JAFP-02-0124, Revision K to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications
| ML021750502 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 06/11/2002 |
| From: | Ted Sullivan Entergy Nuclear Northeast, Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| JAFP-02-0124 | |
| Download: ML021750502 (91) | |
Text
ITS(Updated SUBMITTAL 5/31/01 03/28/01)
Copy Number Assigned to Original ITS Project Repro Master Licensing
____________1__ NRC - Doc'Control Desk 2 NRC - G. Vissing 3 NRC - William Beckner 4 NRC - William Beckner 5 Regional Administrator 6 William Flynn - NYSERDA 8 Verne Childs 9 John Hoddy (ITS) 10 Licensing Manager 11 Licensing (WPO) - Kokolakis 12 Reference Center (JAF) 13 OPS - Phil Russell 14 OPS - Henry Borick 15 Training - Gary Fronk 16 Training - Gary Fronk 17 I&C - Steve Juravich (ITS) 19 Tech Services (ITS - Brian) 20 Tech Support/Sys Eng (Tina) 21 Design Eng (ITS - Tom) 23 Excel - Fred Mizell 24 Excel - Don Hoffman 26 NRC Res. Insp. (Specs & Bases Only) 29 Simulator Control Room Vol.s 2,3 & 4 Only
- PTR Group (Vol's 2-4 only)
- Planning - Dan Johnson (Vol's 2-4 only)
- Excel - Jerry Jones
- Excel - Phil Ballard
- Excel - Gregg Ellis 0 L
RIP - Anne Stark
-__ITS - Doug ITS - Chris ITS - Phil I&C - Mark Cronk
-_ITS- Dale
-__ITS - Ken
Entergy Nuclear Northeast Entergy Nuclear Operations. Inc.
James A. Fitzpatrick NPP P.O. Box 110 EntogyLycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 June 11, 2002 T.A. Sullivan Vice President, Operations-JAF JAFP-02-0124 United States Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59 Revision K to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications
References:
see last page of letter
Dear Sir,
This letter and the associated attachments provides Revision Kto the previously submitted application for amendment to the James A. FitzPatrick Technical Specifications (Reference 1),
as supplemented by References 2, 3, 4, 5, and 7 for converting the current Technical Specifications (CTS) to the Improved Technical Specifications (ITS) consistent with the Improved Standard Technical Specifications (NUREG-1433, Revision 1).
Revision K (Attachment 1) to the Reference 1, 2, 3, 4, 5, and 7 submittals include certain changes requested by the NRC Staff as a result of their review of Revision J (Reference 7).
The submittal also provides revised pages based on the NRC approval of TS Amendment 273.
Additionally, minor changes are made to correct editorial errors in the previous submittals.
Each Chapter/Section includes a summary of the changes to the affected Chapter/Section.
The Insert and Discard Instructions are included in Attachment 2 to allow merging Revision K with the existing submittal. The clean typed ITS and Bases in Volumes 2, 3, and 4, and the CTS markup pages in CTS order in Volume 5 are not being updated since these Volumes are duplicates of each individual Specification located in Volumes 6 through 19.
United States Nuclear Regulatory Commission Attn: Document Control Desk
Subject:
Revision K to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications Page -2 There are no new commitments contained in this letter. Should you have any questions, please contact Mr. Andrew Halliday at (315) 349-6055.
Very Truly Yours, T. A. Sullivan Operations Vice President, Oeain Attachments: 1) Revision K to the JAF ITS Submittal
- 2) Insert and Discard Instructions cc:
Regional Administrator Mr. N. B. Le U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission 475 Allendale Road Mail Stop O-7H3 King of Prussia, PA 19406 Washington, DC 20555 P. 0. Box 134 Mr. Guy Vissing, Project Manager Resident Inspector's Office Project Directorate I James A. FitzPatrick Nuclear Power Plant Division of Licensing Project Management U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission P. 0. Box 134 Mail Stop 8C2 Lycoming, NY 13093 Washington, DC 20555 Mr. William M. Flynn New York State Energy Research and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, New York 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire Plaza Albany, New York 12223 Mr. William D. Beckner, Chief Technical Specifications Branch U. S. Nuclear Regulatory Commission Mail Stop O-7H3 Washington, DC 20555
United States Nuclear Regulatory Commission Attn: Document Control Desk
Subject:
Revision K to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications Page -3
References:
- 1. NYPA letter, J. Knubel to USNRC Document Control Desk, Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JPN-99-008), dated March 31, 1999 (TAC No. MA5049)
- 2. NYPA letter, J. Knubel to USNRC Document Control Desk, Revision B to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JPN-99-018), dated June 1, 1999 3.' NYPA letter, Michael J. Colomb to USNRC Document Control Desk, Revision C to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JAFP-99-0278), dated October 14, 1999
- 4. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, Revisions D, E, F, G, and H to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JAFP-01 0133), dated May 31, 2001
- 5. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, Revision I to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications (JAFP-01-0234), dated October 18, 2001
- 6. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, James A. FitzPatrick (JAF) Improved Technical Specifications (ITS) Submittal (JAFP-02 0029), dated February 6, 2002
- 7. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, Revision J to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications (JAFP-02-0098), dated April 26, 2002
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of )) Docket No. 50-333 Entergy Nuclear Operations, Inc.
James A. FitzPatrick Nuclear Power Plant )
APPLICATION FOR AMENDMENT TO OPERATING LICENSE Entergy Nuclear Operations, Inc. requests an amendment to the Technical Specifications (TS) contained in Appendix A to Facility Operating License DPR-59 for the James A. FitzPatrick Nuclear Power Plant. This application is filed in accordance with Section 10 CFR 50.90 of the Nuclear Regulatory Commission's regulations.
This application for amendment to the FitzPatrick Technical Specifications proposes to convert the FitzPatrick current Technical Specifications (CTS) to be consistent with the Improved Standard Technical Specifications (ISTS) in NUREG-1433, Revision 1, dated April 1995. The proposed license amendment request was prepared considering the guidance of Nuclear Energy Institute (NEI) NEI 96-06, "Improved Technical Specifications Conversion Guidance,"
dated August 1996.
The Proposed license amendment request to convert the FitzPatrick CTS to the FitzPatrick Improved Technical Specifications (ITS) is enclosed with this application.
Entergy Nuclear Operations, Inc. STATE OF NEW YORK COUNTY OF OSWEGO Subscribed and sworn to before me this .14Ž-__day of"3-, nC52002.
T. A. Sullivan / Notary Public Vice President, Operations-JAF
?IIE S. DVSTYiAi 4887051 No",y pubit Startof Rwywk Ow6wgo Couanl .;ý,C My Commifmon Exores Jun 30, IM
ATTACHMENT 2 JAMES A. FITZPATRICK NPP IMPROVED TECHNICAL SPECIFICATIONS REVISION "K" INSERT AND DISCARD INSTRUCTIONS I of 1
JAMES A. FITZPATRICK NPP IMPROVED TECHNICAL SPECIFICATIONS REVISION "K" INSERT AND DISCARD INSTRUCTIONS REMOVE INSERT VOLUME 8 DOCs for ITS 3.3.1.1 pg 9 of 25 DOCs for ITS 3.3.1.1 pg 9 of 25 NUREG Bases markup for ITS 3.3.1.1 pg NUREG Bases markup for ITS 3.3.1.1 pg Insert Page B 3.3-30 Insert Page B 3.3-30 Bases JFDs for ITS 3.3.1.1 pgs I of 4 Bases JFDs for ITS 3.3.1.1 pgs I of 4 through 4 of 4 through 4 of 4 Retyped ITS 3.3.1.1 Bases pgs B 3.3-33 Retyped ITS 3.3.1.1 Bases pgs B 3.3-33 through B 3.3-37 through B 3.3-37 CTS markup for ITS 3.3.2.1 pgs 6 of 10 CTS markup for ITS 3.3.2.1 pgs 6 of 10 and 8 of 10 and 8 of 10 DOCs for ITS 3.3.2.1 pgs 1 of 9 through 9 DOCs for ITS 3.3.2.1 pgs 1 of 9 through 9 of 9 of 9 NUREG ITS markup for ITS 3.3.2.1 pgs NUREG ITS markup for ITS 3.3.2.1 pgs 3.3-19 and 3.3-20 3.3-19 and 3.3-20 JFDs for ITS 3.3.2.1 pgs 1 of 3 and JFDs for ITS 3.3.2.1 pgs 1 of 3 and 2 of 3 2 of 3 NUREG Bases markup for ITS 3.3.2.1 pg NUREG Bases markup for ITS 3.3.2.1 pg B 3.3-54 B 3.3-54 N/A NUREG Bases markup for ITS 3.3.2.1 pg Insert Page B 3.3-54 Retyped ITS 3.3.2.1 pgs 3.3-18, 3.3-19 Retyped ITS 3.3.2.1 p 3.3-18. 3.3-19 and and 3.3-20 3.3-20 Retyped ITS 3.3.2.1 Bases pgs B 3.3-56 Retyped ITS 3.3.2.1 Bases pgs B 3.3-56 through B 3.3-59 through B 3.3-59 CTS markup for ITS 3.3.4.1 pgs 2 of 6 and CTS markup for ITS 3.3.4.1 pgs 2 of 6 and 4 of 6 4 of 6 DOCs for ITS 3.3.4.1 pgs 3 of 8 and DOCs for ITS 3.3.4.1 pgs 3 of 8 and 4 of 8 4 of 8 NUREG ITS markup for ITS 3.3.4.1 pg 3.3- NUREG ITS markup for ITS 3.3.4.1 pg 3.3 35 35 NUREG Bases markup for ITS 3.3.4.1 pg NUREG Bases markup for ITS 3.3.4.1 pg Insert page B 3.3-94 Insert page B 3.3-94 NUREG Bases markup for ITS 3.3.4.1 pgs NUREG Bases markup for ITS 3.3.4.1 pgs B 3.3-96, B 3.3-98 and B 3.3-100 B 3.3-96, B 3.3-98 and B 3.3-100 NUREG Bases markup for ITS 3.3.4.1 pg NUREG Bases markup for ITS 3.3.4.1 pg Insert page B 3.3-100 Insert page B 3.3-100 Retyped ITS 3.3.4.1 pg 3.3-31 Retyped ITS 3.3.4.1 pg 3.3-31 Retyped ITS 3.3.4.1 Bases pgs B 3.3-90 Retyped ITS 3.3.4.1 Bases pgs B 3.3-90 and B 3.3-92 and B 3.3-92 Retyped ITS 3.3.4.1 Bases pgs B 3.3-94 Retyped ITS 3.3.4.1 Bases pgs B 3.3-94 through B 3.3-96 through B 3.3-96 1 of 1
JAMES A. FITZPATRICK NPP IMPROVED TECHNICAL SPECIFICATIONS REVISION "K" INSERT AND DISCARD INSTRUCTIONS REMOVE I INSERT VOLUME 9 CTS markup for ITS 3.3.6.1 pg 3 of 22 CTS markup for ITS 3.3.6.1 pg 3 of 22 DOCs for ITS 3.3.6.1 pgs 7 of 25 and DOCs for ITS 3.3.6.1 pgs 7 of 25 and 22 of 25 22 of 25 NSHCs for ITS 3.3.6.1 pgs 19 of 32 and NSHCs for ITS 3.3.6.1 pgs 19 of 32 and 20 of 32 20 of 32 NUREG ITS markup for ITS 3.3.6.1 pg 3.3- NUREG ITS markup for ITS 3.3.6.1 pg 3.3 57 57 NUREG ITS markup for ITS 3.3.6.1 pg NUREG ITS markup for ITS 3.3.6.1 pg Insert Page 3.3-58 Insert Page 3.3-58 JFDs for ITS 3.3.6.1 pg 4 of 5 JFDs for ITS 3.3.6.1 pg 4 of 5 NUREG Bases markup for ITS 3.3.6.1 pg NUREG Bases markup for ITS 3.3.6.1 pg Insert Page B 3.3-158 Insert Page B 3.3-158 NUREG Bases markup for ITS 3.3.6.1 pg NUREG Bases markup for ITS 3.3.6.1 pg Insert Page B 3.3-161 Insert Page B 3.3-161 NUREG Bases markup for ITS 3.3.6.1 pg NUREG Bases markup for ITS 3.3.6.1 pg Insert Page B 3.3-164b Insert Page B 3.3-164b Retyped ITS 3.3.6.1 pgs 3.3-52 and 3.3-53 Retyped ITS 3.3.6.1 pgs 3.3-52 and 3.3-53 Retyped ITS 3.3.6.1 Bases pg B 3.3-160 Retyped ITS 3.3.6.1 Bases pg B 3.3-160 Retyped ITS 3.3.6.1 Bases pg B 3.3-164 Retyped ITS 3.3.6.1 Bases pg B 3.3-164 1 of 1
JAMES A. FITZPATRICK NPP IMPROVED TECHNICAL SPECIFICATIONS REVISION "K" INSERT AND DISCARD INSTRUCTIONS REMOVE I INSERT VOLUME 16 NUREG Bases markup for ITS 3.8.6 pg NUREG Bases markup for ITS 3.8.6 pg B 3.8-66 B 3.8-66 Bases JFD for ITS 3.8.6 pg 1 of 3 Bases JFD for ITS 3.8.6 pg 1 of 3 Retyped ITS 3.8.7 Bases pg B 3.8-60 Retyped ITS 3.8.7 Bases pg B 3.8-60 1 of 1
ATTACHMENT 1
SUMMARY
OF CHANGES TO ITS SECTION 3.0 - REVISION K Summary of Change Affected Pages Source of Change These editorial changes were identified during the Section 3.0 Misc. editorial corrections preparation of the final ITS submittal:
Revision J inadvertently failed to add a 'bubble' CTS mark-up, p 3 of 5 reference to DOC L4 or revise CTS 4,0.C on CTS mark up page 30a (p 3 of 5).
Or 1
('- VR..,W ,
)niue-.d~ ~e e $ ve*~
rs
-4 t a Surveillance Requirement has not been performed.
-,* ou=
____________y__o__up_[
- GINrequt Ently into an OPERATIONAL CONDITION (mode) or other ha1IibIc
. Toermit the completion of the survellIancek.Wh*R~l shall not be madenot when the conditions for the
-, specified condition for Operation are met and the associated LtO 3.01 Urmiting Condition shutdown if they are not met within a specified "ACTIONrequires a se jjine interval. Entry Into an OPERATIONAL CONDITION (mode miay be made Inaccordance with ACTION or specified conditinconformance requirements when to them permits cntinue D. Entry made into an OPERATIONAL unless CONUI I lUN itinuu-ja, -....
the Surveillance Requirement(s) with the period of time. This thin the o-eration of the facility for an unlimited Limiting Condition for Operation have bee OPERATIONAL otherwise specified. This
-*,sIon shah not prevent passage through ACTION applicable surveillance interval or as CONDITiONS (modes) required to comply with through or to Operational of a shutdown of the plant. provision shall not prevent passagewith ACTION.rulrements or that requirements or that are part "(3 Modes as required to comply are stated in the individual are part of a shutdown of the plant.
Sýxce~ptlons to these requirements ce esing o comoponen s because Its emergencypo shall be applicable as follows:
determined to be inoperable solely Its normal power sourc valves shall be performed in source is inoperable, or solely because Inservice testing of pumps andthe ASME Boiler and OPERABLE for the puroeo accordance with Section Xl of i 4noperable. Itmay be considered Uimiting Condito for Addenda as required satisfying the requirements of its applicable Pressure Vessel Code and applicable where specific SAC or emergency by 10 CFR 50, Section 50.55a(f), except ps~*.g operation, provided: (1)its corresponding normal of Its redundant granted by the NRC pursuant to 10 power source is OPERABLE; and (2)all and device(s) are written relief has been The inservica testing andd "sysstem(s).subsystem(s), train(s), component(s) CFR 50, Section 50.55a(f)(6)(i). an NRC approved edition o of this Is based on OPERABLE, or likewise satisfy the requirements and (2) are satisfied, the )inspection program specification. Unless both conditions (1) ASME Boiler and following 24 and addenda to,Section XI of the within unit shall be placed in COLD SHUTDOWNwhen inCold Shutd the Code which Is In effect 112 mmonths prio t not applicable SPressure Vessel hours. This specification Is o th inpecioninterval..
