ML20079P706
| ML20079P706 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/19/1983 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML19289B797 | List: |
| References | |
| TAC-54169, TAC-54433, TAC-54607, NUDOCS 8401310314 | |
| Download: ML20079P706 (109) | |
Text
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p_ GENERAL ELECTRIC C0MPANY llJ AFFIDAVIT I, R. Artigas, being duly sworn, depose and state as follows:
- 1. I am Manager, BWR Project Licensing, Safety and Licensing Operation, -
General Electric Company, and have been delegated the function of reviewing the information described in paragraph 2 which is sought to be withheld and have been authorized to apply for its withholding.
- 2. " Analog Trip System for Engineered Safeguard Sensor Trip Inputs -
Edwin I. Hatch Nuclear Plant Units 1 and 2", NEDE-22154-1, Rev. 1,
- July 1983.
- 3. In designating material as proprietary, General Electric utilizes the definition of proprietary information and trade secrets set forth in the American Law Institute's Restatement Of Torts, Section 757.
This definition provides:
"A trade secret may consist of any formula, pattern, device or compilation of information which is used in one's business and
, which gives him an opportunity to obtain an advantage over competitors who do not know or use it.... A substantial LT]b element of secrecy must exist, so that, except by the use of improper means, there would be difficulty in acquiring informa-tion.... Some factors to be considered in determining whether given information is one's trade secret are: (1) the extent to which the information is known outside of his business; (2) the extent to which it is known by employees and others involved in his business; (3).the extent of measures taken by him to guard the secrecy of the information; (4) the value of the information to him and to his competitors; (5) the amount of effort or money expended by him in developing the information; (6) the ease or difficulty with which the information could be properly acquired or duplicatsd by others."
- 4. Some examples of categories of information which fit into the definition of proprietary information are:
- a. Information that discloses a process, method or apparatus where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;
- b. Information consisting cf supporting data and analyses, includ-ing test data, relative to a process. method or apparatus, the application of whicM provide a competitive economic advantage, l%[,)\ e.g., by optimization or improved marketability; c
8401310314 840123 gDRADOCK 05000321 PDR AJW: rm/A081919 '
8/19/83 ,
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p c. Information which if used by a competitor, would reduce his
() expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality or licensing of a similar product;
- d. Information which reveals cost or price information, production capacities, budget levels or commercial strategies of General Electric, its customers or suppliers;
- e. Information which reveals aspects of past, present or future General Electric customer-funded development plans and programs of potential commercial value to General Electric;
- f. Information which discloses patentable subject matter for which
. it may be desirable to obtain patent protection;
- g. Information which General Electric must treat as proprietary a; cording to agreements with other parties.
- 5. In addition to proprietary treatment given to material meeting the standards enumerated above, General Electric customarily maintains in confidence preliminary and draft material which has not been subject to complete proprietary, technical and editorial review.
This practica is based on the fact that draft documents often do not appropriately reflect all aspects of a problem, may contain tentative conclusions and may contain errors that can be corrected during O' normal review and approval procedures. Also, until the final document is completed it may not be possible to make any definitive determination as to its proprietary nature. General Electric is not generally willing to release such a document to the general public in such a preliminary form. Such documents are, however, on occasion furnished to the NRC staff on a confidential basis because it is General Electric's belief that it is in the public interest for the staff to be promptly furnished with significant or potentially significant information. Furnishing the document on a confidential.
basis pending completion of General Electric's internal review permits early acquaintance of the staff with the information while protecting General Electric's potential proprietary position and permitting General Electric to insure the public documents are technically accurate and correct.
- 6. Initial approval of proprietary treatment of a document is made by the Subsection Manager of the originating component, the man most i
likely to be acquainted with ti)e value and sensitivity of the information in relation to industry knowledge. Access to such documents within the Company is limited on a "need to know" basis and such documents at all times are clearly identified as proprietary.
- 7. The procedure for approval of external release of such a document is reviewed by the Section Manager, Project Manager, Principal Scientist or other equivalent authority, by the Section Manage _r of the cognizant Marketing function (or his delegate) and by the Legal Operation for AJW:rm/A081319 8/19/83
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<3 technical content, competitive effect and determination of the b accuracy of the proprietary designation in accordance with the standards enumerated above. Disclosures outside General Electric are generally limited to regulatory bodies, customers and potential customers and their agents, suppliers and licensees only in accordance with appropriate regulatory provisions or proprietary agreements.
- 8. The document mentioned in paragraph 2 above has been evaluated in accordance with the above criteria and procedures and has been found to contain information which is proprietary and which is customarily held in confidence by General Electric.
- 9. The information contained herein is the result of extensive analyses performed at considerable cost to the General Electric Company. The development and verification of these methods, as well as their application and execution cost in excess of $2 million.
STATE OF CALIFORNIA COUNTY OF SANTA CLARA
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R. Artigas, being duly sworn, deposes and says:
That he has read the foregoing affidavit and tne matters stated therein
, are true and correct to the best of his knowledge, information, and belief.
A i Executed at San Jose, California, this day of [ ,
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1 R. Artiga g
[4 O General Electric Company Subscribed and sworn before me thisgday of AdAt/5Y 198@
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4 TABLE OF CONTENTS
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I. Introduction II. References III. Analog Transmitter Trip System (ATTS) Installation ,
3A. ATTS Detailed Design Information 3B. Bases for. Technical Specifications Rcvisions of ATTS Equipment 3C. Summary IV. Proposed Plant Modifications to be Incorporated into the ATTS Design 4A. Introduction 4B. Safety Evaluation Report 4B.1 Reactor Water Low Low (Level 2) Trip Setpoint Modification 4B.2 Deletion of High Drywell Pressure Signal for RHR (Shutdown Cooling Mode), RPV Head Spray Valves, and RWCU Isolation 4B.3 Lowered Water Level Trip Setpoint for Isolation of Reactor Water Cleanup and Secondary Containment, and Starting of Standby Gas Treatment System 4B.4 Deletion of Ambient Temperature Loops in the Leak Detection System s
4B.5 Deletion of Drywell Pressure Sensors E11-N011A, B, C, D g 4B.6 Trip Setpoint/ Allowable Value Setpoint Modifications 4B.7 Reactor Vessel Water Level - High (Level 8) Trip Instrumentation Modifications 4B.8 Elimination of the Reactor Pressure Permissive to the Bypass of the MSIV Closure Signal Due to Low Condenser Vacuum V. Summary VI. Proposed Technical Specifications Revisions Appendix 1 - Significant Hazards Review Appendix 2 - NEDE-22154-1, Analog Trip System for Engineered Safeguard Sensor
- Trip Inputs - Edwin I. Hatch Nuclear Plant Units 1 and 2 l
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T I. INTRODUCTION
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The Edwin I. Hatch Nuclear Plant-Unit 2 (HNP-2) safety-related system instru-mentation upgrade program, which includes incorporation of the analog transmitter trip system (ATTS), comprises many equipment modifications and installations to meet many of the requirements and criteria of:
e IE Bulletin 79-01B, Envir:nmental Qualification of Class IE Equipment *
- NUkEG-0737, TMI Lessons Learned
- e NUREG-0696, Safety Parameters Display System Interfaces
- This submittal covers information on'ATTS (Section III) and the proposed plant design modifications (Section IV) being performed at Plant Hatch through the installation of ATTS. The ATTS concept for use in the low low set system has been approved by the Nuclear Regulatory Commission (NRC) for Plant Hatch-Unit 2. This approval was granted by Amendment 33 to the Unit 2 Tech-nical Specifications.
The purpose of this submittal is to provide the bases for the proposed modi-fications and to provide Georgia Power Company's (GPC) proposed revisions to the Technical Specifications. Because of space limitations on some figures located throughout this document, the Unit 2 designator was not included as part of the MPL identifier; that is, instead of 2G11-N004, Gil-N004 was used.
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s-s Therefore, unless otherwise acted, all MPL identifiers are Unit 2 specific.
ATTS and the associated plant modificatious will be incorporated into the FNP-2 system logic during the next refueling outage. The installation is scheduled to be complete by the startup of Cycle 5 l
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- Subject shown is not necessarily title of document.
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l II. REFERENCES i
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- 1. Edwin I. Hatch Nuclear' Plant Unit 2, Docket ~No. 50-366, Proposed Plant Modification-Low Low Set Logic and Lowered MSIV Water Level, GPC Letter NED-83-108, Proposal for Technical Specifications Changes Which Support Cycle 4 Startup, February 23, 1983.
-2. IE Bulletin 79-01B, Environmental Qualification of Class IE Equipment.
- 3. IEEE 323-74, IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations. '
- 4. IEEE 344-75, IEEE Recommended Practices for Seismic Qcalification of IE Equipment.for Nuclear Power Generating Stations.
- 5. NEDO-21617-A, General Electric Licensing Topical Report, Analog Trans-mitter/ Trip Unit System for Engineered Safeguard Sensor Trip Inputs.
- 6. NUREG-0588, Revision 1, Interim Staff Position on Environmental Qualiti-cation of Safety-Related Electrical Equipment.
- 7. NUREG-0696, Functional Criteria for Emergency Response Facilities.
- 8. Regulatory Guide 1.105, Instrument Setpoints.
() 9. Draft Standard Technical Specifications for General Electric Boiling Water Reactor (GE-STS) - BWR/4.
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3 III. ANALOG TRANSMITTER TRIP SYSTEM (ATTS) INSTALLATION
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A safety evaluation report was prepared to evaluate the licensing requirements of ATTS. From that evaluation, it was concluded that the incorporation of ATTS constitutes neither an unreviewed safety question, nor a significant hazards consideration. The following detailed description provides all of the instrumentation to be incorporated into ATTS during the next refueling outage.
The general description of the proposed ATTS is provided in the NEDE-22154-1 General Electric (GE) Licensing Topical Report (Appendix 2). NEDE-22154-1 is a revision to NEDE-22154 which was provided to the NRC in Reference 1. The NEDE-221541 contains the system / component design changes and additional quali-fication test data. A description of qualification testing is documented in Chapter 4 of NEDE-22154-1, and in addition identifies Plant Hatch testing criteria. Actual qualification reports will be provided to the NRC upon request.
Since the equipment comprising the ATTS was demonstrtted to be superior to the mechanical switches currently used at Plant Hatch, certain technical speci-fications surveillance requirements may be revised. These revisions take advantage of the sensor improvements and the decreased drift which was demon-strated for this new equipment. Section III discusses the proposed surveil-lance revisions for the complete ATTS installation. The proposed plant modi-ficaticas associated with ATTS and their juacifications'are discussed in Section IV.
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( _) To keep the Nuclear Regulatory Commission (NRC) apprised of major Plant Hatch design modifications, Georgia Power Company (GPC) provided in Reference 1 a detailed description of the ATTS system to be installed. Since that time, several design modifications have occurred. This section, therefore, updates the information provided in Reference 1.
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3A. ATTS DETAILED DESIGN INFORMATION In NEDO-21617-A, dated December 1978, the NRC identified specific detailed information that each applicant (who uses that topical as a licensing basis) must provide relctive to the specific application of the ATTS hardware into h.s plant. Since NEDE-21154-1 (Appendix 2) describes the same conceptual system as NEDO-21617-A but with differ-eat hardware, GPC is supplying the NRC with the same information (requested of those who use NEDO-21617-A as a licensing basis) regarding the Plant Hatch installation. Therefore, this section pro-vides specific design criteria for the ATTS installation to be com-pleted during refueling at the end of Cycle 4 for Plant Hatch-Unit 2.
3A.1 As identified in the NRC request section of NED0-21617-A, GPC .
offers the information listed below.
3A.1.a The following information for each instrument loop that will be converted to the ATTS is provided in table 3.1:
i e Variable name e MPL number e Engineered safeguards division e Model number and vendor of the transmitter or RTD e Device's associated rack.
3A.1.b The layout of each card file in each trip unit cabinet, showing the trip variable for each card file slot, is provided in figures 3.1 through 3.8. These figures vere developed to illustrate the 12 card slots per card file and the three card files per cabinet arrange-
- ment.
