BVY 08-075, Technical Specifications Proposed Change No. 273, Supplement 4, Response to Request for Additional Information Related to Supplement 2

Technical Specifications Proposed Change No. 273, Supplement 4, Response to Request for Additional Information Related to Supplement 2

ML083540188
Person / Time
Site:	Vermont Yankee
Issue date:	12/15/2008
From:	Michael Colomb Entergy Nuclear Operations, Vermont Yankee
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BVY 08-075, TAC MD8111
Download: ML083540188 (11)

Text

Entergy Nuclear Operations, Inc.

Vermont Yankee P.O. Box 0250 "EnteWgy 320 Governor Hunt Road Vernon, VT 05354 Tel 802 257 7711 December 15, 2008 BVY 08-075 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

References:

(a) Letter, VYNPS to USNRC, 'Technical Specification Proposed Change No. 273 Instrumentation Technical Specifications,"

BVY 08-001, dated February 12, 2008 (b) Letter, USNRC to VYNPS, "Vermont Yankee - Request for Additional Information Regarding Technical Specification Change Relating to Degraded Grid Protection System Instrumentation (TAC No. MD81 11)," NVY 08-080, dated August 19, 2008 (c) Letter, VYNPS to USNRC, "Technical Specification Proposed Change No. 273, Supplement 2, Response to Request For Additional Information," BVY 08-062, dated September 15 2008

Subject:

Vermont Yankee Nuclear Power Station License No. DPR-28 (Docket No. 50-271)

Technical Specifications Proposed Change No. 273, Supplement 4 Response to Request for Additional Information Related to Supplement 2

Dear Sir or Madam,

In Reference (a), Entergy Nuclear Operations Inc. (ENO) submitted a proposed change to the instrumentation sections of the Vermont Yankee Operating License Technical Specifications. Reference (b) provided a request for additional information that was addressed by ENO in Reference (c). to this submittal provides ENO's response to questions concerning the response provided in Reference (c) as discussed with the NRC on a telecom held on September 29, 2008.

This supplement to the original license amendment request does not change the scope or conclusions in the original application, nor does it change ENO's determination of no significant hazards consideration.

There are no new regulatory commitments being made in this letter.

Should you have any questions or require additional information concerning this submittal, please contact Mr. David J. Mannai at (802) 451- 3304.

%Oo(

BVY 08-075 / Page 2 of 2 I declare under penalty of perjury, that the foregoing is true and accurate. Executed on December 15, 2008.

Sincerely, Site Vice President Vermont Yankee Nuclear Power Station Attachments (2) cc: Mr. Samuel J. Collins (w/o Attachments)

Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406-1415 Mr. James S. Kim, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 0 8 C2A Washington, DC 20555 USNRC Resident Inspector Entergy Nuclear Vermont Yankee, LLC 320 Governor Hunt Road Vernon, Vermont 05354 Mr. David O'Brien, Commissioner VT Department of Public Service 112 State Street - Drawer 20 Montpelier, Vermont 05620-2601

Docket 50-271 BVY 08-075 Attachment 1 Technical Specification Proposed Change No. 273, Supplement 4 Vermont Yankee Nuclear Power Station Response to Request for Additional Information

BVY 08-075 / Attachment 1/page 1 of 4 Technical Specification Proposed Change No.273, Supplement 4 Vermont Yankee Nuclear Power Station Response to Request For Additional Information RAI No. 1 In response to RAI number 7, the Licensee added Note (d) to proposed Table 3.2.1.

Since a Note (d) already exists this should be Note (e).

Response to RAI No. 1 The subject note has been renamed to Note (e). A revised Table 3.2.1 is attached.

RAI No. 2 In response to RAI number 10, the Licensee indicated that Note (a), associated with Table 3.2.2 trip functions 3.c and 4.f, is applicable to Startup/Hot Standby, Hot Shutdown and Refuel Modes. The Staff believes that this note should only be required for the refuel mode.

Response to RAI No. 2 A revised proposed Table 3.2.2 applying Note (a) to only the refuel mode is provided in .

RAI No. 3 In support of the response to RAI number 5, submit a discussion regarding the RPS functional test in terms of whether or not the logic portion of the circuitry is bypassed. The licensee stated that the logic portion is not bypassed during the functional test. This information is necessary to confirm the licensee's assertion that the requested change is administrative in nature.