- thebegnnig declared inoperable to Equipment removed from service or
- , 4- may be retured to service under comply with required actions required to
@t0 administrative control solely to perform testingof other eauii ment. *1 demonstrate its operability or the operability (M5*.- 3*75,--
This is an exception to LC 190. 2,', 41, Amendment No. 03. I. 30a 4-REVISIO V5 3.-
REVISION g" k-
ý-(AQO iO LVCAI-fa
SUMMARY
OF CHANGES TO ITS SECTION 3.3 - REVISION K Summary of Change Affected Pages Source of Change TS Amendment 273 Incorporates TS Amendment 273 into ITS. TSA 273 Section 3.3.4.1 revised the ATWS RPT instrument setpoints for Reactor 4 of Pressure - High to a single setpoint, independent of the CTS mark-up, pp 2 of 6 and number.of SRVs that are operable. 6 DOC A9 - deleted (p 3 of 8); DOC M2 (p 4 of 8)
ITS mark-up, p 3.3-35 ITS Bases mark-up, pp Insert page B 3.3-94, B 3.3-96, B 3.3 98, B 3.3-100 and Insert page B 3.3-100 Retyped ITS p 3.3-31 Retyped ITS Bases pp B 3.3-90, B 3.3-92, B 3.3-94, B 3.3-95 and B 3.3-96 Page 1 of 3
SUMMARY
OF CHANGES TO ITS SECTION 3.3 - REVISION K Summary of Change Affected Pages Source of Change NRC telecon The Modes of Applicability for ITS 3.3.6.1, Functions 1.f Section 3.3.6.1 and 2.f (Main Steam Line Radiation - High) is revised to be consistent with CTS. CTS mark-up, p 3 of 22 DOC LI 3 - deleted (p 22 of 25)
NSHC L13 - deleted (pp 19 of 32 and 20 of 32)
ITS mark-up, pp 3.3-57 and Insert page 3.3-58 JFD DB6 (p 4 of 5)
ITS Bases mark-up, Insert page B 3.3-158, Insert page B 3.3-161 and Insert page B 3.3-164b Retyped ITS pp 3.3-52 and 3.3 53 Retyped ITS Bases pp B 3.3-160 and B 3.3-164 Page 2 of 3
SUMMARY
OF CHANGES TO ITS SECTION 3.3 - REVISION K Summary of Change Affected Pages Source of Change Provides a Bases cross reference from ITS SR 3.3.1.1.12 to Section 3.3.1.1 NRC telecon ITS SR 3.3.2.1.8 regarding calibration of the recirculation loop flow signal portion of the channel. An additional CHANNEL NUREG Bases mark-up, p Insert CALIBRATION surveillance (ITS SR 3.3.2.1.8) is added as page B 3.3-30 well as NOTES to SR 3.3.2.1.5 and SR 3.3.2.1.8 regarding calibration of the recirculation loop flow signal portion of the Bases JFD CLB5 (p 1 of 4) channel.
Retyped ITS Bases p 3.3-33 Section 3.3.2.1 CTS mark-up, p 6 of 10 and 8 of 10 DOCs A7 (p 2 of 9) and L5 (p 7 of 9)
NUREG mark-up pp 3.3-19 and 3.3 20 JFDs CLB1 and DB4 (pp 1 of 3 and 2 of 3)
NUREG Bases mark-up, B 3.3-54, Insert page B 3.3-54 Retyped ITS pp 3.3-18, 3.3-19 and 3.3-20 Retyped ITS Bases pp B 3.3-56 through B 3.3-59 Misc. editorial These editorial changes were identified during the Section 3.3.6.1 corrections preparation of the final ITS submittal: DOC M4 (p 7 of 25)
- 1. ITS 3.3.6.1 DOC M4 (p 7 of 25) refers to Note 2.
There is only one Note; therefore, the DOC has been Section 3.3.1.1 corrected. DOC M8 (p 9 of 25)
- 2. ITS 3.3.1.1 DOC M8 (p 9 of 25) refers to Note 3.
This should refer to Note 2; therefore, the DOC has been corrected.
Page 3 of 3
DISCUSSION OF CHANGES ITS: 3.3.1.1 - REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M5 (continued) to the current The addition of new requirements (Surveillances) change necessary Technical Specifications constitutes a more restrictive maintained Operable. This change is to ensure the RPS Functions are is not considered consistent with NUREG-1433, Revision 1. This change to result in any reduction to safety.
the Channel M6 ITS SR 3.3.1.1.1, increases the frequency for performingto every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Checks in CTS Table 4.1-1 from the current Daily the Functions listed below:
Reactor Pressure- High Drywell Pressure- High Reactor Vessel Water Level -Low (Level 3)
Scram Discharge Volume Water Level -High (DP transmitter/trip unit)
Turbine First Stage Pressure Permissive (see LA12) of the current Technical This change to the requirements (Surveillances)change necessary to ensure Specifications constitutes a more restrictive change is consistent the RPS Functions are maintained Operable. is Thisnot considered to result in with NUREG-1433, Revision 1. This change any reduction to safety.
channels overlap prior M7 ITS SR 3.3.1.1.5 was added to verify SRM and IRMrequirements to fully withdrawing SRMs. This change to the constitutes a (Surveillances) of the current Technical Specifications RPS Functions are more restrictive change necessary to ensure the maintained Operable.
reactor protection M8 CTS 4.1.A specifies that the response time of the to be within its system trip functions listed shall be demonstrated limit once per 24 months. Each test shall trip include at least one channel systems shall be tested in each trip system. All channels in both RPS RESPONSE TIME within two test intervals. In ITS SR 3.3.1.1.15 the TEST BASIS.
test must be performed every 24 months on a STAGGERED for the purpose Note 2 of this SR specifies that "n" equals 2 channels Therefore, of determining the STAGGERED TEST BASIS Frequency. response time testing SR 3.3.1.1.15 will require all channels requiring This change is more to be tested in two (2) surveillance intervals. Function 5 (Main Steam restrictive since at least eight (8) ITS 3.3.1.1 ITS 3.3.1.1 Function 8 Isolation Valve-Closure) channels and four (4) tested each interval (Turbine Stop Valve-Clos5jre) channels must be Revision K JAFNPP Page 9 of 25
INSERT SR 3.3.1.1.10 For Functions 8 and 9, this SR is associated with the enabling circuit sensing first stage turbine pressure.
_ *INSERT SR 3.3.1.1.12-1 Physical inspection of the position switches is performed in conjunction with SR 3.3.1.1.12 for Function 5 and 8 to ensure that the switches are not corroded or otherwise degraded. For Function 7.b, the CHANNEL CALIBRATION must be performed utilizing a water column or similar device to provide will assurance that damage to a float or other portions of the float assembly the be detected. For Functions 8 and 9, SR 3.3.1.1.12 is associated with enabling circuit sensing first stage turbine pressure as well as the trip function.
C INSERT SR 3.3.1.1.12-2 Note 3 to SR 3.3.1.1.9 and the Note to SR 3.3.1.1.12 concerns the Neutron Flux-High (Flow Biased) Function (Function 2). Note 3 to SR 3.3.1.1.9 excludes the recirculation loop flow signal portion of the channel, the since this by SR 3.3.1.1.12. Similarly, Note to portion of the channel is calibrated SR 3.3.1.1.12 excludes all portions of the channel except the recirculation loop flow signal portion, since they are covered by SR 3.3.1.1.9. Since the recirculation loop flow signal is also a portion of the Rod Block Monitor (RBM) - Upscale control rod block Function channels (Table 3.3.2.1-1, Control Rod Block Instrumentation, Function 1.a), satisfactory performance of for the SR 3.3.1.1.12 also results in satisfactory performance of SR 3.3.2.1.8 associated RBM-Upscale control rod block Function channels.
Reactor Pressure-High and Reactor Vessel Water Level -Low (Level 3) Function sensors (Functions 3 and 4, respectively) are excluded from the RPS RESPONSE TIME testing (Ref. 19). However, prior to the CHANNEL CALIBRATION of these sensors a response check must be performed to ensure adequate response. This testing is required by Reference 20. Personnel involved in this testing must have been trained in response to Reference 21 to ensure they are aware of the consequences of instrument response time degradation. This response check must be performed by placing a fast ramp or a step change into the input of the each required sensor. The personnel, must monitor the input and output of associated sensor so that simultaneous monitoring and verification may be accomplished.
9 INSERT SR 3.3.1.1.9 The Frequency of SR 3.3.1.1.9 is based upon the assumption of a 92 day calibration interval in the determination of the magnitude of equipment drift in the setpoint analysis.
Insert Page B 3.3-30 Revision K
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.3.1.1 - REACTOR PROTECTION SYSTEM (RPS) INSTRUMENTATION RETENTION OF EXISTING REQUIREMENT (CLB)
CLB1 Function 2.d has been deleted. The Downscale trip has been removed from following Functions the CTS as documented in License Amendment 227. The have been renumbered as required.
test of each RPS automatic CLB2 SR 3.3.1.1.4 has been added (a functional requirements. This scram contactor) consistent with current Frequency extensions of Surveillance was added to allow the Surveillance Technical Specification the automatic RPS Functions per NEDC-30851-P-A. System, since the JAFNPP Improvement Analyses for BWR Reactor Protection used in NEDC-30851-P-A.
design is different than the generic BWR model of the CHANNEL Therefore, the Bases description in ISTS SR 3.3.1.1.5 has been deleted and FUNCTIONAL TEST of the manual scram function test switches.
replaced with the description of the RPS channel the sensor during response CLB3 Consistent with CTS 4.1.A. the measurement of as references time testing is not required. Appropriate Ri. as well Bases have been included consistent with TSTF 322 to require RPS CLB4 The Bases of ITS SR 3.3.1.1.15 has beenthemodified, current licensing basis, and RESPONSE TIME TESTING consistent with as modified in MB.
CLB5 ISTS SR 3.3.1.1.3. the requirement to beenadjust the channels to conform to a calibrated signal every 7 days has deleted since this requirement is currently being performed along with the 92 day channel functional test. This adjustment will be performed in accordance with SR 3.3.1.1.8, the 92 day CHANNEL FUNCTIONAL TEST. This is reflected in the as Bases of SR 3.3.1.1.8. Subsequent SRs have been renumbered,signal portion of applicable. In addition, the recirculation loop flowNotes have been Function 2.b is calibrated by SR 3.3.1.1.12. Thus, and since the added to SR 3.3.1.1.9 and SR 3.3.1.1.12 for clarity the RBM - Upscale recirculation loop flow signal is also a portion of 3.3.2.1.8 has control rod block Function channels. a reference to ITS SR been added to the ITS SR 3.3.1.1.12 Bases.
CLB6 These requirements have been added in accordance with CTS Table 4.1-1 Note 6 and Table 4.1-2 Note 5. as documented in LA11.
has been CLB7 The Channel Functional Test Frequency of SR 3.3.1.1.11 CTS Table increased from 18 months to 24 months in accordance with cycle.
4.1-1. The Frequency is consistent with the JAFNPP fuel CLBB SR 3.3.1.1.10 Surveillance Frequency has been modified to be consistent in License with the frequency in CTS Table 4.1-1 Note 6 and approved Amendment No. 89.
Page 1 of 4 Revision K JAFNPP
REVISION 1 JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, (RPS) INSTRUMENTATION ITS BASES: 3.3.1.1 - REACTOR PROTECTION SYSTEM RETENTION OF EXISTING REQUIREMENT (CLB) have been added to the CLB9 The specific details concerning response checks Amendment No. 235.
Bases of SR 3.3.1.1.12 in accordance with License IMPROVEMENT (PA)
PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL plant specific PAl The Specification has been modified to reflect nomenclature.
clarity or to be PA2 Editorial changes have been made for enhanced consistent with other places in the Bases.
PA3 Grammatical or typographical error corrected.
generic and not plant PA4 This Table has been deleted because it provides in the Table could be specific types of information. The information take credit for these misleading as to which plant specific analyses and transient scenarios.
channels to perform a function during accident PA5 The Reviewer's Note has been deleted.
have been removed. The PA6 The quotations used in the Bases References Writer's Guide does not require the use of quotations.
reflect the PA7 The Bases description has be modified to better Applicability of the Functions in Table 3.3.1.1-1.
PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)
JAFNPP specific design.
DB1 The Bases have been modified to reflect the plant specific reference DB2 The brackets have been removed and the proper have been provided.
Power Range Monitor Flow DB3 The Bases description of Function 2.b, Average has been modified to be Biased Simulated Thermal Power-High Function circuit has been removed consistent with the JAFNPP design. The filter Stability Solutions (Refs. 5 consistent with BWR Owner's Group Long Term as a result of this design and 6). Changes have been made in the Bases as applicable. In difference. References have been renumbered, because the JAFNPP RPS does addition. ISTS 3.3.1.1.14 has been deleted Thermal Power-High time not utilize an APRM Flow Biased Simulated as applicable.
constant. Subsequent SRs have been renumbered, S.... nf A. Revision K JAFNPP rayc %Iu
1 JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION INSTRUMENTATION ITS BASES: 3.3.1.1 - REACTOR PROTECTION SYSTEM (RPS)
PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)
All channels are not required to respond within a specified response DB4 Allowable Values (e.g.
time and all channels do not have a specified the Bases has been revised as Manual Scram Function channels), therefore necessary.
methodology has been revised DB5 The description of the setpoint calculation to reflect the plant specific methodology.
references.
DB6 The Bases has been revised to reflect the appropriate analysis. At low DB7 The Bases has been revised to reflect the safety TSV and TCV is not required; powers (e.g.. < 29% RTP) the scram from theonline (and trip with however, the turbine generator can remain power level. The TSV and TCV resultant pressure transient) below this trip) provide a direct Fast Closure (turbine trip or main generator RTP, a turbine or main reactor scram when k 29% RTP. When < 29X scram, but should the generator trip will not result in a direct the Reactor High Pressure pressure transient reach the setpoint for occur from the Reactor trip, a scram would occur (i.e., is credited tobelow 29% RTP includes High Pressure trip). Since turbine operation of the Reactor High MODE 1 and MODE 2, the necessary applicability 1 and 2. References Pressure trip is consistent with specifying MODE references have been have been included as applicable. Subsequent renumbered as required.
calculation DB8 The Bases has been revised to reflect the setpoint methodology assumptions.
CALIBRATION every DB9 SR 3.3.1.1.9 has been added to perform a CHANNEL Water Level -High, 92 days for Function 7.a (Scram Discharge Volume with CTS Table Differential Pressure Transmitter/Trip Unit) consistent calculation setpoint 4.1-2. The Frequency is consistent with the the Frequency for ISTS SR methodology for this Function. In addition, requirement for the APRM 3.3.1.1.11, the 184 day CHANNEL CALIBRATION 3.3.1.1.9), consistent Functions, has been changed to 92 days (ITS SR reordered and renumbered with the CTS. The Bases description has been as required.
changes made to the DB1O Changes have been made to reflect those Specification.
ray^ : -Al Revision K JAFNPP ravc Um
1 JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION INSTRUMENTATION ITS BASES: 3.3.1.1 - REACTOR PROTECTION SYSTEM (RPS)
DIFFERENCE BASED ON AN APPROVED TRAVELER (TA)
The changes presented in Technical Specification332, Task Force (TSTF)
TA1 Number Revision 1 have been Technical Specification Change Traveler Specifications.
incorporated into the revised Improved Technical been adopted by JAFNPP.
1 has not yet However. NEDO-32291-A, Supplement been incorporated.
Therefore, this portion of the TSTF has not Task Force (TSTF)
TA2 The changes presented in Technical Specification205, Revision 3 have been Technical Specification Change Traveler Number Specifications.
incorporated into the revised Improved Technical Task Force (TSTF)
TA3 The changes presented in Technical Specification231, Revision 1 have been Technical Specification Change Traveler Number Specifications.
incorporated into the revised Improved Technical Task Force (TSTF)
TA4 The changes presented in Technical Specification355. Revision 0, as Technical Specification Change Traveler Number into the revised Improved modified by WOG-ED-25, have been incorporated Technical Specifications.
DIFFERENCE BASED ON A SUBMITTED, BUT PENDING TRAVELER (TP)
None (X)
DIFFERENCE FOR ANY REASON OTHER THAN THE ABOVE NRC Policy Statement" X1 NUREG-1433, Revision 1, Bases reference to "the in accordance has been replaced with 10 CFR 50.36(c)(2)(ii), Subsequent References have with 60 FR 36953 effective August 18, 1995.
been renumbered, as applicable.
have been modified from X2 The SR 3.3.1.1.13 and SR 3.3.1.1.14 Frequencies fuel cycle.