3A.1.c The environmental service conditions for Plant Hatch are presented in tables 4-1 through 4-3 of NEDE-22154-1. The seismic curves identified in these tables are the Unit 2 floor response spectra curves gener-ated for the various mounting locations of the ATTS hardware. Figures 3.9 through 3.21 correspond to or envelop the Plant Hatch safe shutdown earthquake curves. Table 4-5 and figures 4-5 through 4-12 of NEDE-22154-1 l define the seismic levels to which the ATTS hardware was qualified.
' Comparison of the curves presented in figures 3.9 through 3.21 with the corresponding curves presented in figures 4-5 through 4-12 of NEDE-22154-1 shows that the ATTS hardware was seismically qualified to levels which exceed the Hatch requirements. Table 3.2 presents a summary of the various ATTS hardware along with the corresponding 3 mounting locations and references to the applicable Plant Hatch seismic curves.
(SSA.I.d An interconnection diagram, showing the basic interconnections betweer
\s,) the existing logic cabinets and the instrument cabinets and the new trip unit cabinets, is shown in figure 3.22. This interconnection diagram also presents the divisional separation of the cabinets.
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s 3A.I.e The qualification program to the requirements of IEEE 323-74 and IEEE 4
O. 344-75 is presently underway and is scheduled to be completed by the end of 1983. The details of the qualification program are presented in NEDE-22154-1. GPC will submit the final qualification reports (proprietary to GE)'to the NRC upon request.
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TABLF. 3.1 ( T I C7 B)
, INSTRUMENT LOOP INFORNATION a
i Primary Sensor Engineering Eulatina Device New Device Trip Unit Old New Safeguard Variable Nameg ,) HPL No. HPL No. NPL No. Division Nanufacturer Nodel No. Generic Name Manufacturer Nodel No.(c) Associated Back(b) j I) Besctor Steam Dcee B21-N620 NA 521-NI20 ECCS NA NA Pressure Barton 763 A,C-N21-P404A/B Pressure High A,B,C,D A,B,C D Transmitter B,D-N21-P405A/B
- 2) Low Low Set Control B21-N621* NA B23-N120 ECCS NA h4 Slave NA NA NA A,B,C,D A,B,C,D 1
! 3) Low low Set Control B25-N622 NA B21-NI22 ECCS NA NA Pressure Barton 763 A C-N2I-P404A/B l A,B,C,D A,B,C,D Transmitter B,D-N21-P405A/B
- 4) Steam Tunnel B21-W623 B21-N080 821-N123 BPS Fenwell 17002-40 RTD Weed IAOD Local Temperature High A,B,C D A,B,C,D A,B,C,D
- 5) Steam Tunnel B21-N624 B21-Noll B21-N124 BPS Fenwell 17002-40 BTD Weed IAOD Local Tempesature Ilish A,B,C,D A,B,C D A,B,C.D -'
j 6) Steam Tunnel B21-N625 B21-NOI2 B21-N125 RPS
- fenwall 17002-40 BTD Weed IAOD Local j Temperature High A,B,C,D A,B,C.D A,B,C D ,
,, 7 ) Steam Tunnel B21-N626 D21-h013 821-M126 BPS Fenwall 17002-40 BTD Weed IAOD Local 8 Temperature Nigh A,B,C,D A,B,C,D A,B,C.D
! B) Reactor Vessel B21-N64I* B21-N028 821-WO90 ECCS Barton 288 Slave NA NA NA
) Pressure Low B,C B,C, B,C
- 9) Reactor Vessel Steam 821-N678 821-N023 B21-N078 BPS Barksdale B2T-M12SS Pressure Barton 763 A,B-H21-P404C/D j Dome Pressure liigh A,B,C,D A,B,C,D A,B,C.D Transmitter C,D-H21-P405C/D
- Slave trip unit. '
- a. Transmitters having the same HPL No. are the sar.e equipment providing several trip functions.
i i b. Separated.for divisional considerations.
I i c. The romplete HTD model number for,all present ATTS applications is: IA00-6tt-IB-C-4-C-2-A2-0.
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TAB 12 S.I (SNBET 2 0F 8)
IllSTRIBENT IAOP INFORMATICII i
i Primary Sensor Ea81meerin8 Esisties Device New Device Trip Unit Old New Safeguard Vxritble Name4 ,) NPL No. NFL No. NPL No. Nanu f acture r i
Division Model No. Generic Name Manufacturer Model No(.c) Associated Back(b)
- 12) Beactor Vessel Steam 831-N679 B3I-N018 331-N079 BPS Baskadale 52T-N12SS Differential Barton 764 N21-P404E Dome Pressure Low A A A Pressure Tranawitter B31-N679 B31-N018 B31-N079 BPS Static-0-Bing SN-A33- Differential Barton 764 N21-P405E j D B D (X9)STT Pressure j .
Transmitter i II) Heat tor Vessel Water B21-N680 B21-Nol7 B21-N080 BPS Barton 288A Differential Barton 764 A,B-H21-P404C/D I.cvel I.nw (i.evel 3) A,B,C,D A,B,C,0 A,B,C,D Pressure C,D-N21-P405C/D i Transmittes I 12) Beactor Vessel Water B21-N681 B21-N024** B2t-N081 BPS Yarway 441BC Differential Barton 764 A,5-N21-P404C/D Ievel Low Low Low A,8 A,8 A,B Pressure (Level 1) Transmitter B28-N681 821-N025** B21-N081 BPS Yarway 441'8C Differential Barton 764 C,D-N21-P40$C/D
]w C,D A,B C,D Pressure ji Transmitter
, 13) Beactor Vessel Water B21-N682* B21-N024** B21-N081 BPS Yarway 441BC Slave NA NA NA
! Level Low Iow A,8 A,B A,B
) (Level 2)
B21-N682* B21-N025** B21-N081 BPS Ya rway 4418C Slave NA NA NA C,D A,B C,D
- 14) Beactor Shrouel 821-N685 B21-NO36 B21-N085 ECCS Ya rway 4418CE Differential Barton. 764 A-il21-P409 1 Water I.evel Low A,B B21-NO37 A,8 Pressure B-H21-P410 l
(Level 0) Transmittca
- 85) tlain Steam Line A B25-N686 B21-N006 821-N086 BPS Barton 288 Differential Barton 764 A,B-Il21-P415A/B Flow Ifigli A,B C.D A,B,C.D A,B,C,D
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Pressure C,D-li21-P425A/B Transmitter
- Slave t rip unit .
- Not being .lelet ed; os.ly t he safet y f unct ior. is l'eing replaced by ATTS.
- a. Transmitters having the same HPL No. are the same equipment providing several trip functions.
i b. Separated for divisional considerations.
- c. The complete BTD model number for all present ATTS applications is: IA00-611-IB-C-4-C-2-A2-0.
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O O U d [d TABLE 3.1 (SNEET 3 0F 8)
INSTBILENT IAOP INFORNATION Primary Sensor En8ineerin8 Esistina Device New Device Trip Unit Cid New Safe 8uard Vrrinble Name(*} NPL No. NPL No. NFL Wo. Division Manufacturer Nodel No. Generic Name Nanufacturer ModelNo!#iAssociated Back(*
. 16) Niin Steam Line B B23-N687 B21-N007 B21-N087 BPS Barton 238 Differ <atial Barton 764 A.B-N21-P4ISA/B 1
Flow liigh A,B,C,D A,B,C,D A,B,C.D Pressure C.D-N21-P425A/B Transmitter
- 17) Main Steam Line C B21-N688 B21-N008 B21-N088 BPS Barton 248 Differential Barton 764 A,B-N21-P415A/B Flow Iligh A,B.C.D A,B,C.D A,B,C D Pressure C,D-N21-P4,25A/B Transmitter l 18) Hain Steam Line D B21-N689 B21-N009 B21-W089 BPL Barton 288 Differential Barton 764 A,B-N21-P415A/B 1 Finw Iligh A,B,C D A,B,C,D A,B,C,D Pressure C D-N21-P425A/B l Transmitter l 19) Reactor Vessel B21-N690 B21-N021 821-N090ff ECCS' Barksdale B2T-M12SS Pressure Barton 763 A.E-N21-P404A i Pressure Low A,D,E.F A,D,E,F A.D.E.F Transmitter D F-N21-P405A 1 B21-N690 821-N02I B21-N090 ECCS Barton 288 Pressure Barton 763 B-N21-P410 1 uY, B,C B,C 5,C Transmitter C-N21-P409 j20) Reactor Vessel Water B21-N691 B21-NO31 B21-N091f ECCS Ya rway 44ISC Differential Barton 764 A,C-H21-P404A j Level Low Low Low A,B,C,D A,B,C.D A,B,C,D Pressure B,D-Il21-P405A (Level 1) Transmitter I
- 21) Reector Vessel Water B21 N692* B21-NO31 B21-N091~ ECCS Yarway 4418C Slave NA NA NA Level Low Low A,B,C,D A B.C.D A,B,C,D (Level 2)
- 22) Hesctor Vessel Water B21-N693 B21-Nol7 B21-N093 EJCS Barton 288A Differential Barton 764 Local
- level Aligh (Level 8) A,B A,B A,8 Pressure
- Transmitter j B21-N693* B21-NOIF B21-N095 ECCf Barton 288A Slave NA NA NA
, C,D C,D A,B
- Slave trip unit.
f B21-N091A,B will provide reactor water level indication ar>d replace by 821-N026A,B. B21-N091C,D will provide reactor water level recorder irputs and replace B21-N0260,D.
If B28-N090A.D will provide reactor pressure recorder inputs and replace B21-N05IA,B.
- a. Transmitters having the same MPL No. are the same equipment providing several trip functions. *
- b. Separated for divisional considerations.
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/~ s iR TABLE 3.1 (SHsET 4 0F 8)
INSTBtDENT IAOP INFOBHATION Primary Sensor Engineering Existina Device New Device g
Trip Unit Old New Safeguard Varisble Name ,) HPL No. HPL No. HPL No. Division Manufacturer liedel No. Generic Name Manufacturer Model No(.c)
Associated Back(b)
- 23) Reactor Vessel Water B21-N695 B21-N042 821-N09% ECCS Yarway 4418C Differential Barton 764 A-N21-P404B Level Low (Level 3) A,8 A,8 A,8 Pressure B-M21-P405B
- i. Transmitter
- 24) Drywell Pressure C71-N650 C71-N002 C71-N050 BPS Barksdale D2N-MI8SS Differential Barton 764 Local High A,B,C,D A,B,C,D A,B,C D Pressure Transmitter
- 25) HitH hm:p Discharge Ell-N655 Ell-N016 Ell-N055 ECCS Barksdale B2T-Ml2SS Pressure Barton 163 A C-N21-P4188 Pressure liigh A,B,C,D A,B,C,D A B.C.D Transmitter 8,D-N21-P421B
- 26) Rn;t Pump Discharge Ell-N656 Ell-N020 Ell-N056 ECCS Barksdale B2T-MI2SS Pressure Barton 763 N21-P418B Pressure liigh A,C A,C A,C Transmitter Ell-N656 Ell-N020 Ell-N056 ECCS Static-0-Bing SM-AA3- Pressure Barton 763 H21-P421B B,D B,D B,D (Il0)SITT Transmitter 1
- 27) kHR Pump Flow Ell-N682 Ell-N02I EII-Nd82 ECCS Barton 289 Differential Barton 764 A-H21-P418A
- ,,, Low A,8 A,8 A,8 Pressure B-H21-P42tA a Transmitter
- 28) Drywell Pressure Ell-N694 Ell-N010 Ell-N09% ECCS Barksdale D2H-M18SS Differential Barton 764 Local liigh A,B,C D A,B,C,D A,B,C,D Pressure Transmitter j ElI-N011#
A,B,C D j 29) Core Spray Pump E28-N651 E21-N006 E21-N051 ECCS Barton 289 Differential Barton 164 A-H21-P401 Discharge Flow l.ow A,B A,B A,B Pressure B-Il2I-P419 Transmitter .
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f fl51etef irs g>Iant. Functions of Ell-N0llA,B,L,D assigned to Ell-N010A,B,C,D.