Response to RAI No. 3 The individual channel logics are not bypassed during the functional testing. During functional testing, each channel is tested from the sensor contact to the applicable RPS channel logic relay.

BVY 08-075 / Attachment 1/ page 2 of 4 Technical Specification Proposed Change No.273, Supplement 4 Vermont Yankee Nuclear Power Station Response to Request For Additional Information RAI No. 4 In support of RAI Numbers 1 (RPS response time) & 8 (PCIV response time), the Licensee stated that they would look into an NRC concern centered around response time of sensors in terms of reliability such that periodic testing may or may not be needed. This information would be needed to support NRC evaluation of TS changes that discuss response time.

Response to RAI No. 4 The current licensing basis (CLB) requires RPS response time testing of the logic system and does not include sensor response time testing.

Conservative sensor response time assumptions are included in the accident analysis and are provided as an input to General Electric via the Operating Parameters for Licensing (OPL) forms. The Rosemount transmitters used in the RPS system have adjustable damping that is capable of being set as high as 1.67 seconds and all transmitters in service have their damping set to a lower value. The accident analysis assumes a value of 1.8 seconds for the sensor response time. This provides margin in the analysis and based on this sensor response time testing is not performed.

The CLB for the Primary Containment Isolation System (PCIS) does not require response time testing of PCIS instruments and sensors.

Conservative instrumentation response time assumptions are provided as an input to General Electric via the OPL forms. The PCIS logic uses the same type of relays and instrumentation as the RPS Logic and would have a response time of < 50 milliseconds.

The transmitters used in the PCIS system have adjustable damping that is capable of being set as high as 1.67 seconds. That would give a total estimated response time of 1.72 seconds. 2.1 seconds is assumed and used in the accident response for the instrumentation response time.

This is considered to be acceptable based upon, the NRC response to NEDO 32291-A Supplement 1. This report demonstrated that for the' instruments used in PCIS and ECCS and the sensors for RPS, the instruments do not have the potential for significant

,response time degradation before failure. For instrument loops meeting the application criteria of the License Topical Report, performance of ongoing TS required surveillance tests other than response time testing (i.e. calibration test, functional rest and logic system functional test) provides adequate assurance that those instrumentation loops will meet their respective response time requirements.

BVY 08-075 / Attachment 1/ page 3 of 4 Technical Specification Proposed Change No.273, Supplement 4 Vermont Yankee Nuclear Power Station Response to Request For Additional Information RAI No. 5 VY stated that during OPDRVs, EDG startup in a standby status was not required.

However, the VY response to RAI number 7 has the "Also required to initiate diesel generators" Note in the "Required Channels per Trip System" column for the Core Spray System. As a result the capability to initiate the EDG would be required during all applicable Modes, including "When associated ECCS subsystem is required to be operable per specification 3.5," per Note (b) on Function 1.b. There appears to be a discrepancy between the RAI response, and therefore the TS requirement for applicability during OPDRVs, and the statement made during the phone call. Clarification is needed.

Response to RAI No. 5 As a clarification, the VY response to RAI number 7 added the "Also required to initiate diesel generators" Note in the "Required Channels per Trip System" column for the Core Spray System High Drywell pressure and Low-Low Reactor Vessel Water Level trip functions. NRC noted that the new note implied the capability to initiate the EDG during all applicable Modes specified.

During shutdown, TS Section 3.5.H allows multiple core and containment cooling systems to be out of service since there is no possibility of a LOCA. A LOCA can only happen when there is pressure in the reactor vessel. Additionally, a high drywell pressure condition cannot occur during shutdown. The concern during this operating condition is a drain down of the reactor; therefore the operability requirements are tied to performance of operations with the potential to drain the reactor vessel (OPDRV).

TS Section 3.5.H bases states: "...if OPDRVs are in progress with irradiated fuel in the reactor vessel, operability of low pressure ECCS injection/spray subsystems is required to ensure capability to maintain adequate reactor vessel water level in the event of an inadvertent vessel drain down. In this condition, at least 300,000 gallons of makeup water must be available to assure core flooding capability. In addition, only one diesel generator associated with one of the ECCS injection/spray subsystems is required to be operable in this condition since, upon loss of normal power supply, one ECCS subsystem is sufficient to meet this function."

The low pressure ECCS injection/spray subsystems consist of two core spray (CS) and two low pressure coolant injection (LPCI) subsystems. TS Section 3.5.H would allow both CS systems to be inoperable and therefore the instrumentation that provides the EDG start would also be inoperable. EDG start during this condition would either rely on an alternate EDG initiation signal (e.g., Loss of Normal Power initiated) or manual initiation.