18 months to 24 months consistent with the JAFNPP
^A f A Revision K JAFNPP rayc uV
RPS Instrumentation B 3.3.1.1 BASES SR 3.3.1.1.9 and SR 3.3.1.1.12 (continued)
SURVEILLANCE REQUIREMENTS by SR 3.3.1.1.12.
this portion of the channel is calibrated excludes all portions Similarly, the Note to SR 3.3.1.1.12 loop flow signal of the channel except the recirculation SR 3.3.1.1.9. Since the portion, since they are covered by also a portion of the Rod recirculation loop flow signal is block Function Block Monitor (RBM) -Upscale controlRodrodBlock channels (Table 3.3.2.1-1, Control performance of Instrumentation, Function l.a), satisfactoryperformance of SR 3.3.1.1.12 also results in satisfactory control rod SR 3.3.2.1.8 for the associated RBM-Upscale block Function channels.
Water Level -Low Reactor Pressure-High and Reactor Vessel 3 and 4. respectively)
(Level 3) Function sensors (Functions TIME testing (Ref. 19).
are excluded from the RPS RESPONSE of these sensors a However, prior to the CHANNEL CALIBRATION adequate response check must be performed to ensure by Reference 20.
response. This testing is required have been trained in Personnel involved in this testing must are aware of the response to Reference 21 to ensure they time degradation. This consequences of instrument response placing a fast ramp or a response check must be performed by required sensor. The step change into the input of each and output of the personnel, must monitor the input monitoring and associated sensor so that simultaneous verification may be accomplished.
on the assumption of The Frequency of SR 3.3.1.1.9 is baseddetermination of the a 92 day calibration interval in the the magnitude of equipment drift in setpoint analysis.
upon the assumption The Frequency of SR 3.3.1.1.12 is based in the determination of of a 24 month calibration intervalin the setpoint analysis.
the magnitude of equipment drift SR 3.3.1.1.10 of the actual Calibration of trip units provides a check inoperable if trip setpoints. The channel must be declared conservative than the trip setting is discovered to be less If the the Allowable Value specified in Table conservative 3.3.1.1-1.
discovered to be less than trip setting is methodology, but accounted for in the appropriate setpoint the channel performance is not beyond the Allowable Value, (continued)
Revision K D J) o" J' JAFNPP
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.10 (continued)
REQUIREMENTS is still within the requirements of the plant safety setpoint must be analysis. Under these conditions, the than readjusted to be equal to or more conservative methodology. For accounted for in the appropriate setpoint with the enabling Functions 8 and 9, this SR is associated circuit sensing first stage turbine pressure.
the reliability.
The Frequency of 184 days is based ofon the solid-state accuracy, and lower failure rates System components.
electronic Analog Transmitter/Trip SR 3.3.1.1.13 the The LOGIC SYSTEM FUNCTIONAL TEST demonstrates for a specific OPERABILITY of the required trip logic rods channel. The functional testing of control valves (LCO 3.1.8),
(LCO 3.1.3). and SDV vent and drain to provide complete testing of overlaps this Surveillance the assumed safety function.
need to perform this The 24 month Frequency is based on the apply during a plant Surveillance under the conditions that transient if the outage and the potential for an unplannedreactor at power.
Surveillance were performed with the these components usually Operating experience has shown that at the 24 month pass the Surveillance when performed Frequency.
SR 3.3.1.1.14 from the Turbine Stop This SR ensures that scrams initiated Fast Closure, EHC Valve-Closure and Turbine Control Valve inadvertently Oil Pressure-Low Functions will not be This involves bypassed when THERMAL POWER is k 29% RTP.Adequate margins for calibration of the bypass channels. are incorporated into the instrument setpoint methodologies bypass flow can the actual setpoint. Because main turbine (THERMAL POWER is affect this setpoint nonconservatively the main turbine derived from turbine first stage pressure), an inservice bypass valves must remain closed during to ensure that the calibration at THERMAL POWER k 29% RTP calibration is valid.
(continued)
B 3.3-34 Revision K JAFNPP
RPS Instrumentation B 3.3.1.1
-, BASES SR 3.3.1.1.14 (continued)
SURVEILLANCE REQUIREMENTS (i.e.,
If any bypass channel's setpoint is nonconservative either due to open the Functions are bypassed at k 29% RTP.
reasons), then the main turbine bypass valve(s) or other and Turbine Control affected Turbine Stop Valve-Closure Functions are Valve Fast Closure, EHC Oil Pressure-Low considered inoperable. Alternatively, the bypass channel (nonbypass). If can be placed in the conservative condition SR is met and the condition, this placed in the nonbypassOPERABLE.
channel is considered The Frequency of 24 months is based on engineering judgment and reliability of the components.
SR 3.3.1.1.15 response times This SR ensures that the individual channel assumed in the are less than or equal to the maximum values accident analysis. The RPS RESPONSE TIME acceptance criteria are included in Reference 22.
response time RPS RESPONSE TIME may be verified by actualoverlapping, or measurements in any series of sequential, sensors for total channel measurements. However, the RPS RESPONSE Functions 3 and 4 are excluded from specific conditions of Reference 19 are TIME measurement since the response time may satisfied. For Functions 3 and 4, sensor sensor response be allocated based on either assumed design response time. For time or the manufacturer's stated design must be measured.
all other Functions, sensor response time RESPONSE TIME Note 1 excludes neutron detectors from RPS operation testing because the principles of detector time.
virtually ensure an instantaneous response a 24 month RPS RESPONSE TIME tests are conducted onSTAGGERED TEST BASIS STAGGERED TEST BASIS. Note 2 requires This Frequency to be determined based on 2 channels. during two ensures all required channels are testedFunctions 2.b, 2.c, Surveillance Frequency intervals. For during each 3, 4. 6. and 9, two channels must be tested and four channels test; while for Functions 5 and 8. eight (continued)
D J- '
Revision K JAFNPP
RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.15 (continued)
REQUIREMENTS the logic must be tested. This Frequency is based on required to interrelationships of the various channels Frequency is produce an RPS scram signal. The 24 month based upon plant is consistent with the refueling cycle and random failures of operating experience, which shows that response time instrumentation components causing serious infrequent degradation, but not channel failure, are occurrences.
REFERENCES 1. UFSAR, Section-7.2.
- 2. UFSAR, Section 14.5.4.2.
- 3. NEDO-23842. Continuous Control Rod Withdrawal Transient In The Startup Range, April 18, 1978.
- 5. NEDO-31960-A, BWR Owners' Group Long Term Stability Solutions Licensing Methodology, June 1991.
Long
- 6. NEDO-31960-A, Supplement 1. BWR Owners' Group Term Stability Solutions Licensing Methodology.
Supplement 1, March 1992.
- 7. UFSAR, Section 14.5.1.2.
- 8. UFSAR, Section 14.6.1.2.
- 9. UFSAR. Section 14.5.2.1.
- 10. UFSAR, Section 14.5.2.2.
- 11. UFSAR, Section 6.3.
- 12. Drawing 11825-5.01-15D, Rev. D, Reactor Assembly Nuclear Boiler, (GE Drawing 919D690BD).
- 13. UFSAR, Section 14.5.5.1.
- 14. UFSAR, Section 14.5.2.3.
- 15. UFSAR, Section 14.6.1.5.
(continued)
SRevision K 0 -1°**
JAFNPP .- .
RPS Instrumentation B 3.3.1.1 BASES REFERENCES 16. P. Check (NRC) letter to G. Lainas (NRC), BWR Scram (continued) Discharge System Safety Evaluation, December 1, 1980.
- 17. UFSAR, Section 14.5.9.
- 18. NEDC-30851P-A, Technical Specification Improvement Analyses for BWR Reactor Protection System, March 1988.
- 19. NEDO-32291-A System Analyses For the Elimination of Selected Response Time Testing Requirements, October 1995.
- 20. NRC letter dated October 28, 1996, Issuance of Amendment 235 to Facility Operating License DPR-59 for James A. FitzPatrick Nuclear Power Plant.
in NRC Bulletin 90-01, Supplement 1. Loss of Fill-Oil1992.
- 21. December Transmitters Manufactured by Rosemount,
- 22. UFSAR, Table 7.2-5.
Revision K JAFNPP B 3.3-37
/
WP3, 3, 23I-1 C CONTROL ROD BLOCK INSTRUMENTATION Instrument F Instrument Test Calibration Check (Note 4)
Instrument Channel
- 2) APRM - U-scale
/~ 64) 4)
51_
RM - Dopscale iRM - Detectore IRM Upscale IRM - Detector ot inn S~tanupp PPosition
/t SS/U Note 21 S/U (Note 2) / 0 Motes 3 & 6) O mo4s&6
/ NA...Noe_2 I +HbA
-Upscale AR BM 7a)( 26-) R/BM - pcl sR3,3,2,1.1 NA NA ,
- 9) SRM - Detecto Not In tup Position SU Note 2 a0 O 5; 3, 3.
- 10) Scram Discharge Instr ant Volume -
0L High Water e.e.
(
W3z, 233 Amendment No. 3.80, 82 Uv 5
- SAýd'O"' K hho"o
5-ec' JAPH I.-,-
'no
[ APPI; CCIL ('ý I 4.3.B (01 1-'jo 3.3.6 (ContlI 3.ThetcapabfIty oft tdhe Rod Yftlh lrrilizzu' to properly UNdflf Is balm 10% rated rdý,
MM Vk*Wm Rod V*Nlh WýMt:ýýýM ra b eftwceWO o pIsýIba as follows: a. During startup. prior to the start of conrola r 1 a. . hu!teRMb0f cA1 nPr A URA ~~. ~. reactor strtu afther first twelve control rods hwer 33.-193 15 (1) lbe correctness o the R#M programi Z.been ~00ck. or during a reactor shuldmn control rod wmismut may q~fuWPrO!'Id #Wha
~o) second terw 4 .......... W7 (2)ThRWMM orlonofne di lO Utest 0
oper ffr lyrdl kwdpoWjo* LAG shal be zd"x wuiuncla of h seection ai~fM SoordW= With Vieum 1OfC 4.e, one out-oIOUf-l~ece nec ktInsertedgroupshallbe sled.
x CA,1 b. SOid the FMIM bek" "bfrado
~~ begun or beomern kvopeimW goe ~wlcWd withwickaw, stertup, may ccilnu Inaccodenc with Specili~callo 3.32.U. above.
(1) The correcirnes of the RIVM progren sequenice shall be WerNWe. I&2 0,.21 911
,be ON comiplir PIA Ine dlagnostic'"ds (21 Amndmxkent No. A , 155 92 q4 4ta0 /0 K
Palfr~
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION ADMINISTRATIVE CHANGES Nuclear Power Plant Al In the conversion of the James A. FitzPatrick (CTS) to the proposed plant (JAFNPP) Current Technical Specifications (ITS) certain wording specific Improved Technical Specifications do not result in technical that preferences or conventions are adopted and revised numbering are changes. Editorial changes, reformatting, the conventions in NUREG-1433, adopted to make the ITS consistent with Electric Plants, BWR/4",
"Standard Technical Specifications, General Specifications (ISTS)).
Revision 1 (i.e., Improved Standard Technical 4orth Minimizer (RWM) have been added to CTS A2 The requirements of the Rod Fable 3.3.2.1-1 Function 2). This addition Tables 3.2-3 and 4.2-3 (ITS since the requirement concerning RWM is considered administrative CTS 3.3.B.3. This change is consistent OPERABILITY are contained in with NUREG-1433, Revision 1.
A3 Not Used.
functional test and A4 CTS Table 4.2-3 requires both an instrument basis for both the RBM calibration to be performed on a quarterlyRBM-Downscale (Function 7)
Upscale (CTS Table 4.2-3 Function 6) and the performance of a Functions. In the ITS, SR 3.3.2.1.5 requires CHANNEL CALIBRATION. It is not necessary to specify a CHANNEL includes of CHANNEL CALIBRATION FUNCTIONAL TEST since the ITS definition TEST. Therefore, the all the requirements of a CHANNEL FUNCTIONAL included in the ITS. This explicit instrument functional test is not the CHANNEL CALIBRATION is change is considered administrative since all the requirements of a performed on a quarterly basis and fulfillschange, Table 4.2-1 through CHANNEL FUNCTIONAL TEST. Along with this channel function test (This 4.2-5 Note 5 which is associated with the the CTS since the CHANNEL instrument is exempt...) is deleted from The details of this FUNCTIONAL TEST is not required to be performed.
of CHANNEL FUNCTIONAL, therefore Note are included in the ITS definition its removal is also considered administrative.
that instrument checks are A5 CTS Table 4.2-1 through 4.2.5 Note 4 statesnot required to be operable or not required when these instruments are is not retained in ITS 3.3.2.1.
are tripped. This explicit requirement 3.3.2.1 since these allowances This explicit Note is not needed in ITS states that SRs shall be met are included in ITS SR 3.0.1. SR 3.0.1 in the Applicability for during the MODES or other specified conditions in the SR. In addition, the individual LCOs. unless otherwise stated have to be performed on inoperable Note states that Surveillances do not limits. When equipment is equipment or variables outside specified LCO require the equipment to be declared inoperable, the Actions of thiscondition, the equipment is still placed in the trip condition. In this Revision F JAFNPP rage 1 uo
DISCUSSION OF CHANGES ITS: 3.3.2.1 CONTROL ROD BLOCK INSTRUMENTATION ADMINISTRATIVE CHANGES A5 (continued) inoperable but has accomplished the required safety function:' Therefore adequate the allowances in SR 3.0.1 and the associated actions provide are required guidance with respect to when the associated surveillances retained.
to be performed and this explicit requirement is not CTS 3.2.D and 4.2.D provide a cross reference to the Radiological A6 those Radiation Effluent Technical Specification (Appendix B) for and Initiation Function.
Monitoring Systems which provide an Isolation specific requirements and Since CTS 3.2.D and 4.2.D do not prescribe any in Appendix B are since the changes to the current requirements this submittal, this cross discussed in the Discussion of Changes within administrative reference has been deleted. This change is consideredThis change is consistent since it simply eliminates a cross-reference.
with NUREG-1433, Revision 1.
The proposed change adds ITS SR 3.3.2.1.8 for CHANNEL CALIBRATION of the A7 the RBM - Upscale Function recirculation loop flow signal portion of not contain a specific (which is flow biased). CTS Table 4.2-3 does of the recirculation loop flow surveillance requirement for calibration to the signal. The recirculation loop flow signal provided provided to RBM-Upscale control rod block Function is the same signal (ITS Table 3.3.1.1-1, the APRM Neutron Flux-High (Flow Biased) Function (5), and CTS Table 4.1-2. Item (4)).
Function 2.b, CTS Table 3.3-1, Item to provide The proposed change also adds Note 2 to ITS SR 3.3.2.1.5 by ITS SR 3.3.2.1.5 clarification that the CHANNEL CALIBRATION required flow signal portion of the channel for excludes the recirculation loop excludes all Function l.a while the Note in proposed ITS SR 3.3.2.1.8 signal.
portions of the channel except the recirculation loop flow it is Since this change does not change any current requirements, considered administrative.
TECHNICAL CHANGES - MORE RESTRICTIVE An additional Function has been added to CTS Table(Rod 3.2-3 for the Rod M1 Block Monitor Block Monitor. ITS Table 3.3.2.1-1 Function l.b consistent with Inop) will require the "Inop" function to be Operable This the Applicability with the other Rod Block Monitor Functions. block is change is more restrictive but necessary to ensure anotrodavailable to the provided if the minimum number of LPRMs inputs arefunctional test (i.e.,
associated Rod Block Monitor channel. A channel Monitor Inop function.