- a. Transmitters having the same HPL No. are the same equipment providing several trip functions.
1 11 Separated for divisional considerations.
- c. The complete RTD model number for all present ArfS applications is: 1AOD-631-18-C-4-C-2-A2-0.
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TABLE 3.1 (SNEET 5 OF 8) ,
INSTBIBENT LOOP INFORMATION Primary Sensor Ensia:eering Existina Device hew Device Trip Unit Old New Safeguard i Vtrieble NameI " MPL No. MPL No. MPL No. Division Manufacturer Model No. Generic Name Manufacturer Model No. Associated Back(b) l 30) Core Spray Pump E21-N652 E21-N009 E21-N052 ECCS Barksdale B2T-M12SS Pressure Barton 763 A-N21-P401 Dischar2e Pressure A,8 A,8 A,8 Transmitter B-N21-P419 High
- 31) Core Spray Pump E21-N655 E21-N008 E28-N055 ECCS Barksdale B2T-M12SS Pressure Barton 763 A-N21-P408 Discharge Pressure A,8 A,B A,8 Transmitter a-H21-P419 H i g;.
- 32) tilt' Pusup E41-N650 E41-N027 E41-N050 ECCS Barksdale 'B2T-M12SS Pressure Barton 763 N21-P414B Pressure High Transmitter
, 31) IIPCI Pump E41-N651 E41-N006 E41-N051 ECCS Barton 289 Differential Barton 764 H21-P414A Discharge Flow Pressure High, l.ow Transmitter
- 34) IIPCI Pamp E41-N653 E41-NOIO E41-N053 ECCS Static-0-Bing 6N-AA21- Differential Barton 764 Il21-P414B a., Suction (X9)VSTT Pressure
- $ Pressure Low Transmitter
- 35) HPCI Turbine Exhaust E41-N655 E41-N012 E41-N055 ECCS Barksdale D2il-MISOSS Differential Barton 764 A,C-H21-P434 Dirihragm Pressure A,B,C,D A,B,C,D A 2,C,D Pressure B,D-H21-P414A High , Transmitter
, 36) HPCI Turbine Exhaust E41-N656 E41-NOIF E41-N056 ECCS Barksdale 82T-M12SS Pressure Barton 763 H21-P414B Pressure High B,D A,B, B,D Transmitter i 37) HPCI Steam Line E41-N657 E41-NC34 E41-N057 ECCS Barton 288 Differential Barton 164 A-H21-poi 6 l Differential Pressure A,B E41-N005 A,B , Pressure B-II21-P036 liigh (*) Transmitter
- a. Transmitters having the same MPL No. are the same equipment providing several trip functions.
- b. Separated for diGisional considerations.
- c. The complete HTI mo.lel number for all present ATTS applications is: IAOD-6tt-IB-C-4-C-2-A2-0.
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O O C TABLE 3.1 (SilEET 5 QF 8)
INSTRlRENT IDOP INFORMATI0ll 4
Primary Sensor Engl eering Existina Device New Device Trip Unit Old New Safeguard Variable NameI *I HPL No. HPL No. HPL No. Division Namulacturer Nodel No. Generic Name Nanufacturer Model No! AssociateJ Rack (
- 38) CPCI Steam Supply E41-N658 E41-N004 E41-N058 ECCS Barksdale B2T-H12SS Pressure Barton 763 A,C-N21-P016 Pressure Low A,B,C,D A,B,C,D A,B,C,D Transmitter 5,D-N21-P036
- 39) IIPCI Steam Line E41-N6604 E41-N004 E41-N057 ECCS Barton 298 Slave NA NA NA Differential A,B E41-N005 A,8 Pressure High (-)
401 IIPCI Turus Water E41-N662 E41-Nol5 E41-N062 ECCS Nagnetrol 3.5-751-MPG Capillary Barton 764 with Local 1.cvel liigh B,D A,8 B,D Differential Model 352 Pressure Capillary Transmitt<r Sensors
- 41) IIPCI E.guipment E41-N670 E41-NO30** E41-N070 ECCS Pyco NI45C3224PI RTD Weed IAOD Local Ambient Temperature A,B A,B A,8 Type T ,
liigh
- 42) IIPCI Pipe Room E41-N671 E41-N046** E41-N071 ECCS Pyco NI45C3224PI RTD Weed IA00 Local us Ambient Temperature A,B A,8 A,B Type T 1 Iligh
- 43) RCIC Pump Discharge F51-N650 E51-N020 E51-N050 ECCS Barksdale B2T-M12SS Pressure Barton 763 N21-P4178 Pressure High Transmitter
- 44) kCIC Pump Flow E51-N651 E51-N002 E51-N051 ECCS Barton 289 Differential Barton 764 H21-P417A liigh and Low Pressure i
Transmitter i
- 45) HCIC Tushine Exhaust E51-N656 E51-N009 E51-N056 ECCS Barksdale D211-H80SS Pressure Barton 763 H21 P4178 Pressure Iligh A,C A,B A,C Transmitter
- 46) HCIC Steam 1.ine E51-N657 E51-Nol7 E58-N057 ECCS Barton 28S Differential Barton 164 A-H21-P435 AP liigh D ) A,8 E51-Nol8 A,B Pressure B-H21-P038 Transmi t te r t
- Slave trip unit.
l ** Not being deleted; only the safety function is being replaced by ATTS.
I
! a. Transmitters havin'g the same HPL No. are the same equipment providing several trip functions,
- b. Separated for divisional considerations.
1 i
i
TABLE 3.1 (SNEET 7 OF 8) ,
litSTBIBENT IDOP INF0BMATlogi Primary Sensor Engineering Existina Device New Device g
Trip Unit Old New Safeguard c V*rtable Name ,) MPL No. HPL No. hPL No. Division Manufacturer Model No. Generic Name Manufacturer Model No(. ) Associated Rack (b)
- 47) RCIC Steam Supply E51-N658 E51-N019 E51-N058 ECCS Berksdale 32T-M12S8 Pressure Bartoa 763 A C-N21-P035 Pressure Low A,B,C,D A,B,C D A,5,C.D Transmitter B,D-N21-P038
'8)
, RCIC Steam Line E51-N6604 E51-N017 E51-N057 ECCS Barton 284 Slave NA NA NA AP liigh (-) A,B E51-N018 A,B
- 49) HCIC Equipment E51-N661 E51-N023e* E51-N061 WCCS Pyco N145C3224PI RTD Weed IA00 Local Ambient Temperature A,B A,8 A,5 Type T
- liigh
- 50) Torus A=hient E51-N663 E51-N026** E51-N063 ECCS Pyce NI45C3224PI klD Weed IAOD Local Tc=perature (no trip) A,B,C,D A,B,C,D A,B,C,D Type T E51-N664 E51-N027** E51-N064 ECCS Pyco M145C3224PI RTD Weed IAOD Local A,B,C,D A,B,C.D A,B,C D Type T SI) Torus Ambient E51-N666 E51-N025** E51-N066 ECCS Pyce M145C3224PI BTD Weed IAOD Local Temperature Nigh A,B,C,D A,B,C D A,B,C,D Type T
- 52) Terus Differential E51-N665 E51-N604** NA ECCS NA NA NA NA NA NA Temperature Nigh A,B,C D A,B,C,D
- 53) RCIC Pump Suction E51-N683 E51-N006 E51-NC83 ECCS Static-0-Ring 6N-AA21- Differential Barton 764 H21-l'417H Pressure Low (X9)VSTT Pressure Transmitter
- 54) RCIC Turbine Exhaust E51-N685 E51-Nol2 E51-N085 ECCS Barksdale D2H-H80SS Differential Barton 764 A,C-H21-P417A Diaphragm Pressure A,B,C,D* A,B,C,D A,B,C,D Pressure . B,D-H21-P437 Higli Transmitter
- 55) RWCl) Room Temperature G31-N661 G31-N023* G31-N061 RPS Pyco NI45C3224PI RTD Weed IAOD Local Inlet (no trip) A,D,E,H, A,B,C,D, A,D.E,N, Type T J.H E,F J,M i
T Stave trip unit.
- Not being eleleteel; only the safety function P>eing replaced by AlfS.
- s. Iransmitters having the same HPL No. are the same equipment providing several trip functions.
- b. Separated for divisional considerations.
1 c. The complete RTD mu,let number for all present ATTS applications is lA00-631-IB-C-4-C-2-A2-0.
i 1
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A O C
. )s TABIA 3.1 (SNEET B OF 3)
INSTRIBENT la0P INFORNATION Primary Sensor Engineering Esistina Device New Device Trip theit Old New Safeguard Verinble Name(* HPL No. HPL No. HPL No. Division Manufacturer Model No. Generic Name Manufacturer Model No. Associated Back( }
- 56) I:WCU Area Ventflation G31-N663. G31-N602** NA BPS MA NA NA NA NA NA Differential A.D.E.N. A,B,C,D, Temperature Nigh J.H E,F .
- 57) RWCU Hoom Outlet G31-N662 G31-N022** G33-N062 BPS Pyco M145C3224PI BTD Weed IA00 Local Ambient Temperature A D.E,N, A.B.C.D. A,D,E.M. Type T Iligh J,H E,F J,M
- 58) S.sfrty He!ief Valve NA NA B21-N302A ECCS NA NA Pressure PCI All-IP Local open B21-N302B Switch
- B2I-N302C B21-N302D ,
B21-N302E 32I-N302F B21-N302G B21-N302N B21-N302E u B2I-N302L
,'. 321-N302H o
- Not being deleted; only the safety function is being replaced by ATTS. '
- a. Transmitters having the same HPL No. are the same equipment having several trip settings,
- b. Separated for divisional considerations.
I i
TABLE 3.2 CORRELATION BETWEEN EQUIPMENT AND MOUNTING LOCATIONS
\
(For Use With Seismic Figures 3.9 through 3.21)
Description L'ocation Figure
- 1) Cabinets H11-P921 Main Control Room - el 164 ft 3.10, 3.19 to H11-P928 , (Control Building)
- 2) Racks- ,
MPL No.
a) H21-P401 Rer.ctor Building - el 87 ft 3.9, 3.10, 3.11 (NE Corner Room) b) H21-P402 Reactor Building - el 158 ft 3.15, 3.16, 3.17 c) H21-P404 A-E Reactor Building - el 158 ft 3.15, 3.16, 3.17 d) H21-P405 A-E Reactor Building - el 158 ft 3.15, 3.16, 3.17 e) H21-P409 Reactor Building - el 130 ft 3.12, 3.13, 3.14 f) H21-P410 Reactor Building - el 130 ft 3.12, 3.13, 3.14 g) H21-P414 A/B Reactor Building - el 87 ft 3.9, 3.10, 3.11 (HPCI Rootn) h) H21-P415 A/B Reactor Building - el 130 ft 3.12, 3.13, 3.14
'i) H21-P016 (Wall) Reador Building - el 82 ft 3.12*, 3.13*, 3.14*
(HPCI Room) j) H21-P417 A/B Reactor Building - el 87 ft 3.9, 3.10, 3.11 j (RCIC Corner Room) .
k) H21-P418 A/B Reactor Building - el 87 ft 3.9, 3.10, 3.11 (NE Corner Room)
- 1) H21-P016 (Wall) Reactor Building - el 87 ft 3.9*, 3.10*, 3.11*
(SE Corner Room) m) H21-P421 A/B Reactor Building - el 87 ft 3.9, 3.10, 3.11 4
(SE Corner Room) n) H21-P425 A/B Reactor Building - el 130 ft 3.12, 3.13, 3.14
! o) H21-P434 (Wall) Reactor Building - el 87 ft 3.12*,.3.13*, 3.14*
l (HPCI Room) p) H21-P435 (Wall) Reactor Building - el 130 ft 3.15*, 3.16*, 3.17*
q) H21-P036 (Vall) Reactor Building - el 130 ft 3.15*, 3.16*, 3.17*
3-11
4
. TABLE 3.2 (Continued)
Description Location Figure r) H21-P437 (Wall) Reactor Euilding - el 87 ft 3.12*, 3.13*, 3.14*
(RCIC Corner Room) s) H21-P038 (Wall) Reactor Building - el 87 ft 3.12*, 3.13*, 3.14*
(Torus Room)
- 3) Pressure Swit.ches, Use Multiple Locations-Plant 3.20, 3.21 RTDs Hatch Spectrum Peak Envelope Curves
- 4) Locally Mounted Use Multiple Locations - Plant 3.20, 3.21 Transmitters Hatch Spectrum Peak Envelope Curves A
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- The next higher elevation seismic curves should be used for the wall racks.
