accordance with the Surveillance Frequency Control Program (k) May be bypassed when placing the reactor mode switch in the ~HUTDOWN position and all control rods are fully inserted.

(I)

(m) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during reactor operation when THERMAL POWER~ 25% of RATED THERMAL POWER. Adjust the APRM channel if the difference is greater than +2.0/-1.9% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 2.1.2a shall not be included in determining the difference.

(n) Neutron detectors are excluded.

AMENDMENT NO. ~ ~ +&J. 4.e3. 186 203

LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT

c. If Specifications 3.2. 7a and b above are not met, At least once per quarter the feedwater and initiate normal orderly shutdown within one hour INSERT 1 main-steam line power-operated isolation valves and have reactor in the cold shutdown condition shall be exercised by partial closure and within ten hours. subsequent reopening.

d. Whenever fuel is in the reactor vessel and the reactor d. At least once per plant cold shutdown the coolant temperature is less than or equal to 212°F, feedwater and main steam line power-operated the isolation valves on the shutdown cooling system isolation valves shall be fully closed and lines connected to the reactor coolant system shall be reopened, unless this test has been performed operable except as specified in Specification 3.2.7.e within the previous 92 days.

below.

e. In the event any shutdown cooling system isolation valve becomes inoperable whenever fuel is in the reactor vessel and the reactor coolant temperature is less than or equal to 212°F, the system shall be considered operable provided that, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, at least one valve in each line having an inoperable valve is in the mode corresponding to the isolated condition.

f. If Specifications 3.2.7.d and 3.2.7.e above are not met, either:

(1) Immediately initiate action to suspend operations with a potential for draining the reactor vessel (OPDRVs); or (2) Immediately initiate action to restore the valve(s) to operable status.

AMENDMENT NO. 442, 197 109

BASES FOR 3.2.3 AND 4.2.3 COOLANT CHEMISTRY In its May 8, 1997 letter, the NRC required that the licensee submit an application for amendment to address the differences between the current TS conductivity limits for reactor coolant chemistry and the analysis assumptions for the core shroud crack growth evaluations. The purpose of this specification is to limit intergranular stress corrosion cracking (IGSCC) crack growth rates through the control of reactor coolant chemistry. The LCO values ensure that transient conditions are acted on to restore reactor coolant chemistry values to normal in a reasonable time frame. Under transient conditions, potential crack growth rates could exceed analytical assumptions, however, the duration will be limited so that any effect on potential crack growth is minimized and the design basis assumptions are maintained. The plant is normally operated such that the average coolant chemistry for the operating cycle is maintained at the conservative values of< 0.19

µmho/cm for conductivity and < 5 ppb for chloride ions and < 5 ppb for sulfate ions. This will ensure that the crack growth rate is bounded by the core shroud analysis assumptions. Since these are average values, there are no specific LCO actions to be taken if these values are exceeded at a specific point in time. The EPRI "BWR Water Chemistry Guidelines-1996 Revision" (EPRI TR-103515-R1, BWRVIP-

29) action level 1 guidelines suggest that if conductivity is above 0.3 µSiem, or chloride or sulfate ions exceed 5 ppb, that corrective action be initiated as soon as possible and to restore levels below level 1 within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. If the parameters are not reduced to below these levels within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, complete a review and implement a program and schedule for implementing corrective measures.

Specifications 3.2.3a, b, and c are consistent with the licensee's commitment to Table 4.4 of the BWR water chemistry guidelines. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> action time period for exceeding the coolant chemistry limits described in. 3.2.3a and b ensures that prompt action is taken to restore coolant chemistry to normal operating levels. The requirement to commence a shutdown within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, and to be shutdown and reactor coolant temperature be reduced to < 200 degrees F within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> minimizes the potential for IGSCC crack growth. Reactor water samples are analyzed 6RiJY to ensure that reactor water quality remains within the BWR wate~guidelines. These samples are analyzed and compared~ action level 1 values. w-1periodicallyl ~ ~

The conductivity of the re ctor coolant is continuously monitored. The continuous conductivity monitor is visually checked--shtf<<y in accordance with procedur s. The monitor alarms at the local panel. The recorder, which is located in the Control Room, alarms in the Control Room. The sampl s of the coolant which are analyzed for conductivity daffy will serve as a comparison with the continuous conductivity monitor. The p imary sample point for the reactor water conductivity mples is the non-regenerative heat exchanger in the reactor water cleanup syste . An alternate sample point is the #11 recirculation lo p. The reactor coolant samples will also be used to determine the chloride ands l~teconcentrations. ~~~~.~~~~*~~~~~-~~~-~~~~~~~~~~~~~~~~

• * * . However, if the conductivity becomes abnor al (>0.19 µmho/cm), other than short term spikes, chloride and sulfate measure ents will be made within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to assure that then rmal limits(< 5 ppb of chloride or sulfate ions) are maintained. A short term spike is defined as a rise in conductivity(> 0.19 µmho/cm) such as that which could arise from injection of additional feedwater flow for a ration of approximately 30 minutes in time. These ctions will minimize the potential for IGSCC crack growth. I SERT 3 NMP1 will use Noble Metal Chem al Addition (NMCA) as a method to enhance the e ectiveness of Hydrogen Water Chemistry (HWC) in mitigating IGSCC. NMCA will res t in temporary increases in reactor coolant conduct vity values during and following application. During application, the conductivity limit sp citied in 3.2.3a and 3.2.3c.1 is increased to 20 µm o/cm. The application period includes post-NMCA injection cleanup activities conducte prior to returning the plant to power operation. A increase in conductivity is expected principally due to residual ionic species from the N A. However, these species have minor effects o IGSCC and are, therefore, acceptable. During NMCA, samples will be obtained fro the temporary skid which is placed in service duri g the NMCA injection process.

in accordance with the in accordance with the Surveillance Frequency AMENDMENT NO. 142. 16J. 16Q, 172 Surveillance Frequency 98 Control Program, and Control Program

TABLE 4.6.2a INSTRUMENTATION THAT INITIATES SCRAM Surveillance Requirement (1)

Parameter Manual Scram Sensor Check None

-==- \

~

~

Instrument Channel Test 0Roeperweek ~

Instrument CNote1 Channel Calibration I (I)

(2) High Reactor Pressure None iJnse per 3 ffi9AIAs" ~ ,r-fil ce per 3 ll19A!hs1 (3) High Drywell Pressure None jNote 1 ~nee per 3 months' j~ j \JQnoo p~ months~

(4) Low Reactor Water ~e/day ~0Roeper3ffleAtAs \ ~ °'1oeper3tl1eAtAs~

Level !Note 1 (5) High Water Level Scram None Discharge Volume !Note 0A00 ffiOAIAs !Note 1 0Ai9 pf 3 FReAtlos 1

(6) Main-Steam-Line Isolation Valve Position None Onoe per a months Once per operating cycle (7) Deleted

~ ~

AMENDMENT NO. 442, 149 201