SR 3.3.2.1.1) is also proposed for the Rod Block that the The performance of this SR for each RBM channel will ensure Page 2 of 9 Revision K JAFNPP
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES MORE RESTRICTIVE Ml (continued) when it is required to entire channel will perform its intended function of 92 days for SR be Operable. The proposed surveillance frequency provided in NEDC-30851-P 3.3.2.1.1 is based on the reliability analysis concluding that this topical A (see revised DOC L3 for the bases for JAFNPP). Accordingly, the addition report is acceptable for use at the its associated channel of the Rod Block Monitor - Inop function, interval will help to functional test SR and the 92 day surveillance during control rod ensure that the local flux is adequately monitored operator the inoperability of withdrawal by promptly identifying to the component failures.
the Rod Block Monitor as a consequence of certain CTS Table 3.2-3. ITS 3.3.2.1, An additional Function has been added to include M2 the Control Rod Block Control Rod Block Instrumentation, will a required function (Function 3 Function of the Reactor Mode Switch as is that 2 channels of on proposed Table 3.3.2.1-1). The new requirement Position must be the Rod Block function of Reactor Mode Switch-Shutdown position. This Operable whenever the Mode Switch is in the Shutdown Rod Block Instrumentation addition to the Specification for the ControlFUNCTIONAL TEST every 24 will include proposed SR 3.3.2.1.7 (CHANNEL E (Required Actions and months) and proposed LCO 3.3.2.1, Condition ITS SR 3.3.2.1.7 will Completion Times if this function is inoperable).
after the Reactor Mode not be required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> ensures that control rods Switch is placed in Shutdown. This rod block rods are assumed to be are not withdrawn in MODES 3 and 4, since control Revision 1.
inserted. This change is consistent with NUREG-1433, The out of service time in CTS Table 3.2-3 Note 2 Action B.a) has been M3 Required Action A.1) when reduced from 7 days to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (ITS 3.3.2.1 Time is one RBM channel is inoperable. The 24 houranCompletion occurring coincident acceptable, based on a low probability of Thisevent change is more with a failure in the remaining channel. but consistent with NUREG-1433, restrictive since less time .is permitted Revision 1.
SR 3.3.2.1.4 has been added to CTS Table 4.2.3 to verify that the RBM is M4 when a peripheral control not bypassed at Thermal Power x 30% RTP andchange is more restrictive rod is not selected every 92 days. Thisincluded. This will ensure the since a periodic surveillance has been consequences of a single RBM is Operable when required to limit the power operation.
control rod withdrawal error event during 2'f a Revision K JAFNPP rayc Vi
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE surveillance is M5 A new CHANNEL FUNCTIONAL TEST (ITS SR 3.3.2.1.3) in MODE 1 when Thermal proposed to be added similar to CTS 4.3.B.3.a.4 the reactor mode Power is : l0% to ensure the RWM is Operable92 with days and is consistent switch in RUN. The test is required every with NEDC-30851-P-A, "Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation," October 1988.
of CTS 4.3.B.3.
M6 A new SR is proposed to be added to the surveillances Rod Worth Minimizer SR 3.3.2.1.6 will verify every 24 months that: the The RWM may be (RWM) is not bypassed when Thermal Power is the10%. existing specifications bypassed when power is above 10%. However, to verify the setpoint (CTS 4.3.B.3) do not have an explicit requirement an additional of the RWM bypass feature. This change represents to ensure the RWM Function is restriction on plant operations necessary Operable when required.
TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC) of Instrument Channels LA1 The specific details in the "Total Number3.2-3 are proposed to be Provided By Design" column of CTS Table details in the Bases provides relocated to the Bases. Placing these requirements of ITS 3.3.2.1 assurance they will be maintained. The to be OPERABLE, the which require the Control Rod Block Instrumentation Required Action and definition of OPERABILITY, and the proposed are not required to be in surveillances suffice. As such, these details health and safety.
the ITS to provide adequate protection ofbypublic the provisions of the Bases Changes to the Bases will be controlled the ITS.
Control Program described in Chapter 5 of CTS 4.3.B.3.b.2 are proposed LA2 The requirements of CTS 4.3.B.3.a.2. 3 and Manual. The RWM computer to be relocated to the Technical Requirements and CTS 4.3.B.3.b.2 and the on line diagnostic test in CTS 4.3.B.3.a.2 in CTS 4.3.B.3.a.3 are not proper annunciation of the selection error is properly working. ITS SRs required to ensure the rod block function operation of the rod 3.3.2.1.2 and 3.3.2.1.3 demonstrate the proper not need to be included in do block function. Therefore, these tests The requirements of the LCO and the ITS to ensure RWM remains Operable. definition of OPERABILITY the associated RWM surveillances and the required to be in the ITS to suffice. As such, these details are not and safety. Changes to the provide adequate protection of public health by the provisions relocated requirements in the TRM will be controlled of 10 CFR 50.59.
Page 4 of 9 Revision K JAFNPP
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (GENERIC)
The details in CTS 4.3.B.3.a.4 related to Test the performance of the Rod LA3 is proposed to be Worth Minimizer (RWM) Channel Functionaldetails do not need to be relocated to the Bases. These testing the RWM remains Operable. The included in the Specifications to ensure the RWM to be Operable and the requirements of ITS 3.3.2.1 which require Changes to the Bases will be definition of OPERABILITY suffices. Bases Control Program described in controlled by the provisions of the Chapter 5 of the ITS.
LA4 Not Used.
The detail in CTS Table 3.2-3 that the Rod The Block Monitor is Flow-Biased LA5 requirement in ITS LCO is proposed to be relocated to the Bases. for each Function in 3.3.2.1 that the control rod block instrumentation specific requirement in ITS Table 3.3.2.1-1 shall be OPERABLE and the Rod Block Monitor-Upscale Table 3.3.2.1-1 (Function 1.a) for theinstrumentation remains OPERABLE.
Function is sufficient to ensure the channel. As such.
The Bases describes the design of the ininstrumentation the ITS to provide adequate these details are not required to be to the Bases will be protection of public health and safety. Changes Program described in controlled by the provisions of the Bases Control Chapter 5 of the ITS.
The requirement in CTS 3.3.B.3.a and CTS 3.3.B.3.c that the second LA6 or a "reactor engineer" individual be a "reactor" or "senior" operator In addition, the requirement is proposed to be relocated to the Bases. have no other concurrent in CTS 3.3.B.3.c that the individuals shall(when the rod worth minimizer duties during rod withdrawal or insertion is also proposed to be is inoperable and a control rod is being moved) minimizer is inoperable during relocated to the Bases. If the rod worthActions C.2.2 and D.1 require a reactor startup, ITS 3.3.2.1 Required rods is in compliance with bank the verification of movement of control a second licensed operator or by position withdrawal sequence (BPWS) by staff during control rod another qualified member of the technical individuals and, for Required movement. The Bases identifies these individuals shall have no other Action C.2.2 only, states that thesedetails are not required to be in the concurrent duties. As such, these health and safety. Changes ITS to provide adequate protection of public provisions of the Bases Control to the Bases will be controlled by the ITS.
Program described in Chapter 5 of the
__ra F 0I J Revision K JAFNPP ravc %J
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC) 2, Action B), and CTS 3.3.B.5 Li The requirements in Table 3.2-3 (Notecontrol rod pattern have been concerning operations on a limiting rod pattern is defined as operating deleted. Since a limiting control as APLHGR or MCPR), the condition is on a power distribution limit (such power distribution limits does not extremely unlikely. The status of and therefore, no additional affect the OPERABILITY of the RBM required (e.g., that it be tripped requirements on the RBM System are while on a limiting control rod immediately with a channel inoperable distribution limits are pattern). Adequate requirements on power 3.2. Furthermore, due to the specified in the LCOs in ITS Section a limiting control rod pattern, improbability of operating on orbe above required. Therefore, the current all the ACTIONS would almost never by M3 are acceptable for Actions in Table 3.2-3 Action B as modified as ITS 3.3.2.1 ACTIONS and inoperabilities of the RBM and are included B.
CTS 4.2.C (Table 4.2-3) requires an per Instrument Check (Channel Check) of L2 day. ITS 3.3.2.1 does not the RBM Upscale and Downscale onceFunctions. The RBM automatically re require a Channel Check of these selected and retains the latest nulls itself whenever a control rodis isselected, making the performance of setting until another control rod (i.e., a daily channel check) a Channel Check during static conditions time a control rod is of no safety benefit. Specifically, at the readjusts its input and selected for movement, the RBM automaticallly with the rod selected output readings (different LPRM inputs associatedAt this time, the operator is and re-normalization), i.e., "renulling."
of the control rod movement and RBM in direct observation and monitoring instrument check during response: in essence, performingitsa continuous safety function (i.e., during control the time the RBM is performing check of the RBMs during rod withdrawal). Therefore, a routine daily that occurs when a control rod static conditions, prior to the renulling of safety. Accordingly, the is selected for movement, adds no assurance this instrument is acceptable.
elimination of a formal Channel Check for of the rod block function L3 CTS 4.3.B.3.a.4 requires a demonstration rod withdrawal. ITS during startup, prior to the start of control TEST of the RWM every 92 days 3.3.2.1 will require a CHANNEL FUNCTIONAL will be modified by a Note in MODE.2 (SR 3.3.2.1.2). ITS SR 3.3.2.1.2 is not required during a stating that the CHANNEL FUNCTIONAL TESTrod is withdrawn at
- 10% RTP in startup until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control the change in Frequency to 92 MODE 2. The addition of this Note and a CHANNEL FUNCTIONAL TEST less days makes the proposed requirement for is not required until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restrictive because the Surveillance Test and the test is not required after the RWM is required to be Operable.
Revision K JAFNPP rage b uv
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES LESS RESTRICTIVE (SPECIFIC)
L3 (continued) previous 92 days. In to be performed at startup if performed in the required in MODE 1 in addition, a CHANNEL FUNCTIONAL TEST will be after Thermal Power 1 hour accordance with SR 3.3.2.1.3, but not until does not monitor core is g 10% RTP (see M5). The Rod Worth Minimizer rod patterns as a thermal conditions but simply enforces preprogrammed in selecting or backup intended to prevent reactor operator error system, as shown by positioning control rods. The RWM is a reliable successful completion of both a review of maintenance history and by the effect on safety due previous startup surveillances. As a result, the increased testing to the extended Surveillance is small. Also, prior to each startup increases the wear on the instruments, thereby additional Surveillance reducing overall reliability. Therefore, an needed to assure the other than the quarterly Surveillance is not function. In addition, instruments will perform their associated safety CHANNEL FUNCTIONAL TEST.
a 92 day other similar rod block functions have way the required The Note changes are acceptable since the only in the specified condition Surveillances can be performed prior to entry of these devices is not is by utilizing jumpers or lifted leads. Usemay significantly increase recommended since minor errors in their use which is a precursor to a the probability of a reactor ransient or event is allowed to conduct the previously analyzed accident. Therefore, time Surveillances after entering the specified condition.
to verify the L4 The Frequency in CTS 4.3.B.3.a and CTS 4.3.B.3.bstartup, prior to the correctness of the RWM program sequence during prior to attaining start of control rod withdrawal and during shutdown 10% rated power during rod insertion has OPERABLE been changed to require the following loading of verification only prior to declaring RWM since this is when rod the Sequence into RWM. This change is acceptable is consistent with sequence input errors are possible. This change NUREG-1433, Revision 1.
CHANNEL CALIBRATION L5 The proposed change adds Note 1 to the quarterly for the RBM Upscale and Surveillance Requirement in CTS Table 4.2-3 the neutron detectors from Downscale Functions (SR 3.3.2.1.5) excludingis a complete check of the the Surveillance. The CHANNEL CALIBRATION verifies that the channel instrument loop and the sensor. The test the necessary range and responds to the measured parameter within from the CHANNEL accuracy. The neutron detectors are excluded with minimal drift, and CALIBRATIONS because they are passive devices meaningful signal. Changes in because of the difficulty of simulating a
..... 7 f Revision K JAFNPP ravu i
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L5 (continued) by performance of the neutron detector sensitivity are compensated for and the 1000 MWD/T LPRM 7 day calorimetric calibration (SR 3.3.1.1.2) The change is consistent calibration against the TIPs (SR 3.3.1.1.7).
with NUREG-1433, Revision 1.
CTS 4.3.B.5 requires the performance ofprior a functional test on a RBM when L6 to the withdrawal of the a limiting control rod pattern exists is proposed to be deleted designated rod(s). This testing requirement Operation with a limiting from the current Technical Specifications. on a power distribution control rod pattern is analogous to operating no correlation between power limit, such as APLHGR or MCPR. There is the operability of the RBM.
distribution limits and its affect on testing of the RBM based on the Therefore, initiation of surveillancedoes not increase the likelihood of status of power distribution limits operating on a limiting identifying an inoperable RBM. In fact, since this surveillance requirement control rod pattern is extremely unlikely, Furthermore, an analysis of the would most likely never be performed. of RBM instrument operating experience associated with the performance (CTS Table 4.2-3) functional testing and calibration testing tests, which are performed at a 92 demonstrates that these surveillance degree of reliability for a day interval, are indicative of a very high and their associated RBM instrument channel. These testing requirements testing at 92 day test intervals (i.e., functional/calibration As discussed in intervals) are maintained in the ITS by SRall3.3.2.1.5. requirements of a the DOC A4, calibration testing includesbased the on the above evaluation, channel functional test. Accordingly, of this CTS testing the Licensee has concluded that the deletion on nuclear safety. This requirement would have an insignificant affect 1.
change is consistent with NUREG-1433, Revision L7 CTS Table 3.2-3 requires the RBM to be ITS, Operable when reactor power is requirement is greater than or equal to 30%. In the (a) this except when a peripheral maintained in Table 3.3.2.1-1 Footnote acceptable since with a control rod is selected. This change is of control rod withdrawal peripheral rod selected the consequences In a addition, this change is error event will not exceed the MCPR SL. That is when a consistent with the design of the RBM circuitry. automatically bypassed and peripheral control rod is selected the RBM is the output set to zero.
Page 8 of 9 Revision K JAFNPP
DISCUSSION OF CHANGES ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC) submit a report to the LB The requirement in CTS 3.3.B.3.d to prepare and RWM Operable is NRC within 30 days of a plant startup without the Specifications. This special proposed to be deleted from the Technical the action taken to report states the reason for the RWM inoperability, RWM to an operable restore it. and the schedule for returning the to review the status. This special report provides a mechanism but provides no appropriateness of licensee activities after-the-fact, submitted (i.e., no requirement regulatory authority once the report is of 10 CFR 50, for NRC approval). The Quality Assurance requirements corrective actions will Appendix B, provide assurance that appropriate to be provided to the be taken. Given that the report was required report completion and Commission within 30 days following the startup, operation of the facility submittal was clearly not necessary to assure startup of the unit and in a safe manner for the interval between on the above evaluation, submittal of the report. Accordingly, based to be in the current Technical the RWM Special Report is not required with NUREG-1433.
Specifications nor the ITS. This change is consistent TECHNICAL CHANGES - RELOCATIONS Notes to these Tables R1 CTS 2.1.A.1.d, Tables 3.2-3 and 4.2-3 and the Block functions include the Safety Limits, LCOs and SRs for Rod SRMs, and Scram Discharge Volume Level.
associated with the APRMs, IRMs, Requi rements These requirements are being relocated to the Technical Volume (SDV) rod Manual (TRM). The APRM, IRM, SRM, and Scram Discharge when plant blocks are intended to prevent control rod withdrawal there are no safety conditions make such withdrawal imprudent. However, mitigate or analyses that depend upon these rod blocks to prevent, or transients.
establish initial conditions for design basis accidents that the loss of the The evaluation summarized in NEDO-31466 determined blocks would be a non APRM, IRM, SRM, and Scram Discharge Volume rod and offsite significant risk contributor to core damage frequencydetermined to be releases. The results of this evaluation have been does not satisfy applicable to JAFNPP. Therefore, this instrumentation Specifications as 10 CFR 50.36(c)(2)(ii) for inclusion in the Technicalto the JAFNPP documented in the Application of Selection Criteria by reference Technical Specifications. The TRM will be incorporated the TRM will be into the UFSAR at ITS implementation. Changes to controlled by the provisions of 10 CFR 50.59.
Page 9 of 9 Revision K JAFNPP
Control Rod Block Instrumentation 3.3.2.1 FREQUENC SURVEILLANCE REQUIREMENTSSURVEILLANCE continued FREQUENCY SR 3.33.22.1.0@-----Not afterrequired to be NOTE--7-------
reactor mode performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> switch is in the. ...
shutdown position.
Perform CHANNEL FUNCTIONAL TEST.
NOT -----.----
SR 3 . 3 . 2 . 1 .VS
/ Neutron detectors are excluded CHANNEL CALIBRATION.t ~PPerform the Prior to SR 3.3.2. Verify control rod sequences input to 7f 9 RWM are in conformance with BPWS.
declaring RWM OPERABLE following loading of sequence into 3 RWM
- ,-*- ?,,- + -0 . -
-,I't
- 33. o ,
.5.J-17 Rev 1, 04/07/95 BWR/4 STS
Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (pa0 1 of 1)
Control Rod slock 1nstr uwstion APPLICAILE APPLI CABLE NWSOS OR OTNE.