3-11a l - ...- - _ .-.. ,. _ - ...-.._ .__. .-, . _ _ . , _ . . . , _ _ . _ _ _ , .
O O O FIGlstE 3.I ASSIGliED TRIP UNIT LOCATIONS IN ECCS CARD FIIES FOR CABINET Hil-P925II I 2 3 4 5 6 7 8 9 10 11 12 E51-N(58C E51-N657A E51-N660A E51-N685A E51-N665C E51-N656A E51-N656C E51-Nf51 E51-N683 E51-N650 E58-N658Af*
ESI N058A E51-N058C E51-N057A SIAVE E51-N085A E51-N085C E51-N056A E51-N056C E51-N051 E51-N083 E51-N050 RCIC STM LN(# RCIC STH LN RCIC STM LN RCIC SUI LN RCIC fB EX RCTS TB EI RCIC TB EX RCIC TB EX kCIC PlkiP RCIC PUMP RCIC PUMP LOW PR IJDW PR HI AP(*) MI AP(-) DIA MI PP DIA HI PR HI PR HI PR MI FLOW SUC LOW PR Ill PR E48-N658A
- E41-N658C E41-N657A E41-N660A E41-N655A E41-N655C B21-N620A B21-N621A B21-N622A E41-N058A E41-N058C E41-N057A SLAVE Ef.1-N055A E41-N055C B21-NI20A SLAVE B21-N122A IIPCI STH LN(* IIPCI STM LN HPCI STH IJi NPCI STH LN HPCI TR EX NPCI TB EX LLS SCRAN LLS LLS LOW PR LOW PR 111 AP(t) HI AP(-) DIA HI PR DIA HI PR PR PERM CONTROL CONTROL B21-N691A
- B21-N692A B21-N693A 321-N691C 521-N692C B21-N693C B21-N695A B21-N685A Ell-N682A B21-N620C B21-N62 C B?l-N622C B21-N091A SLAVE B21-N093A B21-N091C SLAVE ISIAVE . 321-N095A B21-N085A Ell-N082A B21-NI20C SLAVE B21-NI22C WR LVL II " WR LVL 2 WR LVL 8 W R LVL 1 WR LVL 2 WIR LVL B' WR LVL 3 WR LVL 0 RHR PitlP A/C LLS SCRAM LLS LIS CS/ ADS /Ri!R llPCI/RCIC RCIC CS/ ADS /RHR NPCI/RCIC RCIC ADS RHR LOW FLOW PR PERM CONTROL CONTROL DIESEL DIESEL *
- a. Trir unit HPL No.
t.* te . . Sensor HPL No, d
M
- c. Loop function.
- d. Blank spaces denote empty file slots.
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4 FIGURE 3.2
- ASSIGNED TRIP UNIT LOCATIONS IN ECCS CARD FIIES FOR CABINET Mil-P926(b) ]
. I 2 3 4 5 6 7 R 9 10 11 12 ,
{
E51-N6588 E51-N658D E51-N657B E51-N6608 E51-N6858 E51-N685D E41-N656B E41-N656D E41-N651 E41-N653 E%I-N650 E51-N0588 E51-N058D E51-N057B SLAVE E51-N085B. E51-N085D E41-N056B E41-N056D E41-N051 E41-N053 E41-N050 RCIC STM LN RCIC STN LN RCIC STM LM RCIC STM IJi BCIC TB EX RCIC TB EX NPCI TB EX NPCI TB EX NPCI PUNP NPCI PUNP MPCI PitlP i
LOW PR LOW PR HI AP(+) HI AP(-) DIA MI PR DIA MI PR MI PR NI PR MI FLOW SUC IJ0W PR HI PR ,
E41-N658B E41-N658D E41-P6578 E41-N660B E41-N655B E41-N655D E41-N662B E41-N662D B21-N6208 B21-N6218 B21-N6228 ,
E41-N058B E41-NC58D E41-N0578 SLAVE E41-N055B E41-N055D E41-N0625 E41-N062D B21-N1208 SLAVE B21-NI?2B HPCI STH I.N HPCI STM IJi NPCI STM LN NPCI STM LN NPCI TB EX NPCI TB EX NPCI TORUS NPCI TORUS LLS SCRAM LLS LLS IDW PR IDW PR HI APl+) MI AP(-) DIA MI PR DIA MI'PR MI,WR LVL NI WR LVL PR PREN CONTROL CONTROL B21-N69tB B21-N6928 B21-N6938 B2 4- W69 lD B21-N692D B21-N693n B21-N6958 B21-N6858 Ell-N6828 B21-N6209 B21-N621D B21-N622D B21-N0918 SLAVE B21-N0938 B2 WO9tD SLAVE ' SLAVL B21-N095B B21-N085B Ell-N0828 B21-MI20D SLAVE B21-NI22D WTR LVL 1 WR LVL 2 WR LVL 8 WR i 'L I WR LVL 2 WR LVL B WR LVL 3 WR LVL 0 RHR PUMP B/D LLS SCRAM LLS LLS CS/ ADS /RHR HPCI/RCIC HPCI CS//Ji/RHR NPCI/RCIC NPCI ADS RHR LOW FI4W PR PERM CONTROL CONTROL DIESEt. DIESEL.
- a. See figure 3.1 for description of nomenclature.
l b. Blank spaces denote empty file slots.
Y
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, FICtMtE 3.3I
- ASSIGNED TRIP UNIT IDCATIONS IN ECCS CARD FIEES FOR CABINET Mll-P927 1 2 3 4 5 6 7 8 9 10 Il 12 Ell-N655A Ell-N655C E81-N656A Ell-N656C Ell-N694A Ell-N694C B21-N690A 521-N690C B21-N641C B21-N690E Ell-N055A EII-N055C Ell-N056A Ell-N056C Ell-N094A Ell-N094C B21-N090A B21-N090C SLAVE B21-N090E RHR PUNP A RI:R PUNP C RNR PUNP A RHR PttlP C DRYWELL DRYWELL VESSEL VESSEL VESSEL VESSEL HI PR HI PR NI PR HI PR HI PR NI PR IM PR IM PR IM PR IM PR E21-N651A E21-N655A E21-N652A E51-N663A E51-N664A E51-N665A E51-N663C E51-N664C E51-N665C E21-N051A E21-N055A E21-N052A E51-N063A E51-N064A NA E51-N063C E51-N064C NA CS FilHP A CS POS0 A CS PUNO A RCIC TORUS RCIC TORU',* RCIC TokUS NPCI TORUS NPCI TORUS IN FIM HI PR HI PR AnB Ih(NT)
- AHS NI(NT)' NI AT AHB th(NT)I*) 'MPCI 1LRUSAMB E41-N670A E41-N67tA E51-N66tA E51-N666A E51-N666C E41-N070A E41-N073A E51-N06tA E51-N066A ESI-N066C HPCI EQUIP NPCI PIPE RH RCIC EQUIP RCIC TORUS MPCI TORL'S AH8 Hi TMP AMB NI THP AMB NI THP ,
- a. See. figure 3.1 for description of nomenclature.
- b. Blank spaces denote empty file slots.
, , c. (NT) denotes instrument with no trip function.
9 l
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o s s V 0 FIGURE 3.4(*}
ASSIGNED TRIP UNIT LOCATIONS IN ECCS CARD fills FOR CABIWET Hll-P928 8 2 3 4 5 6 7 8
- 9 10 Il 12 Ell-N6558 Ell-N655D Ell-N6568 Ell-N656D Ell-N694B Ell-N694D b21-N690B B21-b64tB B21-N690D B21-N690F i
Ell-N055B Ell-N055D Ell-N056B Ell-N056D Ell-N094B Ell-N094D B21-N0908 SLAVE B21-N090C B21-N090F RHR PltlP B RHR PUNP D RHR PitIP B RHR PIRIP D DR M LL DR M LL VESSEL VESSEL VESSFL VESSEL HI PR HI PR HI PR HI PR 111 FR 111 PP llM PR IAM PR thW PR IJM PR 628-N651B E21-N6558 E21-N6528 E51-N663B E51-N664B E51-N665B E51-N663D E51-N664D E51-N665D E28-N0518 E21-N0558 E21-N0528 E5I-N0638 E51-N0648 NA . E51-N063D E51-N064D NA l CS PitIP B CS PUNP B CS Pl#tP B RCIC TORUS RCIC TORUS RCIC TORUS HPCI TORUS NPCI TORUS IAM FIDW HI PR HI PR AMR ID(NT)
- AMR HI(NT)I* HI AT AHB IA(NT)I* AHR III(NT)(*}HPCI HI AT T E41-N6708 E41-N671B E51-N6618 E51-N666B E51-N666D E41-N0?OB E41-N071B E51-N061B E51-N066B E51-N066D HPCI EQUIP HPCI PIPE RH LCIC EQUIP RCIC TORUS NPCI TORUS AtlB HI THP AHB HI TitP AMB HI TMP AHB NI TMP AMG HI TMP
- a. See figure 3.1 for description of nomenclatute.
- b. Blank spaces denote empty file slots.
]
- c. (NT) denotes instrument with no trig, function.
um i
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FIGURE 3.5I *)
ASSIGNED TRIP WIT IhCATIONS IN RPS CARD FILES FOR CABINET Mll-P921(bs I
i 1 2 3 4 5 6 7 8 . 9 10 II 12 i B21-N686A B21-N687A B21-N688A B21-N699A B21-N680A 321-N68IA B21-N682A B21-N674A B31-N679A C71-N650A 1
B21-N086A B21-N087A 321-N088A B21-N0894 B21-N080A B21-N081A SLAVE B21-N078A B31-N079A C71-N050A I MN ST LN A ItN ST LN B MN ST LN C HN ST LN D WR LVL 3 WTR LVL 1 W B LVL 2 VESSEL VESSEL DRYWELL l HI FLOW HI FLOW NI FLOW NI FIhW BPS MSIV ISOL NI PR LOW PR NI PR l B21-N623A B21-N624A B21-N625A B21-N626A B21-N123A B21-N124A B21-t'125A B21-N1264
P G31-N66tA G31-N662A G31-N663A G31-N661E G31-N662E G31-N663E G31-N661J G31-N662J G31-N663J G31-N061A G31-N062A NA G31-N061E G31-N062E NA G31-N061J G31-N062J NA-RWCU RM RWCU Ril OITT RWCU AREA RWCu RM RWCU RM OUT RWCU AREA
) RWCU RM TMP(NT)g gy AMB RWCU RM El TMP MIOUT AT RWCll AREA TMPiMT)ggg AMB HI TMP NI AT TMP(NT)ggg AMB HI TMP HI AT I
I i
j a. See figure 3.1 for description of nomenclature.
- b. Blank spaces denote empty file slots. .
- c. (NT) denotes instrument with no trip function.
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i FIGINtE 3.6I *)
l ,
ASSIGNED TRIP IBIIT IDCATIONS IN RPS C/RD FILES FOR CABINET Mll-?922 ,
I 2 3 4 5 6 7 8 9 10 11 12 i B21-N6868 821-N6878 821-N688B B21-N6898 B21-N6863 321-N5813 b21-N6425 521-N6748 C71-N6508 j B21-N0868 B28-N0878 B21-N088B B21-N0495 521-N080B B21-N0815 SIAVE 321-N078B C71-N050B i HN ST LN A HN ST LN B HN ST IN C HH ST LN D W R LVL 3 W R LVL 1 WR LVL 2 WESSEL DRYWELL lli FLOW HI FIAMi MI FLOW HI FIDW RPS MSIV ISOL MI PR HI PR B21-N6238 B21-N6245 B21-N625B B21-N626B
! B21-NI235 B21-N124B B21-N125B B21-N1268 STM TUNN STM TUNN STM TUNN STM TUNN
) Hi TMP NI TMP NI THP !!! TMP I
J
. a. See figure 3.1 for description of nomenclature.