Based on this, the subject note (i.e. Note (e)) is being modified to recognize that the subject instrumentation is required to initiate the EDGs when the Core Spray system is required to be operable per TS Section 3.5.

BVY 08-075 / Attachment 1/ page 4 of 4 Technical Specification Proposed Change No.273, Supplement 4 Vermont Yankee Nuclear Power Station Response to Request For Additional Information RAI No. 6 The UFSAR states that the EDG start on high drywell pressure and low-low reactor water level. It is unclear what trip units cause this to occur. Is it the trip actuation circuit from Core Spray that causes the EDG to start? Does the LPCI trip actuation circuit from stated sensors result in EDG actuation? Or is there some other trip actuation circuit that is independent of the trip actuation of LPCI / Core spray that causes the EDG to start? Where in the FSAR / TS Bases is this discussed?

Response to RAI No. 6 The VY Updated Final Safety Analysis Report (UFSAR) Section 8.5.3 states that both diesel generator units are started automatically by low-low reactor water level and/or high drywell pressure. Additional detail on how this is accomplished is provided in the proposed basis for TS Section 3.2.A/4.2.A. The basis provides a description of the initiation logic and notes that the EDGs receive their initiation signal from the CS system initiation logic. The high drywell pressure and low-low reactor level instruments do not directly start the diesels. The instrument provides signals to actuate of the Core Spray logic. The actuation of the Core Spray logic provides the diesel start signal. The LPCI trip actuation logic does not result in EDG actuation.

As described in UFSAR Section 8.5.4, the EDGs are also started during a degraded emergency bus voltage condition if an accident occurs or if a total loss of normal power occurs. Additional detail is provided in the proposed basis for TS Section 3.2.K/4.2.K.

Should the EDG ignition logic be inoperable, the EDGs are capable of being started from the Control Room.

TS Section 4.1O.A.1. requires the following testing of the EDGs:

• Monthly, each diesel generator must be manually started using the under-voltage automatic starting circuit, the speed increased from idle to. synchronous and then gradually loaded to expected maximum emergency loading not to exceed the continuous rating to demonstrate operational readiness. The test shall continue for a minimum period of one hour.

" Once every six months, in lieu of Specification 4.10.A.l.a.1, each diesel generator must be manually started using the under-voltage automatic starting circuit and loaded to demonstrate that it will reach rated frequency and voltage within specified time limits. The diesel generator shall then be gradually loaded to expected maximum emergency loading not to exceed the continuous rating and run for a minimum period of one hour. The time taken to reach the rated frequency and voltage shall be logged.

" Once per operating cycle, the actual conditions under which the diesel generators are required to start automatically will be simulated and a test conducted to demonstrate that they will start within 13 seconds and accept the emergency loads and start each load within the specified starting time. The results shall be logged.

Docket 50-271 BVY 08-075 Attachment 2 Technical Specification Proposed Change No. 273, Supplement 4 Vermont Yankee Nuclear Power Station Revised Technical Specification Pages

VYNPS Table 3.2.1 (page 1 of 4)

Emergency Core Cooling System Instrumentation ACTIONS WHEN REQUIRED REQUIRED APPLICABLE MODES OR CHANNELS CHANNELS OTHER SPECIFIED PER TRIP ARE TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE TRIP SETTING

1. Core Spray System

a. High Drywell RUN, STARTUP/HOT 2 Note 1

• 2.5 psig Pressure STANDBY, HOT SHUTDOWN, Refuel (a), (e)

b. Low-Low RUN, STARTUP/HOT 2 Note 1 Ž 82.5 inches Reactor STANDBY, HOT SHUTDOWN, Vessel Water Refuel (a), (b), (e)

Level

c. Low Reactor RUN, STARTUP/HOT Note 2 Ž 300 psig and Pressure STANDBY, HOT SHUTDOWN,

• 350 psig (Initiation) Refuel (a), (b) 1

d. Low Reactor RUN, STARTUP/HOT Note 2 Ž 300 psig and Pressure STANDBY, HOT SHUTDOWN,  : 350 psig (System Ready Refuel (a), (b) and Valve Permissive) 2

e. Pump Start RUN, STARTUP/HOT, Note 2 Ž 8 seconds and Time Delay STANDBY, HOT SHUTDOWN, 10 seconds Refuel (a), (b)

f. Pump RUN, STARTUP/HOT Note 8 Ž 100 psig Discharge STANDBY(c), HOT per Pressure SHUTDOWN (C) ,Refuel (C) pump

g. Auxiliary RUN, STARTUP/HOT 1 Note 2 NA Power Monitor STANDBY, HOT SHUTDOWN, Refuel(a), (b)

h. Pump Bus RUN, STARTUP/HOT 1 Note 2 NA Power Monitor STANDBY, HOT SHUTDOWN, Refuel (a), (b)

(a) With reactor coolant temperature > 212 'F.