SPECIFIED COMMO1NS CPANNELS (6
20 -:Lý r.* *.-3j f.
Inop 21.0 divisi.onso
~.Dow~l*U3.3 S*3,3.2.1,* .full scale 1.Syaa i *La ,,,..T,~ ... C=d,)e """""
U ~1 SL3.~
1., 120 m a a] sa,o;;.
SR 3.3.2.1.2 NA pot-] 32 Rod Worth Hinimizer 3
, 3.2.1.3 I,&~
SR 3.3.2. 1.&
- 3, sa 3.3.2.1.&fNA Position 7
SC) THEESAL A E843%
a nld C90MX . an (d) THERPAL R t 90%RTP and isA &. "C~ t(
P POWER~ 643% mold A0TPand Wit 1 (a) THE With THERMAL POWER 5s p L ý
<~ Q
)Reactor switch in thie Shutdown posMtOn.
monde 3.3-20 Rev 1, 04/07/95 BWR/4 STS tZeAJIS'Ovl K\
1 JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433. REVISION ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION RETENTION OF EXISTING REQUIREMENT (CLB)
Calibration Surveillance, is ITS SR 3.3.2.1.5, the 92 day RBM Channel excludes CLB1 the recirculation loop modified by the addition of Note 2 that Table 4.1-2, "Flow Bias flow signal portion of the channel. -CTS test with standard ILŽ Signal," requires an "internal power and flowinterval. This flow bias pressure source" calibration on a refueling Flux-High (Flow Biased) signal provides input to both the APRM Neutron control rod block Function.
RPS scram Function and to the RBM-Upscale bias signal line item, thus CTS 3/4.2.C does not have a specific flow the calibration required by CTS Table 4.1-2 covers the RBM requirements recirculation loop flow signal.
as well as the RPS requirements, of the requirement in SR 3.3.2.1.5 is Therefore, the RBM Channel Calibration signal portion of the modified to exclude the recirculation loop flow of the recirculation channel. ITS SR 3.3.2.1.8 requires calibration and the Bases notes that loop flow signal portion of the channel SR 3.3.1.1.12.
performance of ITS SR 3.3.2.1.8 also satisfies ITS in the COLR. This was CLB2 The Allowable Value of the RBM upscale is located No. 162. This accepted in JAFNPP Technical Specification inAmendment Generic Letter 88-16 for allowance is consistent with the guidance from the Technical the removal of cycle-specific parameter limits Specifications to the COLR.
CLB3 The CTS allows only one startup with the RWM inoperable (i.e.,
inoperable prior to withdrawal of the first Action12 control rods) per C.2.1.2, "performed in calendar year. The words in ISTS Required startups with the RWM the last calendar year" could allow multiple since the check only looks at inoperable in the current calendar year, the last (i.e., previous) calendar year. has Therefore, consistent with the current licensing basis, the word "last" been changed to "current."
IMPROVEMENT (PA)
PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL PAl None PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)
The RWM is required to be Operable at < 10% RTP as specified in CTS bases DB1 the design has been 4.3.B.3.a.4. This requirement is consistent withvalue of 10%
analysis assumptions. Therefore, the bracketed retained in the ITS throughout the Specification.
DB2 The brackets have been removed and the Surveillance Frequency of 92 days Frequency is is retained in ITS SR 3.3.2.1.2 and SR 3.3.2.1.3. This Page 1 of 3 Revision K JAFNPP
REVISION 1 JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, ITS: 3.3.2.1 - CONTROL ROD BLOCK INSTRUMENTATION PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB) with M4. The bracketed DB3 ITS SR 3.3.2.1.4 has been added in accordanceto 92 days and the bracketed Frequency of 18 months has been changed excluded) retained. The Surveillance Note (Neutron detectors are to the JAFNPP plant design.
surveillance has been re-written to conform when required.
The Surveillance ensures the RBM is Operable of RBM Upscale control rod block DB4 ISTS SR 3.3.2.1.7. (Channel Calibration loop from signal portion of the Function (except for the recirculation block Function channels) is channel) and RBM Downscale control rod the surveillance has been currently performed every 92 days therefore as SR 3.3.2.1.5.
placed in its appropriate location and renumbered where applicable. This Subsequent surveillances have been renumbered, methodology in determining the Surveillance Frequency is consistent with Since the Calibration associated Allowable Values for these Functions. for a CHANNEL FUNCTIONAL is performed every 92 days there is no need from these Functions in TEST, therefore SR 3.3.2.1.1 has been removed the Table.
has been added in accordance DB5 SR 3.3.2.1.1, a CHANNEL FUNCTIONAL TEST, bracketed Frequency of 92 days with Ml for the RBM Inop function. The is retained since it is consistent with NEDC-30851-P-A.
The bracketed Surveillance Frequency of theITS SR 3.3.2.1.6 is changed from DB6 associated Bases for this 18 months to 24 months as justified in assumes a Frequency of 24 surveillance. The trip setpoint methodology months between calibrations.
The bracketed Surveillance Frequency of testITS SR 3.3.2.1.7 has been DB7 should be performed during a changed from 18 to 24 months since the transients as described in the plant outage to minimize any unplanned Bases for this SR.
proper number of channels DB8 The brackets have been removed and the The values are included for each Function in Table 3.3.2.1-1.
in CTS Table 3.2.3 for consistent with the current requirements the Rod Worth Minimizer. The Functions l.a, 1.c, and CTS 3.3.B.3 for and Function 3 (Reactor Mode requirements for Function 1.b (RBM-Inop) with M1 and M2. The Switch-Shutdown) have been added in accordance with the plant design.
specified number of channels are consistent are not applicable to JAFNPP.
DB9 Table 3.3.2.1-1 Functions 1.b. 1.c and 1.f from the Table. Subsequent Therefore these Functions have been removed Functions have been renumbered, where applicable.
f 11 2 Revision K r awc V*
Control Rod Block Instrumentation B 3.3.2.1 between successive adjusted to account for instrument drifts s ecific setpoint calibrations consistent with the plant Laethodol ogy f~ tneutron detectors -- -from the CHANNEL CALIBRATION because they are passive devices, with min*
of the difficulty of simulating a drift, and because are adequately tested meaningful signal. Neutron detectors in SL.3.1.14.AdR,3.3. 1,1 sSz S.. 2.
O44 /S The retluency is based n the assumption of ad* 40 interval in the determination of the magnitude calibration of equipment drift in the setpoint analysis..
"ora;X~
gin'6 "I The RWN will only enforce the proper control rod sequence if input into the RWN computer.
the rod sequence is properly into the that the proper sequence is loaded This SR ensures function. The RMN so that it can perform its intended is performed once prior to declaring RbH4 Surveillance , since this OPERABLE following loading of sequence into RWs iswhen rod sequence input errors are possible.
b2o~
SDINSERT SR-1 SR 3.3.2.1.5 is modified by two Notes. Note 1 to SR 3.3.2.1.5 excludes 6 DpltINSERT SR-2 loop flow signal portion of Note 2 to SR 3.3.2.1.5 excludes the recirculation portion of the channel is the channel from the CHANNEL CALIBRATION, since this calibrated by SR 3.3.2.1.8.
k> INSERT SR-3 all portions of channel SR 3.3.2.1.8 is modified by a Note that excludesCHANNEL CALIBRATION.
except the recirculation loop flow signal from results in calibration of the SR 3.3.2.1.5, in conjunction with SR 3.3.2.1.8, entire channel. Since the recirculation loop flow signal is also a portion of RPS scram Function channels (Table the APRM Neutron Flux-High (Flow Biased) 2.b),
3.3.1.1-1. RPS Instrumentation, Function satisfactory performance of completion of SR 3.3.1.1.12 for the SR 3.3.2.1.8 also results in satisfactory Biased) associated APRM Neutron Flux-High (Flow RPS scram Function channels.
A INSERT SR-4 The Frequency of SR 3.372..'.8 is based upon the assumption of a 24 month of the equipment calibration interval in the determination of the magnitude drift in the setpoint analysis.
Insert Page B 3.3-54 Revision K
Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued) FREQUENCY SURVEILLANCE SR 3.3.2.1.2 ------------------ NOTE ------------------
Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after any control rod is withdrawn at S10% RTP in MODE 2.
92 days Perform CHANNEL FUNCTIONAL TEST.
SR 3.3.2.1.3 .................. NOTE ...................
Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is
- 10% RTP in MODE 1.
Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.2.1.4 ...................
NOTE------------------.
Neutron detectors are excluded.
Verify the RBM is not bypassed: 92 days
- a. When THERMAL POWER is k 30% RTP: and
- b. When a peripheral control rod is not selected.
SR 3.3.2.1.5 ..................
- 1. Neutron detectors are excluded.
2.
NOTES--...............
For Function l.a, the recirculation loop flow signal portion of the channel is excluded.
I Perform CHANNEL CALIBRATION.
92 days (continued) 3.3-18 Amendment (Rev. K)
Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued) FREQUENCY SURVEILLANCE SR 3.3.2.1.6 Verify the RWM is not bypassed when 24 months THERMAL POWER is -,10% RTP.
SR 3.3.2.1.7 .................. NOTE ------------------
Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after reactor mode switch is in the shutdown position.
24 months Perform CHANNEL FUNCTIONAL TEST.
SR 3.3.2.1.8 ------------------NOTE ------------------
For Function l.a. all portions of the channel except the recirculation loop flow signal portion are excluded. A 24 months Perform CHANNEL CALIBRATION.
Prior to 1A SR 3.3.2.1.9 Verify control rod sequences input to the declaring RWM OPERABLE RWM are in conformance with BPWS.
following loading of sequence into RWM 3.3-19 Amendment (Rev. K)
Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)
Control Rod Block Instrumentation APPLICABLE MODES OR OTHER ALLOWABLE SPECIFIED REQUIRED SURVEILLANCE CHANNELS REQUIREMENTS VALUE FUNCTION CONDITIONS
- 1. Rod Block Monitor 2 SR SR 3.3.2.1.4 3.3.2.1.5 As specified the COLR in
- a. Upscale (a)
SR 3.3.2.1.8 (a) 2 SR SR 3.3.2.1.1 3.3.2.1.4 NA
- b. Inop (a) 2 SR 3.3.2.1.4 SR 3.3.2.1.5 k 2.5/125 of divisions
- c. Downscale full scale 1 (b), 2 (b) 1 SR 3.3.2.1.2 SR 3.3.2.1.3 NA
- 2. Rod Worth Minimizer SR 3.3.2.1.6 SR 3.3.2.1.9 (c) 2 SR 3.3.2.1.7 NA
- 3. Reactor Mode Switch- Shutdown Position rod selected.
(a) THERMAL POWER z 30% RTP and no peripheral control (b) With THERMAL POWER s 10 RTP.
(c) Reactor mode switch in the shutdown position.
3.3-20 Amendment (Rev. K)
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.2 and SR 3.3.2.1.3 (continued)
REQUIREMENTS state of a single contact of the relay. This clarifies what is an acceptable CHANNEL FUNCTIONAL TEST of a relay. This is acceptable because all of theTechnical other required contacts of Specifications and the relay are verified by'other once per non-Technical Specifications tests at least extensions. The refueling interval with the applicable is" erformed by CHANNEL FUNCTIONAL TEST for the RWM rod no ce with attempting to withdraw a control a control rod block the prescribed sequence and verifying is not required occurs. As noted in the hour SRs, SR 3.3.2.1.2 to be performed until 1 after any control rod is withdrawn at < 10% RTP in MODE 21 and, SR 3.3.2.1.3 is not required to be performed until hour after THERMAL POWER is
< 10% RTP in MODE 1. This allows entry into MODE 2 for SR 3.3.2.1.2. and entry into MODE 1 when THERMAL POWER is required
- 10% RTP for SR 3.3.2.1.3, to performis the not met per day Frequency Surveillance if the 92 allowance 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is based on operating SR 3.0.2. The consideration of providin a reasonable experience and in time in which to complete analysis the SRs. The 92 gay Frequencies are based on reliability (Ref. 9).
SR 3.3.2.1.4 The RBM is automatically bypassed when power is below a control rod is selected.
specified value or if a peripheral from the APRM signals input to The power level is determined must be verified each RBM channel. The automaticInbypass addition, it must also be periodically to be < 30% RTP. bypassed when a non-peripheral verified that the RBM is not one non-peripheral control rod control rod is selected (only If any bypass setpoint is is required to be verified). RBM channel is considered nonconservative, then the affected APRM channel can be placed inoperable. Alternatively, the(i.e., enabling the in the conservative condition the SR is met and nonbypass). If placed in this condition, inoperable. As noted, the RBM channel is not considered from the Surveillance because neutron detectors are excluded drift, and because of they are passive devices, witha minimal meaningful signal. Neutron the difficulty of simulating and detectors are adequately tested in SRis 3.3.1.1.2 based on the actual SR 3.3.1.1.7. The 92 day Frequency for these channels.
trip setpoint methodology utiTized SR 3.3.2.1.5 and SR 3.3.2.1.8 A CHANNEL CALIBRATION is a com.plete check of the instrument loop and the sensor. This test verifies the channel within the necessary responds to the measured parameter (continued)
""31_* Revision K D ,.). o- *Ju JAFNPP
Control Rod Block Instrumentation B 3.3.2.1 BASES SR 3.3.2.1.5 and SR 3.3.2.1.8 (continued) AL\
SURVEILLANCE REQUIREMENTS range and accuracy. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint methodology.
Notes. Note 1 to CHANNEL SR 3.3.2.1.5 is modified by twodetectors I/K SR 3.3.2.1.5 excludes neutron from the with minimal CALIBRATION because they are passive devices, of simulating a drift, and because of the difficulty are adequately tested meaningful signal. Neutron detectors Note 2 to SR 3.3.2.1.5 in SR 3.3.1.1.2 and SR 3.3.1.1.7. portion of the excludes the recirculation loop flow signal since this portion of channel from the CHANNEL CALIBRATION.
the channel is calibrated by SR 3.3.2.1.8.
SR 3.3.2.1.8 is modified by a Note that excludes all portions of channel except the recirculation loop flow SR 3.3.2.1.5. in signal from CHANNEL CALIBRATION. results in calibration of the conjunction with SR 3.3.2.1.8.
entire channel. Since the recirculation loop flow signal is Neutron Flux-Higgh (Flow Biased) also a portion of the APRM 3.3.1.1-1. RPS RPS scram Function channels (Table satisfactory performance of Instrumentation, Function 2.b),
in satisfactory completion of SR 3.3.2.1.8 also results Neutron Flux- High SR 3.3.1.1.12 for the associated APRM channels.
(Flow Biased) RPS scram Function The Frequency of SR 3.3.2.1.5 is based upon the assumption of a 92 day calibration interval in the determination of the in the setpoint analysis. The magnitude of equipment drift based upon the assumption of a Frequency of SR 3.3.2.1.8 is in the determination of the 24 month calibration interval analysis.
magnitude of the equipment drift in the setpoint SR 3.3.2.1.6 The RWM is automatically bypassed when power is above a steam spcified value. The power level pressure. Thefrom is determined automatic flow signals compensated for steamperiodically to be bypass setpoint must be verified setpoint is 1-0% RTP. If the RWM low power inoperable.
nonconservative, then the RWM is considered channel can be placed in Alternately, the low power setpoint If placed in the the conservative condition (nonbypass).
is met and the RWM is not nonbypassed condition, the SRFrequency considered inoperable. The is based on the trip setpoint methodology utilized for the low power setpoint channel.
(continued)
B 3.3-57 Revision K JAFNPP
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.7 REQUIREMENTS is performed for the Reactorentire Mode (continued) A CHANNEL FUNCTIONAL TEST Function to ensure that the Switch-Shutdown Position A successful channel will perform the intendedof function.a channel relay may be test of the required contact(s)of the change of state of a performed by the verification This clarifies what is an single contact of the relay. TEST of a relay. This is acceptable CHANNEL FUNCTIONAL other reguired contacts of the acceptable because all of theTechnical Specifications and relay are verified by other tests at least once per non-Technical Specifications extensions. The refueling interval with the applicable Mode Switch-Shutdown CHANNEL FUNCTIONAL TEST for the byReactor_ to withdraw any Position Function is performedmode atte.pting switch in the shutdown control rod with the reactor rod block occurs.
position and verifying a control As noted in the SR, the Surveillance is not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after the reactor mode switch is in the shutdown position, since testing of this interlock with the reactor mode switch in any other position cannot be prformed without using jumpers,MODES lifted leads, or movable Sinks. This allows entry into 3 and 4 if the 24 month 3.0.2. The Frequency is not met per SR and in consideration 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> allowance is of based on operating experience which to complete the SRs.
providing a reasonable time in The 24 month Frequency is based on that the need to perform this Surveillance under the conditions apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power. pass Oerating experience has shown these components usually t e Surveillance when performed at the 24 month Frequency.
