GNRO-2007/00061, (Ggns), Supplement to Amendment Request Changes to the Condensate Storage Tank Level-Low Setpoints

(Ggns), Supplement to Amendment Request Changes to the Condensate Storage Tank Level-Low Setpoints

ML072550241
Person / Time
Site:	Grand Gulf
Issue date:	09/05/2007
From:	Brian W Entergy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2007/00061, TAC MD4675
Download: ML072550241 (8)

Text

Entergy

"'EnteW P.O. Box 756 Port Gibson, MS 39150 Tel 601 437 6409 William R. Brian Vice President - Operations Grand Gulf Nuclear Station GNRO-2007/00061 September 5, 2007 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Supplement to Amendment Request Changes to the Condensate Storage Tank Level-Low Setpoints Grand Gulf Nuclear Station, Unit I (GGNS)

Docket No. 50-416 License No. NPF-29

REFERENCE:

Letter GNRO-2007/00016 from W. R. Brian, Entergy Operations, Inc.,

to Document Control Desk, USNRC, "License Amendment Request Condensate Storage Tank Level-Low Setpoint Change," dated March 1, 2007 (TAC # MD 4675)

Dear Sir or Madam:

By the above referenced letter, Entergy Operations, Inc. (Entergy) proposed a change to the Grand Gulf Nuclear Station, Unit I (GGNS) Technical Specifications (TS) to incorporate the corrected allowable values in TS Tables 3.3.5.1-1 and 3.3.5.2-1.

Entergy and members of your staff held calls to discuss the technical basis for the proposed TS change. As a result of the calls, six questions were determined to need formal response: four of the questions from the Instrumentation and Controls Branch and two questions from the Reactor Systems Branch. Entergy's response is contained in Attachment 1.

There are no technical changes proposed. The original no significant hazards consideration included in the referenced letter is not affected by any information contained in the supplemental letter. There are no new commitments contained in this letter.

If you have any questions or require additional information, please contact Matt Crawford at 601-437-2334.

GNRO-2007/00061 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on September 5, 2007.

Sincerely, WRB/MLC Attachments:

1. Response to Request For Additional Information

2. Sketch of HPCS/RCIC CST Suction Piping Diagram

3. GGNS Calculation JC-Q1E22-N654-1, Rev. 3

4. GGNS Calculation JC-Q1E51-N635-1, Rev. 1

5. GGNS Calculation M6.7.013, Rev. 1 cc: Dr. Bruce S. Mallett Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 61 1 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U.S. Nuclear Regulatory Commission ATTN: Mr. Bhalchandra Vaidya, NRR/DORL (w/2)

ATTN: ADDRESSEE ONLY ATTN: U.S. Postal Delivery Address Only Mail Stop OWFN/O-7D1A Washington, D.C. 20555-0001 Mr. Brian W. Amy, MD, MHA, MPH Mississippi Department of Health P. 0. Box 1700 Jackson, MS 3921 5-1700 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

Attachment 1 To GNRO-2007100061 Response to Request for Additional Information to GNRO-2007/00061 Page 1 of 3 Response to Request for Additional Information Related to Revising the Condensate Storage Tank Level-Low Setpoints

1. Please provide the physical layout diagrams of the condensate storage tank (CST) and the non-safety related HPCS/RCIC suction piping associated with the new calculated values. Also, provide the calculations used to determine the new values.

Response

A simplified diagram of the piping arrangement labeled "HPCS/RCIC CST SUCTION PIPING DIAGRAM" is provided as Attachment 2. Copies of the setpoint calculations JC-Q1E22-N654-1, Rev 3 (Attachment 3) and JC-Q1E51-N635-1, Rev 1 (Attachment

4) are also included.

2. Please provide the calculations, including the physical dimensions of the CST, used to determine that for a station blackout the 115,278 gallons required to cope with a four hour SBO event is still satisfied with the new setpoint value.

Response

A copy of calculation M6.7.013, Rev 1 is provided as Attachment 5. Per this calculation, a CST level of 18.9 ft (standpipe elevation) ensures a usable reserve volume of 143,000 gallons at 800 gpm with the new 4 ft setpoint.

3. Setpoint Calculation Methodology: Provide documentation (including sample calculations) of the methodology used for establishing the limiting setpoint (or NSP) and the limiting acceptable values for the As-Found and As-Left setpoints as measured in periodic surveillance testing described below. Indicate the related Analytical Limits and other limiting design values (and the sources of these values) for each setpoint.

Response

The Nominal Trip Setpoint (NTSP) values and the proposed tech spec allowable values are derived in calculations JC-Q1E22-N654-1 (Attachment 3) and JC-Q1E51-N635-1 9 (Attachment 4). The GE NEDC 31336P-A setpoint methodology is utilized.