- b. Blank spaces denote empty file slots.
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FIGIRE 3.7I *I ASSIGNED TBIP UNIT IDCATIONS IN BPS CARD FILES FOR CABINET Mll-P'923 I .
. I 2 3 4 5 6 _
7 8 9 to 11 12 l B21-N686C B21-N687C B21-N688C 521-N689C B21-N680C 521-N683C B21-N682C 523-N678C C71-II650C i B21-N086C ' B21-N087C 521-N088C B21-N089C B21-IIOSOC B21'-N081C SIAVE B21-N078C L71-N050C
, MN ST LN A INI ST LN B NN ST LN C HN ST IJi 0 WTB LVL 3 Wrt LVL 1 Wrt LVL 2 VE6SEL DBTWELL NI FLOW NI FIDW NI FIM MI FIM BPS IISIV ISOL NI PR N1 PR 521-N623C B21-N624C B21-N625C 521-N626C 521-N123C B21-N124C B21-N125C B21-M126C STM TUINI STM TUNN STil TUllN STM TUNN MI TMP NI TMP NI TMP NI TMP
/
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- a. See figure 3.1 for description of nomenclature.
I b. Blank spaces denote empty file slots.
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FIGURE 3.8I *I ASSIGNED TRIP UNIT IDCATIONS IN RPS CARD FIIIS FOR CABINET Mll-P924IN I 2 3 4 5 6 7 8 9 10 Il 12 B28-N686D B21-N687D B21-N688D B21-N689D B21-N6800 B21-N681D B21-N682D B21-N678D B31-N679D C71-N6500 B21-N086D B28-N087D B21-N0880 B21-N089D B21-N0400 B21-N081D SIAVE B21-N0780 B31-N0790 C71-N0500 MN ST LN A MN ST LN B HN ST LN C HN ST LN D VTR LVL 3 WTR LVL 1 WTR LVL 2 VESSEL VESSEL I4tYWELL HI FI.0W NI FIDW NI Fim MI FIM RPS MSIV ISOL MI PR la et NI PR B21-N623D B21-N624D B21-N625D B21-N6260 B21-N1230 B21-N124D B21-NI25D B21-N!26D 4
STM TUNN STM TUNN STM TUNN STM TUNN MI TMP MI TMP MI TMP NI TMP G31-N661D G31-N662D G31-N6630 G31-N661N G31-N662H G31-N663M G31-N66tM G31-N662M G31-N66'Al G31-N061D G31-N062D NA G31-N06118 G31-N062H NA G31-Nc61M G31-N062M M4 RWCU RM RWCU RM OUT RWCU AREA RWCU RM RWCU RM OUT RWCU AREA RWCU RM jg TMP(NT)g AMS RWCU HI TMP RM NI OlfT AT RWCU AREA TMP(NT)ggy AMB MI TMP HI AT TMP(NT)ggy AMB NI TMP NI AT d
- a. See figure 3.1 for description of nomenclature.
- h. Blank spaces denotes empty file slots.
- c. (NT) denotes instrument with no tril, function.
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FIGURE 3.22 BASIC CABLE ARRANGEMENT AND SEPARATION LAYOUT FOR PLANT HATCH ATTS
3B. BASES FOR TECHNICAL SPECIFICATIONS REVISIONS FOR ATTS EQUIPMENT With the incorporation of ATTS into the HNP-2 design, two types of Technical Specifications revisions are desirab'le:
i 4 e Nomenclature changes e Modifications to the surveillance frequency of ATTS equipment.
- The bases for the Technical Specifications changes over and above these ATTS changes are included in Section IV.
3B.1 The bases for these changes are as follows:
3B.1.a Nomenclature Changes to the Technical Specifications The installation of ATTS replaces mechanical switches with trans-mitters. In the past, the Technical Specifications identified the switches as such under the instrument description. Therefore, the Technical Specifications require a revision to reflect this change.
3B.1.b Modifications of the Surveillance Frequency As evidenced by NEDO-21617-A, the NRC previously approved the following surveillance frequencies for ATTS equipment:
() e Once per shift for channel check e Once per month for channel functional test e Once per operating cycle for channel calibration.
The revisions contained in the enclosed proposed changes to the Technical Specifications reflect the above surveillance frequencies.
Additional bases for the surveillance frequency revisions are con-tained within Chapter 6 of NEDE-22154-1 (Appendix 2).
! 3B.2 The proposed revisions illustrated in this submittal are the revi-l sions that will be incorporated into the Plant Hatch-Unit 2 system during the next refueling outage.
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3-34
l 3C.
SUMMARY
From the preceding discussions, it may be concluded that the proposed
, incorporation of the ATTS equipment into the Plant Hatch design and the proposed nomenclature and surveillance requirement changes to the Technical Specifications do not introduce an unreviewed safety ques-tion or a significant hazards consideration. This proposed change takes advantage of the sensor improvements and the decreased drift which was demonstrated for this new equipment.
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3-35
/] IV. PROPOSED PLANT MODIFICATIONS TO BE INCORPORATED INTO THE ATTS DESIGN V
4A. INTRODUCTION Section IV provides the bases for the proposed Technical Specifica-tions revisions, in addition to an overview of the evaluation per-formed to conclude that the proposed modifications constitute neither an unreviewed safety question, nor a significant hazards consideration.
Within this section three values are discussed: the analytical limit, the trip setpoint/ allowable value, and the trip setpoint. The analytical limit is the value used in the plant safety analyses. The trip setpoint/ allowable value was developed using the criteria of Regulatory Guide 1.105, taking into account instrument inaccuracies, and is the value that will be listed in the Technical Specifications.
The actual trip setpoint will take into consideration instrument drift.
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4B. SAFETY EVALUATION REPORT a.
4B.1 Reactor Water Low Low (Level 2) Trip Setpoint Modification The proposed Technical Specifications trip setpoint/ allowable value for the reactor vessel water level 2 signal is 1 -55 in. Reactor vessel water level 2 is for the initiation of high pressure coolant injection
- (HPCI) and reactor core isolation cooling (RCIC), and the recirculation pump trip.
Using Appendix K models, the Final Safety Analysis Report (FSAR) smer-gency core cooling system (ECCS) analysis was performed using a nominal analytical limit of -38 in. for level 2 HPCI actuation. However, the level 2 analytical limit is not a significant parameter in the Appendix K calculations. Therefore, as explained below, the ECCS calculations are insensitive to the variation in HPCI actuation water level so that a lower water level for level 2 has no signifi-cant effect on the ECCS system performance.
Sensitivity studies of the effect the level 2 analytical limit has on ,
the ECCS analysis have shown that level 2 may be lowered all the way to the level 1 analytical limit (-152.5 in.) without affecting the results of that analysis. However, it has not been recommended' '
that the level 2 analytical limit. be lowered to -152.5 in. The proposed analytical limit of -58 in. was selected to provide the oest flexibility and protective margin for the plant.
. V For small breaks, the limiting single failure is,the HPCI failure. Even if HPCI is assumed operable, the HPCI actuation at -58 in. rather than
-38 in, has an insignificant effect on core heatup.
Fortheprogedchangeinthelevel2analyticallimitto-58in.,
limit will not be changed.
the MAPLHGR The requirements of 10 CFR 100 will still te met because:
e Reactor power level or inventory is not changed.
e Engineering standards are not changed.
- e The probability of radio ative release will not be increased.
f The trip setpoint/ allowable value of 1 -55 in. was developed from the I
newly established analytical limit using the criteria of Regulatory
, Guide 1.105. The designated trip setpoint for the plant will take into consideration instrument drift.
Therefore, this modification constitutes neither an unreviewed safety question, nor a significant hazards consideration. No limiting condi-l tions of operation (LCOs) or surveillance frequencies are changed by
( this modification. Appendix 1 (page Al-2) provides the results of the significant hazards review.
- a. Maximum average planar linear heat generation rate.
i 4-2 L
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c., 4B.2. Deletion of High Drywell Pressure Signal for Residual Heat Removal (Shut-down Cooling Mode), RPV Head Spray Valves, and Reactor Water Cleanup j
( }i System Isolation High Drywell pressure has been used as a signal to isolate reactor water cleanup (RWCU) and the shutdown cooling mode of RHR. Small steam leaks in the drywell can cause a high drywell pressure signal which would prohibit an acceptable normal shutdown orocedure by preventing operation of the RHR and RWCU systems during the shutdown cooling mode. To resolve this operational concern, the high drywell pressure signal would be deleted from the isolation logic for the RHR shutdown cooling suction and discharge valves, as well as the reactor pressure vessel (RPV) head spray isolation valves and RWCU isolation valves.
The use of high drywell pressure as an isolation signal has little ef-fect in preventing coolant losses due to an RHR or RWCU pipe break inside the drywell since the inboard isolation valves are located as close as possible to the drywell wall. Such pipe breaks do not present a site boundary dose problem since the leaked fluid and associated radioactivity are completely retained within the primary containment boundary. The high drywell pressure signal for the RPV head spray isolation valves will also be deleted. Since the RPV head spray valves are used as part of the shutdown cooling procedures, this change is consistent with above mentioned proposed RHR (shutdown cooling mode)
System Modification.
This change does not affect the Appendix K calculation results presented in the FSAR. The requirements of 10 CFR 100 will still be met. This modification, theref ore, constitutes neither an unreviewed saf ety ques-tion, nor a significant hazards consideration. Appendix 1 (page Al-3) provides the results of the significant hazards review.
This modification has been implemented on other boiling water reactor (BWR)/4s.
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1 j) 4B.3 Lowered' Water Level Trip Setpoint for Isolation of Reactor Water Cleanup
\ s/ System and Secondary Containment, and Starting of Standby Gas Treatment System (SGTS)
Reactor scram from normal power levels (above 50 percent of rated) usually result.s in a reactor vessel water level transient due to void collapse that causes isolation of the RWCU system at reactor water level 3.
The result is typically the dropping of the cleanup filter cake, added r2dwaste processing, loss of ability to remove water from the reactor vessel immediately after scram, and other undesirable operational problems. These results adversely affect plant availability and opera-bility. By lowering the isolation setpoint to reactor water level 2, these problems may be resolved without any adverse safety impact.
The lowering of the level trip for isolation of RWCU from reactor water level 3 to reactor water level 2 will not have any adverse effect on plant transient and accident analysis. For any reactor pressure coolant boendary line breaks inside the primary containment, the LOCA design basis accident (DBA) analysis shows that the ECCS is capable of mitigating all break sizes including and up to the re-
! circult ion line break. For a RWCU line break outside the primary containment, the break detection is provided by the high differential flow, area high temperature, or area high ventilation differential tem-perature rather than by water level variation.
L By lowering the SGTS actuation and secondary containment isolation from
'f reactor water level 3 to reactor water level 2, a potential for spurious
, . g? trips is reduced. The ECCS analysis design basis assumes that the i
SGTS will initiate at the same time as the ECCS which initiates at reactor water level 2. -
This modification has been implemented on other BWR/4s. The require-ments of 10 CFR 100 are still met. These changes to the plant constitute neither an unreviewed safety question, nor a significant I i hazards consideration. Appendix 1 (page A1-4) provides the results of
- the significant hazards review. .
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O 4-4'
4B.4 Deletion of Ambient Temperature Loops in Leak Detection System j
Typically, the leak detection system uses ambient and differential tem-peratures to detect the small high-temperature leaks. In the earlier
- design, the ambient temperature trip was provided by an independent temperature element and trip device, and the differential temperature trip was provided by two independent temperature elements and a AT trip
' device as shown on figure 3.1. By using the ATTS, the ambient tem-perature trip may be obtained from one leg of the differential tempera-ture trip shown on figure 3.2. In figures 3.1 and 3.2, it can be seen that K1 will provide the high ambient temperature trip, and K2 will provide the differential temperature trip. With this arrangement, the
~
sensitivity of leak detection may be changed slightly, dependent on heating, ventilation, and air-conditioning design; but it will not defeat the intended function of the system. This arrangement is suit-able for the small rooms containing leak detection temperature monitor-ing as part of the, isolation logic, because only large rooms, such as the turbine building, need the spatial location of sensors to adequately protect the room against leaks. This scheme will allow the deletion of several unnecessary temperature loops in the RWCU system.