SR 3.3.2.1.9 I The RWM will only enforce the properinto control rod sequence if the rod sequence is properly input the RWM computer.
This SR ensures that the proper sequence is loaded into the function. The RWM so that it can perform its intended (continued)
"D*_Revision
'*~ K D O.O'*Q JAFNPP
Control Rod Block Instrumentation B 3.3.2.1 BASES SURVEILLANCE SR 3.3.2.1.9 (continued)
REQUIREMENTS Surveillance is performed once prior to declaring RWM OPERABLE following loading of sequence into RWM, since this is when rod sequence input errors are possible.
REFERENCES 1. UFSAR, Section 7.5.8.2.
- 2. UFSAR, Section 7.16.5.3.
- 3. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, Supplement for United States, Section S.2.2.1.5, (Revision specified in the COLR).
- 5. UFSAR, Section 14.6.1.2.
- 6. NRC SER, Acceptance of Referencing of Licensing Topical Report NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, Revision 8, Amendment , December 27, 1987.
- 7. Letter from T.A. Pickens (BWROG) to G.C. Lainas (NRC),
Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A, BWROG-8644, August 15, 1986.
GENE-770-06-1-A, Addendum to Bases for Changes to
- 8. Out-of-Service Surveillance Test Intervals and Allowed Times for Selected Instrumentation Technical Specifications, December 1992.
- 9. NEDC-30851P-A, Supplement 1, Technical Specification "Improvement Analysis for BWR Control Rod Block Instrumentation, October 1988.
B 3.3-59 Revision K JAFNPP
- p&C;-fCa 4
ioV 3-.-*4 -*I JAFNIJP Minimum Number of IfAprfecoA iOV (4 Operable Instrument Aipplicable Modes Channels Per Trip System (Notes 1 & 21 Trip Function -(
- 1 psig Run [moi *i-J 1-*
Reactor Pressure - High Run P4 ) t6.J
[oE
[LCD 3-q.I.lJ2 *10:5.4 iný II.53 Reactor Water Level - Low Low ELto '3.t10A.'12 Amendment No. 227r3;-2--64, 2 73 76a r,*e s* of 6I(
3>
JAFNPP THIS PAGE INTENTIONALLY BLANK p-
"I Amendment No. 237 , 2 7 3 76c pa 1 v/oP (o KJp.ev t o'
DISCUSSION OF CHANGES ITS SECTION 3.3.4.1: ATWS-RPT INSTRUMENTATION ADMINISTRATIVE CHANGES A7 (continued) for the Dete1rmination of Setpoints far Nuclear Safety-Related' change revises the terminology used in the CTS Instrumentation." This Since the from "Trip Level Setting" to "Allowable Value". at the same numerical value, instrumentation will be declared inoperable This change is consistent this change is considered administrative.
with NUREG-1433, Revision 1.
for operations for A8 CTS 3.2.G makes reference to the limiting conditionspumps in CTS Table the instrumentation that trip(s) the recirculationPump Trip instrumentation 3.2-7. CTS 4.2.G requires the Recirculationtest the associated logic as to be functional tested, calibrated and to to the Tables has been indicated in Table 4.2-7. This cross-reference All of the deleted since ITS 3.3.4.1 does not include aandTable.4.2-7 are included in the technical requirements of CTS Tables 3.2-7 Since this change ITS 3.3.4.1 LCO, Applicability, and Surveillances. is considered simply deletes this cross-reference, this change NUREG-1433, Revision 1.
administrative. This change is consistent with A9 Not used.
TECHNICAL CHANGES - MORE RESTRICTIVE Channel M1 CTS Table 4.2-7 requires a daily performance ofto anbeATWS-RPT performed every 12 Check. ITS SR 3.3.4.1 will require this test is to ensure that a gross hours. The purpose of the Channel Check Thus, performance of the failure of instrumentation has not occurred. outright channel channel check helps to ensure that an undetected is consistent with NUREG failure is limited to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. This change 1433, Revision 1.
Revision K JAFNPP Page 3 of 8
DISCUSSION OF CHANGES ITS SECTION 3.3.4.1: ATWS-RPT INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE (continued)
Values (A7) in CTS Table M2 This change replaces the setpoints or Allowable with : 1153 psig (ITS 3.2-7, Reactor Pressure-High
- 1155 psigThe Allowable Value (to be SR 3.3.4.1.4, Reactor Pressure-High): and the Trip Setpoint (to be included in the Technical Specifications)established consistent with the included in plant procedures) have been Loop Accuracy and NYPA Engineering Standards Manual, IES-3A, "Instrument used to determine Setpoint Calculation Methodology." The methodologythe methodology discussed in the "Allowable 4 Value" is consistent with the Determination of ISA-$67.04-199 , Part II, "Methodologies for The proposed Setpoints for Nuclear Safety-Related Instrumentation." is met. All design value will ensure the most limiting requirement as ensuring that limits, applied in the methodologies, were confirmed system is maintained.
applicable design requirements of the associated has been added to CTS M3 A NOTE (ITS 3.3.4.1 Required Action A.2 Note) the action to place a channel Table 3.2-7 Note l.a which specifies that channel is a result of an in trip is not applicable if the inoperable for opening, ATWS-RPT inoperable breaker. If a breaker is inoperable operating trip capability is not maintained for the associatedchannel in trip would not be recirculation pump, therefore placing the the channel would not cause an appropriate action to take since tripping the action should be the inoperable breaker to trip. In this 1:condition, however, the CTS does not taken according to CTS Table 3.2-7 Note a tripped condition for this explicitly prohibit placing a channel in above, has been added to the situation. Therefore, a NOTE, as described the addition of this NOTE to the CTS Table 3.2-7 Note l.a. Accordingly, This change is consistent CTS is considered a more restrictive change.
with NUREG-1433, Revision 1.
TECHNICAL CHANGES LESS RESTRICTIVE (GENERIC)
Level Setting of the Reactor LA1 The detail in CTS Table 3.2-7 that the Trip from the Top of Active Water Level - Low Low Trip Function is referenced the Bases. CTS 1.0.Z Fuel (TAF) is proposed to be relocated to of Active Fuel, corresponding to the definition specifies that the Top bundle, is located top of the enriched fuel column of each fuel lowest point in the inside 352.5 inches above vessel zero, which is the bottom of the reactor pressure vessel. (See General Electric drawing to be relocated to the No. 919D690BD). These details are also proposed the ATWS instrumentation Bases. The requirement in ITS LCO 3.3.4.1 that be OPERABLE, the requirements for each Function in Table 3.3.4.1-1 shall the definition of in the Table including the Allowable Value, Requirements are Operability, the proposed Actions, and Surveillance
-^ a ^Revision K JAFNPP Page 4 v=
ATWS-RPT Instrumentation 3.3.4.0 0--l 2.G7\
rkit1 v-'O ASt Rev 1, 04/07/95 BWR/4 STS 3.3-35 REVISION K
0 INSERT B 3.3.4.1-1 The Allowable Value was derived from the analysis performed in Reference 4. It
( j INSERT Function a coolant injection (HPCI) also provides an opportunity for the high pressure to recover water level if and reactor core isolation cooling (RCIC) systems is referenced from a level of feedwater is not available. The Allowable Value inside bottom of the RPV and water 352.56 inches above the lowest point in the (Ref. 3).
also corresponds to the top of a 144 inch fuel column (as described in Table The HPCI, RCIC and ATWS-RPT initiation functions 1; and LCO 3.3.4.1.a 3.3.5.1-1, Function 3.a: Table 3.3.5.2-1, Functionthe reactor vessel water level including SR 3.3.4.1.4, respectively) describeThe Allowable Values associated initiation function as "Low Low (Level 2)." is different from the Allowable with the HPCI and RCIC initiation function function as the ATWS function Value associated with the ATWS-RPT initiation consistent with the has a separate analog trip unit. Nevertheless, the "Low Low (Level 2)"
nomenclature typically used in design documents.these three initiation designation is retained in describing each of functions.
Insert Page B 3.3-94 Revision K
ATWS-RPT Instrumentation B 3.3.4.22
.- CPDA BASES ACTIONS Ll
& Required Action B.1 is intended to ensure that appropriate actions are taken if multiple, inoperable, untripped channels within the same Function result in the Function not maintaining ATWS-RPT trip capability. A Function is when
'considered to be maintaining ATWS-RPT trip capability sufficient channels are OPERABLE or in trip such that the ATWS-RPT System will generate a trip signal from the given Function on a valid signal, and both recirculation pumps can be tripped. This requires: channelgAof the Function in-)
-- - t~ rip system to each be OPERABLE or in trip, and the tWi~irccuattion pump drive mtrbreakers to be OPERABLE or in trip.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is sufficient for the operator to take corrective action (e.g., restoration or tripping of channels) and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period and that one Function is still maintaining ATWS-RPT trip capability.
Required Action C.1 is intended to ensure that appropriate Actions are taken if multiple, inoperable, untripped channels within both Functions result in both Functions not maintaining ATWS-RPT trip capability. The description of a Function maintaining ATbIS-RPT trip capability is discussed in the Bases for Required Action B.1 above.
The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is sufficient for the operator to take corrective action and takes into account the likelihood of an event requiring actuation of the ATWS-RPT instrumentation during this period.
not With any Required Action and associated Completion Time met, the plant must be brought to a MODE or other specified this condition in which the LCO does not apply. To achieve 2 within status, the plant must be brought to at least MODE 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (Required Action D.2). Alternately, the associated this recirculation pump may be removed from service since (continued)
B 3.3-96 Rev 1, 04/07/95 BWR/4 STS REVISION Y /<
ATWS-RPT Instrumentation B 3.3.4. P BASES (Th 2) rk ) )
S3.3.4 M1 (continued)
SURVEILLANCE SP "sn T Dr-Mnff something even more serious. itA CHANNEL CHECK will detect W%16W4 is key to verifying the gross channel failure; thus, operate properly between each instrumentation continues to
,.CHANNEL CALIBRATION.
greement criteria are determined by the plant staff based
\. _-"--J n a combination of the channel instrument uncertainties, including indication and readability. If a channel the is outside the criteria, it may be an indication that instrument has drifted outside its limit.
The Frequency is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the required channels of this LCO.
SR 3.3.4.,M.2 A CHANNEL FUNCTIONAL TEST Is performed on each required the channel to ensure that the Cn*ft channel will perform intended function.
d'- Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.
TM The Frequency of 92 da s is based on the reliability analysis of Reference Calibration of trip units provides a check of the actual must be declared inoperable if trip setpoints. The channel conservativ the trip setting is discovered to be less If the trip the Allowable Value specified in SR 3.3.4. 4. than the setting is discovered to be less conservative setting accounted for in the appropriate setpoint the methodology, but is not beyond the Allowable Value, (continued)
Rev 1, 04/07195 BWR/4 STS B 3.3-98 REVISION p X
ATWS-RPT Instrumentati B 3.3.4.0 on REFERENCES Ijgure FT7SAM, Q agt#P Srveillance Test ases for Changes ToSTimes 4p.770-06-n
- Allowed 0ut-of-ServicB For ineras 66NSelected Instrumentation Technical Speclficatlofls,(ýj Rev.-1, 04/07/95 UNK/4 ZIQ B 3.3-100 REVISON r
INSERT REF
Boiler. (GE
- 3. Drawing 11825-5.01-15D, Rev. D, Reactor Assembly Nuclear Drawing 919D690BD).
- 4. "ATWS Overpressure Analysis for FitzPatrick," GE-NE-A42-00137-2-01, March 2000.
Insert Page B 3.3-100 Revision K
ATWS -RPT Instrumentation 3.3.4.1 SURVEILLANCE REQUIREMENTS
. . . . NOTE ------------------------------------
status solely for performance of When a channel is placed in an inoperable Conditions and Required Actions required Surveillances, entry into associated the associated Function maintains may be delayed for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided ATWS-RPT trip capability.
..... ....-- .... o.....---- °--- -' ' . . . . . - - . . . . . .
cl- -
FREQUENCY FREQUENCY SURVEILLANCE 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.4.1.1 Perform CHANNEL CHECK.
92 days SR 3.3.4.1.2 Perform CHANNEL FUNCTIONAL TEST.
184 days SR 3.3.4.1.3 Calibrate the trip units.
24 months Perform CHANNEL CALIBRATION. The SR 3.3.4.1.4 Allowable Values shall be:
- a. Reactor Vessel Water Level- Low Low (Level 2): : 105.4 inches; and
- b. Reactor Pressure-High: K 1153 psig.
24 months SR 3.3.4.1.5 Perform LOGIC SYSTEM FUNCTIONAL TEST including breaker actuation.
I 3.3-31 Amendment (Rev. K)
ATWS-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE b. Reactor Pressure-High (continued)
SAFETY ANALYSES, LCOand that result in a pressure increase, counteracting the APPLICABILITY pressure increase by rapidly reducing core powerthe RPT generation. For the overpressurization event, aids in the termination of the ATWS event and, along with the safety/relief valves (S/RVs), limits the peak RPV pressure to less than the ASME Section III Code Service Level C limits (1500 psig).
The Reactor Pressure-High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure. Four channels of Reactor Pressure-High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Pressure-High Allowable Value is chosenIIIto provide an adequate margin to the ASME Section ode Service Level C allowable Reactor Coolant System pressure. The Allowable Value was derived from the analysis performed in Reference 4.
A Note has been provided to modify the ACTIONS related to ACTIONS Section 1.3, Completion ATWS-RPT instrumentation channels.
Times, specifies that once a Condition has been entered, subsequent divisions, subsystems, components, or variables or expressed in the Condition, discovered to be inoperable into not within limits, will not result in separate entry the Condition. Section 1.3 also specifies that Required Actions of the Condition continue to apply for each initial additional failure, with Completion Times based onActions for entry into the Condition. However, the Required inoperable ATWS-RPT instrumentation channels provide appropriate compensatory measures for separate inoperable channels. As such, a Note has been provided that allows separate Condition entry for each inoperable ATWS-RPT instrumentation channel.
(continued)
Revision K JAFNPP B 3.3-90
ATWS-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 (continued)
OPERABLE or recirculation pump MG drive motor breakers to be in trip.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is sufficient for thetripping 1 perator restoration or of to take corrective action (e.g., of an event channels) and takes into account the likelihood during requiring actuation of the ATWS-RPT instrumentation maintaining this period and that one Function is still ATWS-RPT trip capability.
C.1 appropriate Required Action C.1 is intended to ensure that untripped Actions are taken if multiple, inoperable, Functions not channels within both Functions result in both of a maintaining ATWS-RPT trip capability. The description is discussed Function maintaining ATWS-RPT trip capability in the Bases for Required Action B1 above.
operator to The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time is sufficient for the the likelihood take corrective action and takes into account ATWS-RPT of an event requiring actuation of the instrumentation during this period.
D.1 and D.2 Time not With any Required Action and associated Completion or other specified met, the plant must be brought to a MODE condition in which the LCO does not apply. To achieve this MODE 2 within status, the plant must be brought to at least the associated 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (Required Action D.2). Alternately, since this recirculation pump may be removed from service performs the intended function of the instrumentation Time of (Required Action D.1). The allowed Completion both 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is reasonable, based on operating experience, and to remove a to reach MODE 2 from full power conditions manner and recirculation pump from service in an orderly without challenging plant systems. Required Action D.1 is Action is modified by a Note which states that the Required the result of only applicable if the inoperable channel is the an inoperable RPT breaker. The Note clarifies Action would situations under which the associated Required be the appropriate Required Action.
(continued)
Revision K JAFNPP B 3.3-92
ATWS - RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.2 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the channel will perform the intended of a function. A successful test of the required contact(s) of the thannel relay may be performed by the verificatiofi change of state of a single contact of the relay. This of a clarifies what is an acceptable CHANNEL FUNCTIONAL TEST required relay. This is acceptable because all of the other contacts of the relay are verified by other Technical at Specifications and non-Technical Specifications tests least once per refueling interval with applicable extensions.
Any setpoint adjustment shall be consistent with the specific setpoint assumptions of the current plant methodology.