The As-Found and As-Left tolerances of the setpoints specified in the plant surveillance procedures 06-IC-1 E22-Q-0002, 06-IC-1 E51 -Q-0002, 06-IC-1 E22-R-0002 and 06-IC-1 E51-R-0002 is 0.25% span per JC-Q1E22-N654-1 and JC-Q1E51-N635-1. The associated analytical limits are also discussed in the calculations.

4. Safety Limit (SL)-Related Determination: Provide a statement as to whether or not the setpoint is a limiting safety system setting for a variable on which a safety limit (SL) has been placed as discussed in 10 CFR 50.36(c)(1)(ii)(A). Such setpoints are described as "SL-Related" in the discussions that follow. In accordance with 10 CFR 50.36(c)(1)(ii)(A), the following guidance is provided for identifying a list of functions to GNRO-2007/00061 Page 2 of 3 to be included in the subset of LSSSs specified for variables on which SLs have been placed as defined in Standard Technical Specifications (STS) Sections 2.1.1, Reactor Core SLs and 2.1.2, Reactor Coolant System Pressure SLs. This subset includes automatic protective devices in TSs for specified variables on which SLs have been placed that: (1) initiate a reactor trip; or (2) actuate safety systems. As such these variables provide protection against violating reactor core safety limits, or reactor coolant system pressure boundary safetý limits.

Examples of instrument functions that might have LSSSs included in this subset in accordance with the plant-specific licensing basis, are pressurizer pressure reactor trip (pressurized water reactors), rod block monitor withdrawl blocks (boiling water reactors), feedwater and main turbine high water level trip (boiling water reactors), and end of cycle recirculation 'pump trip (boiling water reactors).

For each setpoint, or related group of setpoints, that you determined not to be SL-Related, explain the basis for this determination.

Response

The affected setpoints are not "SL-Related" since they do not meet the specified criteria.

TS 2.1.1, Reactor Core Safety Limits, requires the reactor vessel water level to be greater than the top of active irradiated fuel (TAF). A Loss of Feedwater Event described in GGNS Updated FSAR section 15A.6.3.3, Event 20 and in section 15.2.7 assumes that initial core cooling and reactor water level are maintained by either HPCS or RCIC. RCIC and HPCS are automatically actuated in this event by the reactor water level -Low Low, Level 2 signal. The CST low level signal is not required for HPCS/RCIC actuation or reactor vessel injection. The CST low level signal along with the associated automatic transfer function ensures a continuous inventory supply for the HPCS and RCIC systems after the reactor vessel is refilled as part of its long-term cooling function.

The CST level setpoint is based on ensuring the suction source transfer is performed and is not related to protecting the TAF or any other safety limit. Therefore, this setpoint is not SL-related.

5. For setpoints that are determined to be SL-Related: The NRC letter to the NEI SMTF dated September 7, 2005 (ML052500004), describes Setpoint-Related TS (SRTS) that are acceptable to the NRC for instrument settings associated with SL-related setpoints. Specifically: Part "A" of the Enclosure to the letter provides LCO notes to be added to the TS, and Part "B" includes a check list of the information to be provided in the TS Bases related to the proposed TS changes.

a. Describe whether and how you plan to implement the SRTS suggested in the September 7 letter. If you do not plan to adopt the suggested SRTS, then explain how you will ensure compliance with 10 CFR 50.36 by addressing items 3b and 3c, below.

b. As-Found Setpoint evaluation: Describe how surveillance test results and associated TS limits are used to establish operability of the safety system. Show that this evaluation is consistent with the assumptions and results of the setpoint calculation methodology. Discuss the plant corrective action processes (including plant procedures) for restoring to GNRO-2007/00061 Page 3 of 3 channels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the criteria for determining operability of the instrument being tested are located in a document other than the TS (e.g. plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

c. As-Left Setpoint control: Describe the controls employed to ensure that the instrument setpoint is, upon completion of surveillance testing, consistent with the assumptions of the associated analyses. If the controls are located in a document other than the TS (e.g. plant test procedure) explain how the requirements of 10 CFR 50.36 are met.

Response

The affected setpoints are not "SL-Related"; therefore, this question is Not Applicable.

6. For setpoints that are not determined to be SL-related: Describe the measures to be taken to ensure that the associated instrument channel is capable of performing its specified safety functions in accordance with applicable design requirements and associated analyses. Include in your discussion information on the controls you employ to ensure that the as left trip setting after completion of periodic surveillance is consistent with your setpoint methodology. Also, discuss the plant corrective action processes (including plant procedures) for restoring channels to operable status when channels are determined to be "inoperable" or "operable but degraded." If the controls are located in a document other than the TS (e.g., plant test procedure), describe how it is ensured that the controls will be implemented.