(b) When associated ECCS subsystem is required to be operable per specification 3.5.

(c) With reactor steam pressure > 150 psig.

(e) Required to initiate the emergency diesel generators when core spray is required to be operable per specification 3.5.

Amendment No. 35

VYNPS Table 3.2.2 (page 2 of 3)

Primary Containment Isolation Instrumentation ACTIONS WHEN ACTIONS APPLICABLE MODES REQUIRED REQUIRED REFERENCED OR OTHER CHANNELS CHANNELS FROM SPECIFIED PER TRIP ARE ACTION TRIP TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE NOTE 1 SETTING.

3. High Pressure Coolant Injection (HPCI) System Isolation a.High Steam Line RUN, STARTUP/HOT 6 Note 1 Note 2.d

• 196 OF Space STANDBY, HOT Temperature SHUTDOWN, Refuel(a) b.High Steam Line RUN, STARTUP/HOT Note 1 Note 2.d

• 195 d/p (Steam Line STANDBY, HOT inches of Break) SHUTDOWN, Refuei(a) water c.Low Steam RUN, STARTUP/HOT 4 Note 1 Note 2.d Ž 70 psig Supply Pressure STANDBY, HOT SHUTDOWN, Refuel(a) d.High Main Steam RUN, STARTUP/HOT 2 Note 1 Note 2.d 5 200 OF Line Tunnel STANDBY, HOT Temperature SHUTDOWN, Refuel (a) e.High Main Steam RUN, STARTUP/HOT 1 Note 1 Note 2.d

• 35 Line Tunnel STANDBY, HOT minutes Temperature Time SHUTDOWN, Refuel(a)

Delay

4. Reactor Core Isolation Cooling (RCIC) System Isolation a.High Main Steam RUN, STARTUP/HOT 2 Note 1 Note 2.d

• 200 OF Line Tunnel STANDBY, HOT Temperature SHUTDOWN, Refuel (a) b.High Main Steam RUN, STARTUP/HOT Note 1 Note 2.d S 35 Line Tunnel STANDBY, HOT minutes Temperature Time SHUTDOWN, Refuel(a)

Delay c.High Steam Line RUN, STARTUP/HOT 6 Note I Note 2.d S 196 OF Space STANDBY, HOT Temperature SHUTDOWN, Refuel(a)

(a) With reactor coolant temperature > 212 OF.

Amendment No. 45

VYNPS Table 3.2.2 (page 3 of 3)

Primary Containment Isolation Instrumentation ACTIONS WHEN ACTIONS APPLICABLE MODES REQUIRED REQUIRED REFERENCED OR OTHER CHANNELS CHANNELS FROM SPECIFIED PER TRIP ARE ACTION TRIP TRIP FUNCTION CONDITIONS SYSTEM INOPERABLE NOTE 1 SETTING

4. RCIC System Isolation (Continued) d.High Steam Line RUN, STARTUP/HOT 1 Note 1 Note 2.d

• 195 d/p (Steam Line STANDBY, HOT inches of Break) SHUTDOWN, Refuel (a) water e.High Steam Line RUN, STARTUP/HOT 1 Note 1 Note 2.d Ž 3 seconds d/p Time Delay STANDBY, HOT and SHUTDOWN, Refuel(a) 7 seconds f.Low Steam RUN, STARTUP/HOT 4 Note 1 Note 2.d Ž 50 psig Supply Pressure STANDBY, HOT SHUTDOWN, Refuel (a)

5. Residual Heat Removal Shutdown Cooling Isolation a.High Reactor RUN, STARTUP/HOT 1 Note 1 Note 2.d .5 150 psig Pressure STANDBY, HOT SHUTDOWN, Refuel (a)

(a) With reactor coolant temperature > 212 OF.

Amendment No. 46