The Frequency of 92 days is based on the reliability analysis of Reference 5.
SR 3.3.4.1.3 Calibration of trip units provides a check of the actual if trip setpoints. The channel must be declared inoperablethan the trip setting is discovered to be less conservative trip the Allowable Value specified in SR 3.3.4.1.4. If the the setting is discovered to be less conservative than setting accounted for in the appropriate setpoint the methodology, but is not beyond the Allowable Value, of the channel performance is still within the requirements setpoint plant safety analysis. Under these conditions, the must be readjusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.
The Frequency of 184 days is based on the reliability, accuracy, and low failure rates of these solid-state electronic components.
SR 3.3.4.1.4 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary (continued)
B 3.3-94 Revision K JAFNPP
ATWS-RPT Instrumentation B 3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.4 (continued)
REQUIREMENTS range and accuracy. CHANNEL CALIBRATION leaves the channel successive adjusted to account for instrument drifts betweensetpoint calibrations consistent with the plant specific ifethodology.
of a 24 month The Frequency is based upon the assumption of the magnitude calibration interval in the determination of equipment drift in the setpoint analysis.
SR 3.3.4.1.5 the The LOGIC SYSTEM FUNCTIONAL TEST demonstratesfor a specific OPERABILITY of the required trip logic the pump breakers is channel. The system functional test of overlaps the LOGIC included as part of this Surveillance and testing of the SYSTEM FUNCTIONAL TEST to provide complete a breaker is assumed safety function. Therefore, if instrument channels incapable of operating, the associated would be inoperable.
to perform this The 24 month Frequency is based on the need during a plant Surveillance under the conditions that apply unplanned transient if the outage and the potential for an at power.
Surveillance were performed with the reactor usually pass Operating experience has shown these components month Frequency.
the Surveillance when performed at the 24 REFERENCES 1. UFSAR, Figure 7.4-9 Reactor Recirculation System (FCD).
- 3. Drawing 11825-5.01-15D, Rev. D, Reactor Assembly Nuclear Boiler, (GE Drawing 919D690BD).
GE-NE
- 4. "ATWS Overpressure Analysis for FitzPatrick,"
A42-00137-2-01, March 2000.
(continued)
B 3.3-95 Revision K JAFNPP
ATWS - RPT Instrumentation B 3.3.4.1 BASES REFERENCES (continued)
- 5. GENE-770-06-1-A, Bases for Changes To Surveillance Test Intervals And Allowed Out-of-Service Times for I&~
Selected Instrumentation Technical Specifications, December 1992.
B 3.3-96 Revision K JAFNPP
JAFNPP 33
& - n 2 (1)Reactor LOWWater LOW (Sao8 a] 1 (25) Reactor HIig ProssuUFS Ar L P- -~(StIvtdmn Cooin Isolation)
I'8 IM ~ )1 2 (If) Reactor .LOW-LOWLo WaterLeveli 2 112 ,W111(g Drywell HIPh Pressure (Notes 4 &Mr421ý27 Psig el H~g Prsue\-L s2.7 psig
]2 (6)
Steam V2n <3 x Noimal Rated ser Limu Tw ivAý uj power eaccgrun' 2main High RditOOM 6]2Main Steam l~hn oW Pressure 9125 p6Ig 1 6]~ 2 (Pal CN511 er 2mSL MaIn Steam Line Hig Flow L ree 038j(10) M SamLm
~'ii) (1)Eq~m6~ Area High Tempetatu' A7 22 C(lid HI __A 69L- ~ )Y(7 o Amendiment No. 227 11 REVISIONr- 1<
DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - MORE RESTRICTIVE M3 Not Used.
M4 CTS 4.2.A spdtifies that the main steam isolation valve (MSIV)-actuation must be instrumentation response time for the specified trip functions Each test shall demonstrated to be within its limit once per 24 months.
channels in both include at least one channel in each trip system. All In ITS trip systems shall be tested within two test intervals. test must be SR 3.3.6.1.8 the ISOLATION INSTRUMENTATION RESPONSE TIME The Note for this performed every 24 months on a STAGGERED TEST BASIS. of determining SR specifies that "n" equals 2 channels for the purpose will the STAGGERED TEST BASIS Frequency. Therefore, SR 3.3.6.1.8 to be tested in two require all channels requiring response time testing is more restrictive since two (2) surveillance intervals. This change l.a and 1.b (2) channels must be tested each interval for Functions 1.c instead while 8 channels must be tested each interval for Function This change of one channel in each trip system required by the CTS.
each interval to will ensure a sufficient number of channels are tested identify any significant response time degradation.
M5 Not Used.
system in CTS M6 The required number of OPERABLE channels in each trip and HPCI and RCIC Table 3.2-1 for HPCI and RCIC Steam Line Low Pressure Turbine High Exhaust Diaphragm Pressure Functions (proposed Functions 3.b, 4.b, 3.c and 4.c for Table 3.3.6.1-1) are proposed to be increased receive inputs from 1 to 2. The two trip systems for these Functions the associated from two channels, both of which must trip to isolate system. The valve(s), yielding a two-out-of-two logic for each trip a more increase in channels required to be OPERABLE constitutes instrument restrictive change and is necessary to ensure no single failure can preclude the isolation function.
in cold shutdown M7 CTS Table 3.2-1, Note 3.A requires the reactor to be associated with within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the ACTIONS or Completions Times be satisfied.
inoperable Primary Containment instrumentation cannot ITS 3.3.6.1 Required These requirements are proposed to be replaced by main steam line Actions D.2.1 (for isolation Functions associated with with primary isolation) and H.1 (for isolation Functions associated MODE 3 within 12 containment isolation) which require the plant be in 3.3.6.1 Required hours under the same conditions. In addition, ITS Action D.2.2 and H.2 requires the plant to be in MODE 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (L11). This change is more restrictive because it provides an additional 3 in 12 hours. The allowed requirement to place the plant in MODE are reasonable, based Completion Times in Required Action D.2.1 and H.1 conditions from on operating experience, to reach the required plant Page 7 of 25 Revision K JAFNPP
DISCUSSION OF CHANGES ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L12 (continued) a plant transient Operable. This extra time reduces the potential for is consistent with that could challenge safety systems. This change NUREG-1433, Revision 1.
L13 Not used.
those portions L14 The details in CTS Tables 4.1-1 and 4.1-2, that identify testing (trip channel of the instrument channel which require functional pressure source),
and alarm) and the method of calibration (standard respectively, are proposed to be deleted. This information is not for Channel Functional Test necessary because the proposed definitions This change is and Channel Calibration provide the necessary guidance.
consistent with NUREG-1433, Revision 1.
L15 Not used.
Value (A16) of L16 This change replaces the Trip Level Setting or Allowable dP for the HPCI water
- 160 inches of water dP to : 168.24 inches of 3.3.6.1 Function 3.a).
(ITS Turbine Steam Line High Flow trip function Technical Specifications)
The Allowable Values (to be included in the have been and the Trip Setpoints (to be included in plant procedures)
Standards Manual, IES established consistent with the NYPA Engineering Methodology."
3A, "Instrument Loop Accuracy and Setpoint Calculation Values" are consistent The methodology used to determine the "Allowable Part II, with the methodology discussed in ISA-S67.04-1994, for Nuclear Safety "Methodologies for the Determination of Setpoints analysis limits, Related Instrumentation." Any changes to the safety and confirmed as ensuring applied in the methodologies, were evaluated Revision K JAFNPP Page 22 of 25
NO SIGNIFICANT HAZARDS CONSIDERATIONS INSTRUMENTATION ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION TECHNICAL CHANGES LESS RESTRICTIVE (SPECIFIC)
L13 CHANGE Not used.
Revision K Page 19 of 32 JAFNPP
NO SIGNIFICANT HAZARDS CONSIDERATIONS ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION TECHNICAL CHANGES - LESS RESTRICTIVE (SPECIFIC)
L13 CHANGE (continued)
Not used.
This page intentionally blank Revision K JAFNPP Page 20 of 32
primary containment isolation instrumentation 3.3.6.1 Tai. 3.3.6.1,1 (pass I of 61 inatunntatiaI' Pviinr-y Cantaif'? eaokation APPLICABLE UPT~M oNMSa REUIRIEDS Msii DIL* VIIALUE SpecIFIE PER TRIP .. JI.UUTS A MuTTIO SUIW ACTION W.
pijICION 4.
I.* i SteM LOWI's1LatiU'
[rT7.(q]) OR 3.3.6.1.1()
LO Lo 3.3.6.1.
-11,Lo LW~
S S 3.3.6A.1S 3.3.6-1.1w6 e~s~l
- b. *&in $to= LiI's I Presaur - LAM Id ~~3.3.6.1.21Se IT SR MOR3.3.6.1.2j 3..
EU 3.3.6.1-1 LOW' ,t,2,3 00NILper U 3.3.6 1AA Main Ste 3.3.6:1:9
. 0 NO vamflfl Candsir V~WAU - Low 3C&
('7g sU 3.3.6.1.7 2( o SR 3.3.6.1 0 1 n~i. atmlwwI~i UR 3.3.6.1 1.2.3
~33.3.6.
~~i~tui - N gMI L 7 1()
sk 3.36. .7 3.3.6.1.1 Iq1 oD S3.3.6.1.
U 3.3.6.1.
CT(
CTa)M]
REVISIONrK R
INSERT Functions 2.d, 2.e, 2.f, 2.g CODP-(-
2(c) F SR 3.3.6.1.1 2.7 psig
- r3. Z-(faJ d. Drywell Pressure - 1.2,3 High SR 3.3.6.1.4 T3.2- (1 SR 3.3.6.1.5
_ _ _ _ __ j _ _ jSR 3.3.6.1.7 [ _ _
2 9 3.3.6.1.1 x IS inches Reactor Vessel ater 1,2.3 T 3.240(J3 e.
Level - Low Low Low (Level
_ 1) 1SSR
_evel 1SR SR 3.3.6.1.2 SR 3.3.6.1.4 3.3.6.1.5 3.3.6.1.7 . _
2 F SR 3.3.6.1.1 3 times T'3.,1-((7)] ain Steam Line m.l 1,2.3 Normal Full SR 3.3.6.1.3 Radiation - High SR 3.3.6.1.6 Power SIc SR 3.3.6.1.7 SR33... 1 Background 7 i nches SR 3.3.6.1. 177 1,2.3 2(c)
Reactor Vessel Water 3.z-I () g.
3)
Level - Low (Level I F'SR 1 SR 3..6 ISR 1.
3.3.6.1.4 3.3.6.1.7 Insert Page 3.3-58 Revision K
1 JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION ITS: 3.3.6.1 - PRIMARY CONTAINMENT ISOLATION INSTRUMENTATION PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)
DB6 (continued)
Function Line Penetration (Drywell Entrance) Area Temperature-High; Function 5.c, RWC 5.a, Suction Line Penetration Area Temperature-High; 5.f, Drywell Heat Exchanger Area Temperature-High; and Function added for those Pressure-High. Functions 2.d and 2.g have been penetration flow Functions which include only one trip system to certain (c) was added to Table paths to simplify the Required Actions. Footnote Notes have been 3.3.6.1-1 to identify these Functions. Subsequent have been renumbered, where applicable. Subsequent Functions renumbered, as required.
Table 3.3.6.1-1 since DB7 This change deletes various ITS Functions from the l.f. Main Steam Tunnel they are not included in the design: Function Function 1.g, Turbine Building Area Differential Temperature-High; Exhaust Temperature-High; Function 2.e, Refueling Floor Pressure-High; Function Radiation-High; Functions 3.d and 4.d, Drywell Temperature-Time Delay 3.g and 4.f. HPCI and RCIC Suppression Pool Area Pool Area Relays; Functions 3.h and 4.g, HPCI and RCIC Suppression4.h, Emergency Area Differential Temperature-High; Function 3.i and Room Differential Cooler Temperature-High; Function 4.j, RCIC Equipment and Function 5.c Temperature-High; Function 5.a Differential Flow-High Subsequent Functions Area Ventilation Differential Temperature-High.
have been renumbered. as required.
for ITS Function DB8 The correct trip level Function has been incorporated design.
3.3.6.1 Function 5.e in accordance with the JAFNPP to identify the valves DB9 ITS Table 3.3.6.1-1 Footnote Cd) has been revised JAFNPP design.
isolated by the Function consistent with the specific value or DB1O The brackets have been removed and the proper plant requirements incorporated.
Function (ITS DB11 This change separates the RWC Pump Area Temperature-High (Pump Room A and 3.3.6.1 Function 5.b) Allowable Value into two areasare different.
Pump Room B) since the proposed "Allowable Values" Revision K JAFNPP Page 4 of 5
0 INSERT ASA-2 removal of heat from the In addition, the setting is low enough to allow the prevent isolation on a reactor for a predetermined time following a scram, to the safety/relief valves partial loss of feedwater and to reduce challengesa level of water (S/RVs). The Allowable Value is referenced from bottom of the RPV and also 352.56 inches above the lowest point in the inside(Ref. 13).
corresponds to the top of a 144 inch fuel column Insert Page B 3.3-158 Revision K
INSERT Function 1.f l.f. Main Steam Line Radiation- High has been removed from the The Main Steam Line Radiation-High isolation signal this isolation Function has MSIV isolation logic circuitry (Ref. 1); however, other valves discussed under been retained for the MSL drains valves (and utilized to determine that Function 2.f) to ensure that the assumptions (CRDA) are acceptable offsite C"doses resulting from a control rod drop accident maintained.
generated from four radiation Main Steam Line Radiation-High signals are located near the main steam lines elements and associated monitors, which are channels of the Main Steam Line in the steam tunnel. Four instrumentation to be OPERABLE to ensure Radiation-High Function are available and required the isolation function.
that no single instrument failure can preclude that a high radiation trip The Allowable Value was selected to be low enough the CRDA. In addition, the results from the fission products released in radiation level in the setting is adjusted high enough above the background trips are avoided at rated vicinity of the main steam lines so that spurious power.
This Function isolates the MSL drain valves.
Insert Page B 3.3-161 Revision K
60 INSERT Functions 2.e and 2.f (continued) 2.f. Main Steam Line Radiation - High signal has been removed from The Main Steam Line Radiation - High isolation 1); however, this isolation Function the MSIV isolation logic circuitry (Ref. loop sample valves to ensure that the has been retained for the recirculation acceptable offsite doses resulting from assumptions utilized to determine that a CRDA are maintained.
are generated from four radiation Main Steam Line Radiation - High signals are located near the main steam lines elements and associated monitors, which channels of the Main Steam Line in the steam tunnel. Four instrumentation and required to be OPERABLE to ensure Radiation - High Function are availablepreclude the isolation function.
that no single instrument failure can enough that a high radiation trip The Allowable Value was selected to be low in the Design Basis CRDA. In results from the fission products released above the background radiation addition, the setting is adjusted high enough lines so that spurious trips are level in the vicinity of the main steam avoided at rated power.
loop sample valves.
This Function isolates the recirculation Insert Page B 3.3-164b Revision K
Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 1 of 6)
Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM PER TRIP REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED VALUE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS FUNCTION
- 1. Main Steam Line Isolation 1.2.3 2 SR 3.3.6.1.1 z 18 inches
- a. Reactor Vessel Water SR 3.3.6.1.2 Level - Low Low Low SR 3.3.6.1.4 (Level 1) SR 3.3.6.1.5 SR 3.3.6.1.7 SR 3.3.6.1.8 1 SR 3.3.6.1.1 S825 psig
- b. Main Steam Line SR 3.3.6.1.2 Pressure - Low SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 SR 3.3.6.1.8 2MSL per SR 3.3.6.1.1 S125.9 psid
- c. Main Steam Line 1.2.3 Flow - High SR 3.3.6.1.2 SR 3.3.6.1.4 3.3.6.1.5 SR 3.3.6.1.7 SR 3.3.6.1.8 SR 2 D SR 3.3.6.1.1 3.3.6.1.2 ý 8vacuum Hg inches
- d. Condenser Vacuum- Low 1.
SR 3.3.6.1.4 2 (a), 3 (a) 3.3.6.1.5 SR SR 3.3.6.1.7 D SR 3.3.6.1.1 : 195°F 1.2,3 8 SR 3.3.6.1.2
- e. Main Steam Tunnel Area SR 3.3.6.1.4 Temperature - High 3.3.6.1.5 SR SR 3.3.6.1.7 IAK F SR 3.3.6.1.1 1.2.3 2 SR 3.3.6.1.3 S3 times Normal Full
- f. Main Steam Line 3.3.6.1.6 Power Radiation - High SR SR 3.3.6.1.7 Background (continued)
(a) With any turbine stop valve not closed.