. The RWCU temperature and differential temperature sensors sense the tem-perature in the two pump rooms and the heat exchanger room. Each room
'has a redundant set of temperature instrumentation that provides input i to the RWCU isolation logic. By using the hot leg of the differential
'g*
temperature sensor for the high ambient trip as shown on figure 3.2, several devices may be deleted without any loss of protective function.
For these modifications, single-failure criteria will be maintained.
The following six temperature loops are proposed for deletion:
, MPL No.
Sensor Switch G31-N016A G31-N600A G31-N016B G31-N6005 G31-N016C G31-N600C G31-N016D G31-N600D G31-N016E G31-N600E G31-N016F G31-N600F The proposed Technical Specifications revisions will reference the trip unit loop from which the ambient temperature trip is taken in place of the existing ambient temperature trip instrument. Included in these-proposed changes are new surveillance frequencies which correspond with the surveillance requirements of the ATTS. This modification does not constitute an unreviewed safety question or a significant hazards consideracion. No LCO requirements will change as a result of this modification. Appendix 1 (page Al-5) provides the results of the significant hazards review.
O 4-5
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TYMCAL ECulPMENT ROOM TEMPERATURE ELEMENT AIR FLOW d h CHILLER
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DIFFERENTIAL TERWERATditE SWITCH (K1 K2 101634
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4B.5- Deletion of Drywell Pressure Sensors E11-N011A, B, C, D
- The original design of Plant Hatch has the high drywell pressure signals ,
for the ECCS coming from eight sensing devices. For example, E11-N011A, B, C, D (existing MPL numbers) provide signals to RHR, core spray, and HPCI; E11-N010A, B, C, D (existing MPL numbers) provide signals to ADS.
This configuration is inconsistent with the inputs for the reactor water levels 1 and 2 trips which are provided by only four sensing devices, namely.B21-NO31A, B, C, D (existing MPL numbers). To make drywell pressure sensor configuration consistent with that for the water levels 1 and 2 sensors, drywell pressure sensors E11-N010A, B, C, D may be used to provide signals for all four systems of the ECCS and still maintain single-failure criteria. Plant safety margin is not being reduced since the level of redundancy to serve a trip function is maintained.
This change deletes instruments E11-N011A, B, C, D and transfers their associated trip function to instruments E11-N010A, B, C, D. Since these instruments (E11-N010A, B, C, D) are being incorporated into the ATTS modification, the instrument number was changed to E11-N694A, B, C, D.
It is proposed that the surveillance frequencies be modified to those of instruments E11-N010A, B, C, D. It-should be noted that the surveillance frequencies of both E11-N010A, B, C, D, and E11-N011A, B, C, D are the same in the existing Unit 2 Technical Specifications. As a result of n s the above discussion, this modification represents neither an unre-y,,) viewed safety question, nor a significant hazards consideration.
Appendix 1 (page Al-6) providec the results of the significant hazards review.
i 4-8
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l T 4B.6 Trip Setpoint/ Allowable Value*Setpoint Modifications
,-)
4B.6.1 The instruments to be incorporated into the ATTS possess less drift and greater accuracy than the existing instruments in use at Plant Hatch.
Therefore, new calculations were performed to determine the setpoint value for each instrument. The setpoint calculations were made using the criteria of Regulatory Guide 1.105. The Plant Hatch. analytical limits were used (where applicable) to develop the allowable values and trip setpoints. Unless Identified in the text, the analytical limits used to develop these setpcints are the values used in the design basis of Plant Hatch. The values that are proposed to be inserted into the Technical Specifications are the calculated allowable values. The Technical Specifications also have a column to indicate the trip set-point. To avoid inconsistency in the specifications, this table re-flects that the trip setpoint is more conservative than the allowable value. The setpoints used at Plant Hatch wil' take into consideration instrument drift and will be developed from the allowable values. The proposed Technical Specifications revisions include modification of the trip setpoints/ allowable values for the following instruments:
, (RPS) Trip Function Trip Unit MPL No.
- 1. Main steamline flow - high B21-N686A,B,C,D B21-N687A,B,C,D O B21-N688A,B,C,D B21-N689A,B,C,D
- 2. Main steamilne tunnel B21-N623A,B,C.n temperature - high B21-N624A,B,C,s B21-N625A,B,C,D B21-N626A,B,C,D 3.
D Reactor vessel steam dome B21-N678A,B,C,D pressure - high
, 4. Reactor vessel water level - B21-N680A,B,C,D level 3
- 5. Reactor vessel steam dome B31-N679A,D pressure - low
- 6. Reactor vessel water level - B21-N681A,B,C,D level 1
- 7. Drywell pressure - high C71-N650A,B,C,D I
- 8. RWCU room ambient tempera- G31-N662A,D,E,H,J,M ture - high
(
() RWCU room differential temperature - high G31-N663A,D,E,H,J,M G31-N661A,D,E,H,J,M 4-9
( _ _ _ _ __ . _ __ _ _ _ -
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ECCS Trip Function
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- Trip Unit MPL No.
- 1. Reactor vessel water level - B21-N691A,B,C;D level 1
- 2. Reactor vessel water level - B21-N693B,D level 8
- 3. Reactor shroud water level - B21-N685A,B level 0 ,
- 4. Reactor vessel pressure - B21-N690A,B,C,D low-
- 5. Reactor vessel pressure - B21-N690E,F low B21-N641B,C
- 6. Reactor vessel water B21-N695A,B level - level 3
- 7. Drywell pressure - high E11-N694A,B,C,D
- 8. RHR pump discharge pres- E11-N655A,B,C,D ,
sure - high E11-N656A,B,C,D
- 9. Core spray pump discharge E21-N655A,B (s~_~f) pressure - high E21-N652A,B
- 10. HPCI steam supply pres- E41-N658A,B,C,D sure - low
- 11. HPCI steamline differ- E41-N657A,B ential pressure - high E41-N660A,B
- 13. Suppression chamber water E41-N662B,D level - high
- 14. Emergency area cooler E41-N670A,B i
temperature - high
- 15. HPCI pipe penetration room E41-N671A,B l ambient temperature - high
- 16. RCIC steamline pressure - E51-N658A,B,C,D low
- 17. RCIC turbine exhaust dia- E51-N685A,B,C,D phragm pressure - high
- 18. RCIC steamline differ- E51-N660A,B ential pressure - high E51-N657A,B 4-10
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- 19. Suppression chambir ambient E51-N666A,B,C,D_ _
temperature - high
- ^
Suppr'essica chamber differ- E51-N664A,B,C,D -
, s ential temperature - high E51-N665A,B,C,D E51-N663A,B,C,D t.
. -. _ ~
- 20. Emergency area cooler E51-N661A,B
~ ,
t'emperature - high .
, 1
~ ~ w.
$ t g
= k, s
w w
%_ o, -
-gut. -% 9
.g k i, w;
pA f AN -g
= -
j T" ? *s N
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=
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4 s I
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w 4-11
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i 4B.6.2 The bases for these proposed changes are as follows:
4B.6.2.1 Setpoint Bases for Trip Functions Assigned to RPS Cabinets i .1. Main steamline flow - high (B21-N686A,B,C,D; B21-N687A,B,C,D; B21-N688A,B,C,D; B21-N689A,B,C,D)
The analytical limit of 120 psid corresponds to 140 percent of rated flow. The 140 percent of rated flow is used as an input to the.high-energy line break'(HELB) calculations. The trip
' ~" '~
setp,oint/ allowable value of 5 125 psid was developed using the
('
, . . ~~'di~i teria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into account instrument drift.
- 2. Main steam line tunnel temperature - high (B21-N623A,B,C,D; B21-N624A,B,C,D; B21-N625A ,B ,C ,D ;
B21-N626A,B,C,D) ,
The value of 200*F was designed to detect a small steam line break. In addition, it provides early isolation of the main steam line to meet 10 CFR 100 requirements. The trip setpoint/ allowable value of 5 194*F was developed using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into account instrument drift.
- 3. Reactor vessel steam dome pressure - high (B21-N678A,B,C,D)
- The analytical limit of 1071 psig is the value used in the plant transient analysis and is documented in the Plant Hatch-Unit 2 FSAR. The intent of this reactor scram function is to provide the nuclear system process barrier overpressure protection, because a rupture to the process barrier may result in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by collapsing voids, thus adding reactivity. A reactor scram will quickly reduce the neutron flux, counteracting the pressure increase. The trip set-point / allowable value of 5 1054 psig was developed using the criteria of Regulatory Guide 1.105. The designated trip set-point for the plant will take into consideration instrument drift.
- 4. Reactor vessel water level - level 3 (B21-N680A,B,C,D)
Reactor vessel water level 3 is for isolation of primary con-taiment, initiation of reactor scram, and for closure of RHR
, shutdown cooling isolation valves. The analytical limit of 7.5 in. was used to determine the trip setpoint and allow-able value. A trip setpoint/ allowable value of > 8.5 in. was developed using the criteria of Regulatory Guide 1.105. The i
O. designated trip setpoint for the plant will take into considera-tion instrument drift.
4 4-12
,g_,--_ _ ,, . _ . - . . , , _ _ , . . - , , . _ . .rm. , . _ _ _ _ . , _ , , . , . . ,m_m,. . _ _ , , , , , , , . _ _
- 5. Reactor vessel steam dome pressure - low (B31-N679A,D)
(}
The trip function provides the low-pressure permissive signals to the RHR shutdown cooling mode loops. The analytical limit of 162 psig was determined based on plant-specific RHR system piping design and layout. The trip setpoint/ allowable value of 5 145 psig was developed using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into account instrument drift.
- 6. Reactor vessel water level - level 1(B21-N681A,B,C,D)
The analytical limit fcr this application is -152.5 in. The trip setpoint/ allowable value of 2 -221.5 in. was develcped using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into consideration instru-ment drift.
- 7. Drywell pressure - high (C71-N650A,B,C,D)
The value used as the analytical limit for this RPS scram function is 2.0 psig. The trip setpoint/ allowable value of 5 1.85 psig was developed using the criteria of Regulatory Guide 1.105. The de-signated trip setpoint for the plant will take into consideration instrument drift.
- 8. RWCU room ambient temperature - high (G31-N662A,D,E,H,J,M)
RWCU area differential temperature - high (G31-N663A,D,E,H,J,M; G31-N661A,D,E,H,J,M)
The leak detection system uses ambient and differential tempera-tures to detect small high-temperature leaks. The present design for this system uses a value of 130*F for the RWCU room outlet ambient temperature and 75*F for the RWCU area differential tem-perature. -
The trip setpoint/ allowable values of 5124*F for the RWCU ambient temperature trip and $ 67*F for the EWCU differential temperature trip were developed using the criteria of Regulatory Guide 1.105.
4B 6.2.2 Setpoint Basis for Trip Functions Assigned to ECCS Cabinets
- 1. Reactor vessel water level - level 1 (B21-N691A,B,C,D)
The analytical limit of -152.5 in. is the value in the FSAR ECCS analysis for automatic depressurization system (ADS), low-pressure coolant injection (LPCI), and core spray initiation.
The trip setpoint/ allowable value of 1 -121.5 in. was developed
()
'ss) using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into ceasideration instrument drift.
4-13
, - . ~ - - . - - _ . - - . . - . .. - . . .. -- _ = , _ .