Response

The trip unit setpoints are checked approximately every 92 days per surveillance procedures 06-IC-1 E22-Q-0002, 06-IC-1 E51 -Q-0002. The transmitter calibration and trip unit calibration is checked approximately every 18 months per surveillance procedures 06-IC-1E22-R-0002 and 06-IC-1E51-R-0002. The As-Found and As-Left tolerance of the transmitter and trip units is 0.25% span per JC-Q1E22-N654-1 (Attachment 4) and JC-Q1E51-N635-1 (Attachment 5). If one of the transmitters or trip units is found out of calibration, it will be recalibrated per the surveillance procedure. In addition, a condition report (CR) will be initiated if the Technical Specification allowable value is exceeded.

Actions taken with an inoperable channel are in accordance with the Technical Specification.

Attachment 2 To GNRO-2007/00061 Sketch of HPCS/RCIC CST Suction Piping Diagram

40' I.D.

NON-SAFETYSýSTEM STAND PIPES-IPIIAOo2 (CST)

,,-AUA. TUILDINDWALL

-HPCSCC TEST RETURN LINE 10.90 FT.

9.413FT.

-=Milf-11---lil-lliff WIl~i--ll--M--l 011-=- CST DIKE AREA 3.0 F T. = II-IU -UI SAFETYRELATEDO ý-ýNON-OAFITV ýRELATESI 1.68 FT

-D.33 FT. Il*U-qI-Ir-Jlll-I.... '=[l_--1:

l =-Illi AVORTSEXA "JlIM ;Ei"- .REAVER-viIm

-l-----nI.I*II I1.A FT.

- *U~ill I 1

S~~~ . ~ ~ ~~~~~

"I.08 FII

?

Ill U1ll

-1.16 FT.

1liffiiff

-=fill-= [Ill=-Jill L. alill

-- 11110=1111 1111 Jill=

miI -l TRANSMITTERS IE2ZN054C&G I.S' 1, ý-

E5IN035A&E I 3- 125 3' 1.51 NOTES, I. DRAWING IS VERY SIMPLIFIED AND IS NOT TO SCALE.

ONLYMAJORCOMPONENTS/PIPING ARE INCLUDED.

2. ELEVATIONSARE REFERENCEDTO CST LEVEL TRANSMITTER INSTRUMENT ZERO THAT IS 1.08 FT. ABOVETHE BOTIOM OF 6". S"A 105 THE CST.

REFERENCES.

1. C-143.0-NIPIIA002-I.3-014 REV.000. CSO 12" STANDPIPE(NOZZLEEl.

2. C-143.0-NIPlIA002-1.3-0I5 REV.002. CST 12" STANDPIPE(NOZZLEFl.

3. C-143.0-NIPlIAA02-I.3-016 REV.002. CST 12" STANDPIPE[NOZZLEG).

4. C-143.0-N1PIIA002-I.3-018 REV.002, CST 10" STANDPIPE (NOZZLEJ).

S. C-143.0-NIPIIA002-I.3-0IT REV.002. CST I1" OVERFLOW (NOZZLELI.

20.,R 6. C-143.0-NIPIIA000-I.3-017 REV.002. CST 20" HPCS/RCIC SUCTION A REDUCER (NOZZLEHA.

7. C.143.1;-hIPIIA002 "1.3"*0 2 REV.003. CST OUTLINEDRAWING.

IE5IFOIO 0. M-10S5 RE% 041. CONDENSATE AND REFUELINGWATERSTORAGEAND TRANSFERýfSTEM P&iD.

9. M-1003R R.V. 033, RCIC SYSTEM P&ID.

10. M-1086 REA.030. HPCS SYSTEM P&ID.

II. M-1400 RE':. 016. YARDPIPING CST AND RWSTTANKAREA.

M 1E72F0O1 I2. M- 414 PE;, OO. YARDPIPING SECTIONSAND DETAILS.

13. M-13368 019. APCS AND RCIC PIPING ISOMETRIC.

• LV.

14. J-IGSOB LEVEL SETTING DIAGRAM. DORN 05-1560.

FROM SUPPRESSION IS. MS-03 REV. I. MECHANICAL PIPING CLASS STANDARD.

POOL IA. CR-GGN-2001-01367. CONDITIONREPORT.

IE51FO31 PROM SUPPRESSION RCIC PUMP 0- POOL IE51CO01 IE22FOI1

-26.37 FT.

TREACTOR --

/CST 000 SIP HPCS PUMP IE22C001

-27.10 FT. GRAND GULF NUCLEAR SIATION UNIT I TO REACTOR

/CST NUCLEAR PLANT ENGINEERING HPCS/RCIC CST SUCTION PIPING DIAGRAM SCALE: NONE SKETCHN-,

VGA.

DEN: t-sMketch.d 9 n,