(b) Not used. 19L 3.3-52 Amendment (Rev. K)
Primary Containment Isolation Instrumentation 3.3.6.1 Table 3.3.6.1-1 (page 2 of 6)
Primary Containment Isolation Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM PER TRIP REQUIRED SURVEILLANCE ALLOWABLE SPECIFIED VALUE CONDITIONS SYSTEM ACTION C.1 REQUIREMENTS FUNCTION
- 2. Primary Containment Isolation H SR SR 3.3.6.1.1 3.3.6.1.2 S177 inches Reactor Vessel Water 1.2,3 2
- a. SR 3.3.6.1.4 Level - Low (Level 3) SR 3.3.6.1.5 SR 3.3.6.1.7 2 H SR SR 3.3.6.1.1 3.3.6.1.2 S2.7 psig
- b. Drywell Pressure-High 1.2.3 SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 F SR 3.3.6.1.1 s 450 R/hr 1,2.3 1 SR 3.3.6.1.2
- c. Containment 3.3.6.1.5 Radiation - High SR SR 3.3.6.1.7 F SR 3.3.6.1.1 f 2.7 psig 1.2.3 SR 3.3.6.1.2
- d. Drywell Pressure-High SR 3.3.6.1.4 SR 3.3.6.1.5 SR 3.3.6.1.7 F SR 3.3.6.1.1 ý 18 inches 1,2,3 2 SR 3.3.6.1.2
- e. Reactor Vessel Water SR 3.3.6.1.4 Level - Low Low Low 3.3.6.1.5 (Level 1) SR SR 3.3.6.1.7
- f. Main Steam Line 1.2.3 2 F SR SR 3.3.6.1.1 3.3.6.1.3 3.3.6.1.6 S3 times Normal Power Full IL Radiation - High 3.3.6.1.7 Background SR SR 3.3.6.1.1 k 177 inches F
1.2,3 2 (c) SR 3.3.6.1.2
- g. Reactor Vessel Water 3.3.6.1.4 Level - Low (Level 3)
SR SR 3.3.6.1.5 SR 3.3.6.1.7 (b) Not used.
(continued)
I&~
(c) Only one trip system provided for each associated penetration.
3.3-53 Amendment (Rev. K)
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES 1.f. Main Steam Line Radiation-High (continued)
APPLICABLE SAFETY ANALYSES. to LCO, and Function 2.f) to ensure that the assumptions utilized from a determine that acceptable offsite doses resulting APPLICABILITY (CRDA) are maintained.
control rod drop accident from Main Steam Line Radiation-High signals are generated are which four radiation elements and associated monitors,tunnel. Four located near the main steam lines in the steam instrumentation channels of the Main Steam Line to be Radiation-High Function are available and required can OPERABLE to ensure that no single instrument failure preclude the isolation function.
that a The Allowable Value was selected to be low enough high radiation trip results from the fission products adjusted released in the CRDA. In addition, the setting isin the high enough above the background radiation level trips are vicinity of the main steam lines so that spurious avoided at rated power.
This Function isolates the MSL drain valves.
Primary Containment Isolation 2.a. 2.g. Reactor Vessel Water Level- Low (Level 3) to cool Low RPV water level indicates that the capability penetrations the fuel may be threatened. The valves whose to communicate with the primary containment are isolated of the The isolation limit the release of fission products. actions to ensure primary containment on Level 3 supports (continued)
B 3.3-160 Revision K JAFNPP
Primary Containment Isolation Instrumentation B 3.3.6.1 BASES 2.f. Main Steam Line Radiation-High (continued)
APPLICABLE SAFETY ANALYSES, are generated from LCO, and Main Steam Line Radiation-High signals monitors, which are APPLICABILITY four radiation elements and associated Four the steam tunnel.
located near the main steam lines in finstrumentation channels of the Main Steam Line to be Radiation-High Function are available and required failure can OPERABLE to ensure that no single instrument preclude the isolation function.
that a The Allowable Value was selected to be low enough high radiation trip results from the fission products the setting released in the Design Basis CRDA. In addition, radiation level is adjusted high enough above the backgroundthat spurious in the vicinity of the main steam lines so trips are avoided at rated power.
sample valves.
This Function isolates the recirculation loop Core Isolation High Pressure Coolant Injection and Reactor Cooling Systems Isolation 3.a, 4.a. HPCI and RCIC Steam Line Flow-High to detect a Steam Line Flow-High Functions are provided initiate closure break of the RCIC or HPCI steam lines and of the steam line isolation valves of the appropriate flowing out of system. If the steam is allowed to continue and the core can the break, the reactor will depressurize initiated on high uncover. Therefore, the isolations are The isolation flow to prevent or minimize core damage. of the RPS, ensures action, along with the scram function (continued)
Io
'_1rA Revision K JAFNPP D *. * - J.U"r
K
SUMMARY
OF CHANGES TO ITS SECTION 3.8 - REVISION Affected Pages Source of Change Summary of Change Section 3.8.6 Misc. editorial These editorial changes were identified during the corrections preparation of the final ITS submittal: ITS Bases mark-up, p 3.8-66 B.1 and Bases JFD DB1 to be Revise ITS Bases Action consistent with the changes made in Revision J (battery Bases JFD DB1 (p 1 of 3) electrolyte temperature limit was revised from 60 degrees F to 65 degrees F). Retyped ITS Bases p 3.8-60
Battery Cell Parameters B 3.8.6 BASES ACTIONS A.1. A.2. and A.3 (continued)
Continued operation is ohly permitted forto 31 days btfore battery cell parameters must be restored within Taking into consideration that, Category A and B limits. capacity while battery capacity is degraded, sufficient allow time to exists to perform the intended function andto tonormal limits, fully restore the battery cell parameters the this tim is acceptable for operation prior to declaring DC batteries inoperable.
La When any battery pavmeter is outside the Category C limit cell, sufficient capacity to supply the for any connected and the maximum expected load requirement subsystem is not ensured corresponding DC electrical power must be ddeclan Additionally, other 1a__Y'23&MW !
conditioL...stch as taeu Copeto Required ActionS averay o-Lý S*Iplotion lll"Aar average temperature of representative cells also are cause for imediately declaring the XC electrical power subsystem inerl. 4 es.7,4o .
ru-sZ 'WVt' voc'.a-C A. V jvos'4. ?JA SURVEILLANCE REQUIWMTS cell parameters are This SR verifies that Category :Abattery consistent with IEEE-0*4 (Ref. , which recomends regular battery inspections (at least one per month) including voltage, specific gravity, and electrolyte temperature of C41nt palls ets I t sliits.
Iae Trasients, tack as motor s, ring ie ch wy m ntarily c so batted vol age t vop to lll03 V, do not constitut a battery dis, -n.v vided he ba yoinal Vol++ an lnoat (continued)
B 3.846 Rev 1. 04/07125 SUR/4 STS
JUSTIFICATION FOR DIFFERENCES FROM NUREG-1433, REVISION 1 ITS BASES: 3.8.6 - BATTERY CELL PARAMETERS RETENTION OF EXISTING REQUIREMENT (CLB)
SR 3.8.6.2 is revised to omit the Frequencies of "Once within after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> CLB1 a after a battery discharge < 110 V" and "Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> battery overcharge > 150 V" since no-similar CTS Surveillancef a battery Requirement exists at JAFNPP. The Frequencies associated with with discharge or overcharge are omitted since, they are inconsistent do the content of typical STS Surveillances, revised ISTS Surveillances not typically contain "abnormal condition" related frequencies and, battery discharge or overcharge are adequately covered by administrative controls. In addition, this change is currently submitted as aand is Technical Specification Task Force Change Traveler, TSTF-201, pending.
(PA)
PLANT-SPECIFIC WORDING PREFERENCE OR MINOR EDITORIAL IMPROVEMENT PAl Changes have been made (additions, deletions, and/or changes to the NUREG) to reflect the plant specific system/structure/component nomenclature, equipment identification or description.
PA2 Battery Cell Parameters support the operation of the DC electrical power to subsystems and the Battery Cell Parameter Specification is required "DC be applicable during the same MODES and conditions as in LCO 3.8.4, Sources-Operating," and LCO 3.8.5, "DC Sources-Shutdown." The LCO same safety analyses discussions as those discussed in the Bases for 3.8.4 and LCO 3.8.5 are also applicable to the Battery Cell Parameter Specification. As a result, the Bases for the Battery Cell Parameter Specification in the Applicable Safety Analyses Section have been revised accordingly.
a PA3 Editorial changes have been made for enhanced clarity or to correct grammatical/typographical error.
PLANT-SPECIFIC DIFFERENCE IN THE DESIGN (DB)
DB1 ITS 3.8.6.3 Condition B.1 has been revised to reflect specific JAFNPP requirements of, ! 65°F for 125 VDC batteries and ! 50°F for 419 VDC LPCI MOV independent power supply batteries based on JAF Electrical Calculations.
DB2 ITS 3.8.6 has been revised to reflect the specific JAFNPP requirements of, UFSAR Chapter 6, Emergency Core Cooling System.
DB3 ITS 3.8.6 has been revised to reflect the specific JAFNPP requirements Page 1 of 3 Revision K JAFNPP
Battery Cell Parameters B 3.8.6 BASES ACTIONS A.1, A.2, and A.3 (continued) initial verification because specific gravity measurements must be obtained for each connected cell. Taking into consideration both the time required to perform the required Verification and the assurance that the battery cell parameters are not severely degraded, this time is at 7 considered reasonable. The verification is repeatedCategory day intervals until the parameters are restored consistentto A and B limits. This periodic verification is with the guidance provided in IEEE-450 (Ref. 4) of (not to monitoring battery conditions at regular intervals exceed one week) while completing corrective actions.
Continued operation is only permitted for 31 days before battery cell parameters must be restored to within that, Category A and B limits. Taking into consideration while battery capacity is degraded, sufficient capacity time to exists to perform the inteded function and to allow limits, fully restore the battery cell parameters to normal the this time is acceptable for operation prior to declaring DC batteries inoperable.
B.1 C limit When any battery parameter is outside the Category the for any connected cell, sufficient capacity to supply load requirement is not ensured and the maximum expected corresponding DC electrical power subsystem must be declared inoperable. Additionally, other potential conditions, such as any Required Action of Condition A and associated Completion Time not met, or average electrolyte temperature of representative cells < 65°F for each 125 VDC battery, or MOV independent power supply
< 50°F for each 419 VDC LPCI the battery, also are cause for immediately declaring associated DC electrical power subsystem inoperable.
SURVEILLANCE SR 3.8.6.1 REQUIREMENTS are This SR verifies that Category A battery cell parameters regular consistent with IEEE-450 (Ref. 4), which recommends battery inspections (at least one per month) including (continued)
B 3.8-60 Revision K JAFNPP
Entergy Nuclear Northeast Entergy Nuclear Operations, Inc.
James A. FitzPatrick NPP Entergy( P-o. Box 110 Lycoming, NY 13093 Tel 315 349 6024 Fax 315 349 6480 T. A. Sullivan Vice President, uiperations-JAF April 26, 2002 JAFP-02-0098 United States Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop O-P1-17 Washington, D.C. 20555
Subject:
James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59 Revision J to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications
References:
see last page of letter
Dear Sir,
This letter and the associated attachments provides Revision J to the previously submitted application for amendment to the James A. FitzPatrick Technical Specifications (Reference 1),
as supplemented by References 2, 3, 4, and 5 for converting the current Technical Specifications (CTS) to the Improved Technical Specifications (ITS) consistent with the Improved Standard Technical Specifications (NUREG-1433, Revision 1).
Revision J (Attachment 1) to the Reference 1, 2, 3, 4 , and 5 submittals include: certain Technical Specification Task Force Traveler related changes; a change to close out a remaining NRC question; numerous typographical, editorial, and consistency corrections; changes due to the engineering analysis performed as discussed in Reference 6; and a few new additional changes. Each Chapter/Section includes a summary of the changes to the associated Chapter/Section (with the exception of the Split Report, whose summary for the change is included in the Summary of Changes to Section 3.7).
The Insert and Discard Instructions are included in Attachment 2 to allow merging Revision J with the existing submittal. The clean typed ITS and Bases in Volumes 2, 3, and 4, and the CTS markup pages in CTS order in Volume 5 are not being updated since these Volumes are duplicates of each individual Specification located in Volumes 6 through 19.
We request that you approve the James A. FitzPatrick ITS no later than July 31, 2002.
United States Nuclear Regulatory Commission Attn: Document Control Desk
Subject:
Revision J to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications Page -2 There are no new commitments contained in this letter. Should you have any questions, please contact Mr. Andrew Halliday at (315) 349-6055.
Very Truly Your Vice President, Operations - JAF Attachments: 1) Revision J to the JAF ITS Submittal
- 2) Insert and Discard Instructions cc:
Regional Administrator Mr. N. B. Le U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission Mail Stop O-7H3 475 Allendale Road Washington, DC 20555 P. 0. Box 134 King of Prussia, PA 19406 Resident Inspector's Office Mr. Guy Vissing, Project Manager James A. FitzPatrick Nuclear Power Plant Project Directorate I U. S. Nuclear Regulatory Commission Division of Licensing Project Management P. 0. Box 134 U. S. Nuclear Regulatory Commission Lycoming, NY 13093 Mail Stop 8C2 Washington, DC 20555 Mr. William M. Flynn New York State Energy Research and Development Authority Corporate Plaza West 286 Washington Avenue Extension Albany, New York 12203-6399 Mr. Paul Eddy NYS Department of Public Service 3 Empire Plaza Albany, New York 12223 Mr. William D. Beckner, Chief Technical Specifications Branch U. S. Nuclear Regulatory Commission Mail Stop O-7H3 Washington, DC 20555
United States Nuclear Regulatory Commission Attn: Document Control Desk
Subject:
Revision J to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications Page -3
References:
- 1. NYPA letter, J. Knubel to USNRC Document Control Desk, Proposed Technical SpecificatiorL.Change (License Amendment) Conversion to Improved Standard Technical Specifications (JPN-99-008), dated March 31, 1999 (TAC No. MA5049)
- 2. NYPA letter, J. Knubel to USNRC Document Control Desk, Revision B to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JPN-99-018), dated June 1, 1999
- 3. NYPA letter, Michael J. Colomb to USNRC Document Control Desk, Revision C to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JAFP-99-0278), dated October 14, 1999
- 4. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, Revisions D, E, F, G, and H to Proposed Technical Specification Change (License Amendment) Conversion to Improved Standard Technical Specifications (JAFP-01 0133), dated May 31, 2001
- 5. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, Revision I to Proposed Technical Specification Change (License Amendment)
Conversion to Improved Standard Technical Specifications (JAFP-01-0234), dated October 18, 2001
- 6. Entergy Nuclear Northeast letter, T. A. Sullivan to USNRC Document Control Desk, James A. FitzPatrick (JAF) Improved Technical Specifications (ITS) Submittal (JAFP 02-0029), dated February 6, 2002
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of )
Entergy Nuclear Operations, Inc. ) Docket No. 50-333 James A. FitzPatrick Nuclear Power Plant )
APPLICATION FOR AMENDMENT TO OPERATING I ICENSE Entergy Nuclear Operations, Inc. requests an amendment to the Technical Spesifications (TS) contained in Appendix A to Facility Operating License DPR-59 for the James A. FitzPatrick Nuclear Power Plant. This application is filed in accordance with Section 10 CFR 50.90 of the Nuclear Regulatory Commission's regulations.
This application for amendment to the FitzPatrick Technical Specifications proposes to convert the FitzPatrick current Technical Specifications (CTS) to be consistent with the Improved Standard Technical Specifications (ISTS) in NUREG-1433, Revision 1, dated April 1995. The proposed license amendment request was prepared considering the guidance of Nuclear Energy Institute (NEI) NEI 96-06, "Improved Technical Specifications Conversion Guidance,"
dated August 1996.
The Proposed license amendment request to convert the FitzPatrick CTS to the FitzPatrick Improved Technical Specifications (ITS) is enclosed with this application.
Entergy Nuclear Operations, Inc. STATE OF NEW YORK COUNTY OF OSWEGO Subscribed and ýworn to before me this I-. day oft 2002.
P ub c Vice President, Operations-JAF
~~...............
"*".-**.'." *