(
,{ } 2. . Reactor vessel water level - level 8 (B21-N693B,D)
After HPCI has activated, this trip function prevents the water level in the reactor vessel from reaching the height of the main steam outlet. Its' intended protective function is accom-plished by tripping the HPCI steam turbine when the water level in the reactor vessel has reached the level 8 setting. The trip function is to protect the HPCI turbine system from po-tential damage. The cualytical limit for this function is 59.5 in. The trip setpoint/ allowable value of < 56.5 in. was devel-
- oped using the criteria of Regulatory Guide 1.105. The desig-nated trip setpoint for the plant will take into consideration instrument drift.
r
- 3. Reactor shroud water level - level 0 (B21-N685A,B)
This trip function acts as a bypass interlock to prevent con-tainment spray initiation during LPCI operation when the water level of the reactor shroud is low. Under post DBA conditions, water pumped from the suppression chamber through the RHR heat exchanger may be diverted for containment spray to remove the energy from the drywell. If the water level in the vessel.is low, this function prevents the water from being diverted. An analytical value of -211 in. is used to develop the allowable i value and trip setpoint. The trip setpoint/ allowable value of
>-207 in. was developed using the criteria of Regulatory Guide i 1.105. ~The~ designated trip setpoint for the plant will take into consideration instrument drift.
l i 4. Reactor vessel pressure - low (B21-N690A,B,C,D)
These instrument channels monitor the reactor vessel pressure to provide the permissive signal for core spray and the injec-tion permissive signal for RHR (LPCI). The analytical limit ,
j for this pressure permissive function has a range from 400 to l 500 psig. The criteria for this setpoint are that it should be l low enough to assure no overpressurizatioa of the low pressure
- ECCS and yet high enough so as not to delay the flow injection
, of the low-pressure system. The trip setpoint/ allowable value '
of 2 422 psig was developed using the criteria of Regulatory Guide 1.105 and an analytical limit of 400 psig. Since the important parameter is to ensure that RHR and core spray are I
operating as the pressure drops, a sign change of 1 is required.
The existing requirement that the setpoint be <500 psig is also retained in the Technical Specifications. The designated trip
, setpoint for the plant will take into consideration instrument drift. A new post-accident surveillance recorder (2B21-R623A,B) will take its signal from the B21-N690A,D transmitter.
(
- 5. Reactor vessel pressure - low (B21-N690E,F; B21-N641B,C)
-( An analytical limit of 300 psig is used in the Appendix K ECCS l calculations. The intent of this trip function is to provide a l permissive for closure of the recirculation discharge valves.
The present Hatch setpoints of 309 psig and 314 psig are above 4-14
('] -
t the analytical limit of 300 psig. Plant Hatch, therefore, has
. sd operated within its design basis. However, a sign change is being proposed to correct the Technical Specifications deficiency. The trip setpoint/ allowable value of 1 325 psig was developed using the criteria of Regulatory Guide 1.105. '
i' The designated trip setpoint for the plant will take into
. consideration instrument drift.
- 6. Reacter vessel water level - level 3 (B21-N695A,B)
Reactor vessel level 3 provides an ADS permissive. The analytical a
limit of 7.5 in. was used in the FSAR ECCS analysis. The trip Jetpoint/ allowable value of 1 8.5 in, was developed from the
, analytical limit using the criteria of Regulatory Guide 1.105.
The designated trip'setpoint for the plant will take.into con-sideration instrument drift.
- 7. Drywell pressure - high (E11-N694A,B,C,D) i' The vclue used for the analytical limit for the ECCS function is 2.0 psig. The trip setpoint/ allowable value of < 1.85 psig was developed using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into consideration instrument drift.
. - 8. RHR pump discharge pressure - high (E11-N655A,B,C,D; x,,
E11-N656A,B,C,D)
These instruments provide the RHR pump high pressure permissive
, signal for ADS. The ADS interlock to sense if low-pressure ECCS
- ' pumps are running is not an direct input to the ECCS calculaticas; however, correct operation of the interlock is an analytical assumption. To avoid any false indication during an accident, the trip setpoint should be above the suction relief valve set-point of approximately 100 psig. The range of the analytical limit is between 100 and 150 psig.
Selecting 100 psig at the lower bound for the analytical limit, the trip setpoint/ allowable value of 1 105 psig was developed using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into consideration instru-ment drift.
l i
- 9. Core spray pump discharge pressure - high (E21-N655A,B; E21-N652A,B)
The core spray pump discharge pressure - high trip is part of the
! ADS interlock function. This trip is not used as input to any l transient or safety analysis; however, correct operation of the j trip is assumed. To avoid a false indication during an accident, l this ADS permissive function should be set above the suction relief valve setpoint. To assure this, the value from which to x determine the trip setpoint and allowable value has been conser-vatively selected to be at 1 125 psig. The trip setpoint/
~
l allowable va~1ue of 1 130 psig was developed using the criteria j 4-15
.. ._ _ _ _ . _ . _ _ _ _ _ _ . _ _ , _ _ _ _ _ . _ . _ _ - _ _ _ . . _ _ _ _ _ _ . _ _ _ ~ , , _ . . _ -
____._;_, _- _- w- -u _ .= _ a- __ a
- l 4
O of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into consideration instrument drift.
- 10. HPCI steam supply pressure - low (E41-N658A,B,C,D)
This trip function is intended to prevent HPCI turbine stall and possible HPCI turbine damage. The signal of low HPCI steamline pressure will close the HPCI isolation valve and trip the HPCI turbine.. Even though the HPCI design specification tutd Tech-nical Specifications only require HPCI operability down to 150 psig, it is desirable to operate HPCI to as low a steam I
pressure as allowable. It has been determined that the use of an allowable value of > 100 psig will prevent the possibility
, of damage to the HPCI equipment due to turbine stall. The designated terip setpoint for the plant will take into con- t sideration instrument drift.
l
- 11. HPCI steamline differential pressure - high (E41-N657A,B; E41-N660A,B)
The purpose of this instrumentation is to detect HPCI steamline 4
breaks and to isolate the HPCI system to confine the resulting radioactivity release and limit the reactor inventory loss.
The HELB analysis assumes that the HPCI turbine trips and, the system isolates at 300 percent of rated flow. However, the O HELB analysis is used for guillotine breaks which have flows
'u/ several times higher than 300 percent of rated flow. A con-cervative analysis shows the leakage detection instrumentation isolates in the 400 percent of rated flow range with less inventory loss and less peak qualification parameters than the inventory
- loss and the qualification parametera calculated in the extreme HELB analysis. Since operability problems are a concern with
, setpoints derived from an analytical limit of 300 percent (using i
Regulatory Guide 1.105 methodology for this function), it is proposed that the present Plant Hatch setpoint be maintained.
This setpoint was proven to be acceptable from operability con-
[ siderations. Using this setpoint, an allowable value of 307 percent of rated flow was selected taking into account instrument drift.
The trip setpoint for this instrumentation shall be low enough such that'when the rupture disc blows, the transmitter / trip unit will activace imroediately, causing isolation of the HPCI system, yet high enough such that atmospheric variations will nat cause I unnecessary HPCI isulation. This isolation minimizes steam re-l leases to the secondary containment. Since this trip function is not directly considered in any safety or transient analysis performed for the FSAR, no analytical limit is available.
An allowable value of 5 20 psig was determined to be adequate in preventing turbine stall. Also, since high-pressure conditions only occur when the diaphragm is ruptured, an abnormally I
! 4-16 l
L___ _ . _ _
- high pressure condition at the turbine exhaust is detected always by these pressure switches with an allowable value of 20 psig. The trip used for the instruments at the plant will take into consideration instrument drift using the criteria of Regulatory Guide 1.105.
- 13. Suppression chamber water level - high (E41-N662B,D)
The purpose of this t.-ip function is to transfer the HPCI suc-tion from the plant condensate storage tank to the suppression chamber when the water in the suppression chamber rises above a predetermined level.
Although no analytical limit has been developed for this func-tion, the requirements of Mark 1 Long-Term Program have been considered. The trip setpoint was developed to minimize undesir-able trips. The allowable value of 5 33.2 in. was determined by adding the design drift allowance to the trip setpoint.
- 14. Emergency area cooler temperature - high (E41-N670A,B)
The leak detection system uses ambient temperature elements to detect any small high-temperature leaks. This trip function provides a signal to trip the HPCI turbine and closes the HPCI isolation valves. The present design for this system uses a O value of 175"F to isolate the HPCI equipment. The trip set-point / allowable value of 5 169*F was developed using the criteria of Regulatory Guide 1.105. - The designated trip setpoint for the plant will take into consideration instrument drift.
- 15. HPCI pipe penetration room ambient temperature - high (E41-N671A,B) 4 The leak detection system uses ambient temperature elements to de'tect any small high-temperature leaks. This trip function provides a signal to trip the HPCI turbine and close the HPCI isolation valves. The present design for this system uses a value of 175*F to isolate HPCI equipment. The trip setpoint/
, allowable value of 5 169*F was developed using the criteria of
, Regulatory Guide 1.105. The designated trip setpoint for the i
plant will take into consideration instrument drift. The iden-
! tification of this trip function was modified to clarify the location of the instrument. However, the general area where the equipment is located will not change.
, 16. RCIC steam supply pressure - low (E51-N658A,B,C,D) i Per the Unit 1 Technical Specifications, the RCIC system is only required to operate down'to 150 psig, but the system is capable of operating at lower pressures. The intent of this trip func-tion is to prevent RCIC turbine stall at low reactor pressure.
i O- An allowable value of > 60 psig was determined to be adequate l in preventing turbine stall. The trip setpoint used for the I
instruments at the plant will take into consideration instrument l drift using the criteria of Regulatory Guide 1.105.
4-17 I- , . , . - . - - - , - - - - - - - - - . . - - - - _ _ _ _ _ .
. m _. . _ _ .__ . .
The bases for the trip setpoint/ allowable value of 20 psig are identical to that for the HPCI turbine exhaust diaphragm pres-sure - high trip provided in item 12.
- 18. RCIC steamline differential pressure - high (E51-N660A,B; E51-N657A,B)
For' reasons similar to those ,for the HPCI steamline differential pressure - high trip (provided in item 11), it is proposed that the present Plant Hatch setpoint be maintained. This setpoint was proven to be acceptable from operability considerations.
Using this setpoint, an allowable value of 312 percent of rated flow was calculated taking into account instrument drift.
- 19. Suppression chamber ambient temperature - high (E51-N666A,B,C,D)
Suppression chamber differential temperature - high (E51-N665A, B,C,D; E51-N663A,B,C,D; E51-N664A,B,C,D)
The leak detection system uses ambient and differential tem-peratures to detect small high-temperature leaks. The present design for this system utilizes a trip value of 175*F for the suppression chamber ambient temperature - high and 50*F for the e suppression chamber differential temperature - high. The trip.
setpoint/ allowable values of 5 169'F for the ambient temperature and 5 42*F for the differential temperature were developed using the criteria of Regulatory Guide 1,.105. The designated trip set-point for the plant will take into consideration instrument drift.
- 20. RCIC emergency area cooler temperature - high (E51-N661A,B)
The leak detection system uses ambient teaperature elecents to detect any small high-temperature leaks and provides a signal to trip the RCIC turbine and close the RCIC isolat. ion valves. The present design uses a trip value of 175'F for providing the intended RCIC isolation function. The trip setpoint/ allowable value of 5 169*F was developed using the criteria of Regulatory
- Guide 1.105. The designated trip setpoint for the plant will take into consideration instrument drift.
4B.6.3 Cooclusion For the trip functions identified in this section, the analytical limits that are presented are such that neither transient nor safety analysis
! results documented in the FSAR are adversely affected. The uncertain-ties associated with instrument accuracy, calibration, and drift are considered in the setpoint determination. Therefore, it was concluded that the proposed instrumentation setpoint changes do not reduce the margin of safety of the current Technical Specifications or change the g FSAR setpoint bases and, therefore, constitute neither an 10 CFR 50.59 unreviewed safety question, nor a significant hazards consideratica as i described in 10 CFR 50.92. Appendix 1 (page Al- ) provides the results
- of the significant hazards review.
4-18 L , _ ._ _ _ . . _ _ - ..-_ _ _ _ _ _ .__ _ _ __._ _ . _ _.-~ _, ___. _ , _ _ - . .. ._ _ .~._ _ _
4B.7 Reactor Vessel Water Level - High (Level 8) Trip Instrumentation N
Modifications After HPCI and RCIC have activated, this trip function prevents the water level in the reactor vessel from reaching the height of the main steam outlet. Its intended protective function is accomplished by tripping the HPCI steam turbine enclosing the HPCI and RCIC' steam supply valves when the water level in the reactor vessel reaches the level 8 setting. The trip function is to protect the HPCI and RCIC steam turbine system from potential damage.
This trip function is currently assigned to B21-N017A,B,C,D which controls the RPS reactor vessel water level 3 instrumentation. To separate the RPS and ECCS functions, the ATTS design assigns the reactor vessel water level 8 trip function to ECCS instrumentation.
Since the functions of the lev &l 8 trip remain the same, this modification constitutes neither an unreviewed safety question, nor a significant safety hazards consideration. Appendix 1 (page Al-8) provides the results of the significant hazards review.
The analytical limit for this function is 59.5 in. The trip setpoint/
allowable value of < 56.5 in, was developed using the criteria of Regulatory Guide 1.105. The designated trip setpoint for the plant will take into consideration setpoint drift.
n-s- .
e I
I
- s_/
i 1 --
4-19 i
t
-4B.8 El'inination of the Rea$ tor Pressure Permissive to the ~ Bypass of t}ie MSIV' Closure Signal-Due'to Low Condenser Vacuum-O It is proposed to delete the reactor steam' dome pressure permissive . 1 which prevents. the group 1 isolation valves signal from being bypassed on a low condenser vacuum isolation.
The-manual bypass is provided to facilitate the following operations:
A. The-bypass allows cold shutdown testing of the main steamline
- f. isolation logic and allows stroking the MSIVs open and closed for maintenance even though there is no condenser vacuum.
' ~ ~
B. The bypass allows the MSIVs to be opened so seal steam and ejector steam can be available at.the turbine and condenser, thereby allowing restart of the reactor from a hot pressurized condition.
Attempting to establish condenser vacuum without seal steam from the hot condition by the mechanical vacuum pump may damage the turbine shaft seals.
Thus, the manual'byp[ ass _of the MSIV closure is. performed only when the
- . reactor is not operating at full power. In addition, the manual
- bypass, which is arnunciated in.the control room, has the following
! three pemissive conditions:
'A. When 'the keylocked manual s. witches located on the back cab'inets housing the MSIV logic are in the bypass position. One keylocked switch is'in each isolation logic string.
B. When the turbine stop valves are less than 90 percent full open, j
the four independent contacts of the turbine stop valve position i switch sensor relay of the RPS will trip.
I C. When the reactor is below 1045 pig. s l
Of these three permissives on the manual bypass of the MSIV closure, only the reactor pressure permissive.(item C above).does not have a safety function.
'This is also the only permissive that.is proposed for deletion. With the .
septoint at 1045 psig, which is the same setpoint for the high reactor pressure scram, the manual bypass can be performed at any operating reactor pressure provided the other two permissives are cleared..- When the manual bypass is activated _, plant protection is provided by those two other permissives by the I
normal scram and isolation signals, e.g., turbine stop valve position, low reactor water level, high steam flow, high steac tunnel temperature, and turbine building temperature, and by the annunciators.in the control room. ~
liminating the reactor pressure permissive does not affect the existing plant protection in any way. Therefore, elimination of the reactor pressure permissive on the manual bypass of the MSIV closure (item C above) does not constitute an unreviewed safety question or a significant hazards consideration, Appendix 1 (page Al-9) provides the results of the significant hazards review.
O 4-20
1 l
~
V.
SUMMARY
( ,
i This submittal provides an overview of the evaluation and the justification for the incorporation of the proposed modifications into the Plant Hatch
- design. Proposed Technical Specifications revisions are . supplied to cover l these modifications. None of these modifications constitute an unreviewed safety question or a significant hazards consideration; and the plant safety margin has not been reduced.
I t
r l
1 1
5-1
-_u-._-__.------.~._..--..___.._ - _ _ _ .-. .,,____ _ - _ _ _ - _ _ _ . _ . , _ . - _ , . __
/ VI. PROPOSED TECHNICAL SPECIFICATIONS REVISIONS The HNP-2 Technical Specifications (Appendix A to Operating License NPF-5) are proposed for revision as presented in this section. Table 6.1 provides the instructions for incorporating the revision (s).
e e
s l-00 _
6-1
TABLE 6.1 INSTRUCTIONS FOR INCORPORATING TECHNICAL SPECIFICATIONS REVISIONS If the Technical Specifications revisions are accepted as proposed, the HNP-2 Technical Specifications (Appendix A to Operating License NPF-5) should be incorporated as follows:
Deletions Insertions Applicable Safety Evaluation' ~
Ites (Pane) (Page) Report (SER) Sections (a) 1 2-4 2-4 ATTS, 4B.6 2 3/4 3-2 3/4 3-2 ATTS 3 3/4 3-7 3/4 3-7 ATTS 4 3/4 3-11 3/4 3-11 ATTS, I,B.3, 4B.3 5 3/4 3-12 3/4 3-12 ATTS, 4B.2, 4B.3, 4B.4 6 3/4 3-13 3/4 3-13 Editorial, ATTS, 4B.4, 4B.5 7 3/4 3-14 3/4 3-14 Editorial, ATTS, 4B.3, 4B.4, 4B.5 8 3/4 3-15 3/4 3-15 Editorial, 4B.6, 4B.8 9 3/4 3-16 3/4 3-16 4B.3, 4B.6, 4B.1 10 3/4 3-17 3/4 3-17 Editorial, 4B.1, 4B.3, 4B.6 11 3/4 3-18 3/4 3-18 ~
4B.6 12 3/4 3-19 3/4 3-19 4B.3 13 3/4 3-20 3/4 3-20 Editorial 14 3/4 3-21 3/4 3-21 ATTS, 4B.3, 4B.8 15 3/4 3-22. 3/4 3-22 Editorial, ATTS, 4B.3 16 -
3/4 3-23 3/4 3-23 ATTS 17 3/4 3-26 3/4 3-26 ATTS, 4B.5 18 3/4 3-27 3/4 3-27 Editorial, ATTS, 4R.S 19 3/4 3-28 3/4 3-28 4B.6 20 3/4 3-29 3/4 3-29 4B.1, 4B.6 21 3/4 3-31 3/4 3-31 ATTS 22 3/4 3-32 3/4 3-32 ATTS 23 3/4 3-34 3/4 3-34 ATTS 24 3/4 3-35 3/4 3-35 4B.1 25 3/4 3-36 3/4 3-36 -ATTS ~
.'_3[4'3-54'~,
~
26 3/43-51 ATTS,,_,_ _;_ -
l 27 3/4 4-18_ _- 3/4 4-18 4B.6 , ;.
28 3/4 6-41 3/4 6-41 4B.3 29 B 3/4 3-6 B 3/4 3-6 Editorial, ATTS, 4B.6 l
o a. 1. ATTS refers to proposed revisions already presented in refe rence 1.
l 2. 4B.1 to 4B.8 refers to justifications presented in section 4B of this submittal.
- 3. Editorial refers to a typographical error. The proposed revision j corrects the error; no additional justification is necessary.
t,, ,- -- . _ . . _ . . . - - - - __ . - - - _ - .-_ . - _
O O O g TABLE 2.2.1-1 s ,
I REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS h !
G FUNCTIONAL UNIT TRIP SETPOINT . ALLOWABLE VALUES
- 1. Intermediate Range Monitor, Neutron Flux-High 1 120/125 divisions 1 120/125 divisions l-(2C51-K601 A,B,C,D,E,F,G,H) of full scale of full scale F o
- 2. Average Power Range Monitor:
(2C51-K605 A,B,C,D,E,F)
- a. Neutron Flux-Upscale, 15% 1 15/125 divisions i 20/125 divisions of full scale of full scale j b. Flow Referenced Simulated Thermal 5 (0.66 W + 51%), 1 (0.66 W + 54%), -
Power-Upscale with a maximum with a maximum i 113.5% of RATED 1 115.5% of RATED j THERMAL POWER THERMAL POWER
'f" c. Fixed Neutron Flux-Upscale, 118% ~< 118% of RATED ~< 120% of RATED ,.
THERMAL POW 2R THERMAL POWER i
2
- 3. Reactor Vessel Steam Dome Pressure - High 5 1054 psig i 1054 psig l i (2B21-N678 A,B,C,D) l i 4. Reactor Vessel Water Level - Low (Level 3) ~> 8.5 inches above -> 8.5 inches above ! f (2B21-N680 A,B,C,D) instrument zero* instrument zero*
, 5. Main Steam Line Isolation Valve - Closure 1 10% closed i 10% closed i j (NA)
- 6. Main Steam Line Radiation - High 5 3 x filli power 1 3 x full power (2D11-K603A,B,C,D) background background
- 7. Drywell Pressure - High 1 1.85 psig i 1.85 psig l (2C71-N650A,B,C,D)
- See Bases Figure B 3/4 3-1.
O O TABLE 3.3.1-1 O :
I REACTOR PROTECTION SYSTEM INSTRUMENTATION APPLICABLE MINIMUM NUMBER .
OPERATIONAL OPERABLE CHANNELS h FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a) ACTION l!
9' I
h (2051-K601, A, B, C, D, E, F, G, H) a N a. Neutron Flux - I!igh 2(*}, 5(b) 3 3 2
3,kb) 2
- b. Inoperative 2,5 3 1 3,4 .2 2
- 2. Average Power Range Monitor:
(2C51-K605 A, B, C, D, E, F) i
- a. Neutron Flux - Upscale, 15% 2, 5 2 1
- b. Flow Referenced Simulated >
y Thermal Power - Upscale 1 2 3-
- c. Fixed Neutron Flux -
[ Upscale, 118% 1 2 3
- d. Inoperative 1, 2, 5 2 4 -
- e. Downscale 1 2 3 s
- f. LPRM 1,2,5 (d) NA
, A, B, C, D) l l
- 4. Reactor Vessel Water Level - .
I Low (Level 3) (2B21-N680 A, B, C, D) 1, 2 2(3 ' p) 5 l
l
- 5. Main Steam Line Isolation Valve -
Closure (NA) 1( 4 3
- 6. Main Steam Line Radiation - High 1, 2(*) 2 6 (2Dll-K603 A, B, C, D)
- 7. Drywell Pressure - High 1, 2(8) 2 5 (2C71-N650 A, B, C, D) l
, -,m , - m -_m.__ u
_ -- _ . - . . . - u- *- 4 a
O O O :.
}.
1 1 -
TABLE 4.3.1-1 (
f h REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS a.
I C h
H CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH N FUNCTIONAL UNIT CHECK TEST CALIBRATION (*) SURVEILLANCE REQUIRED
- a. Neutron Flux - High D S/U(b)(c) R 2 D W R 3,4,5
- b. Inoperative NA -W NA 2,3,4,5
- 2. Average Power Range Monitor.
- y a. Neut.ron Flux - Upscale, 15% S S/U(b)(c) , W(d) S/U(b) , W(d) 2 Y 5 g b. Flow Referenced Simulated S S/U(b) ,W W(e)(f) , SA 1
] 4 Thermal Power - Upscale i c. Fixed Neutron Flux - Upscale, S S/U(b) ,W W(e) , SA 1 118% ;
- d. Inoperative NA W NA ,
1, 2, 5 1
- e. Downscale NA W NA 1
- f. LPRM D NA (g) 1, 2, 5 1
l 3. Reactor Vessel Steam Dome ,
S M R 1, 2 l l Pressure - High I u j 4. Reactor Vessel Water Level - S M R 1, 2 l Low (Level 3) i i S. Main Steam Line Isolation Valve -
Closure NA M R( .# 1 f;
)
- 6. Main Steam Line Radiation - High D W(I) .R(N 1, 2
- 7. Drywell Pressure - High S M R 1, 2 l ;
]
- 8. Scram Discharge Volume Water NA M R( ) 1, 2, 5
- Level - High r - -- __
O~ n TABLE \~ ,d.2-1 O\ .
ISOLATION ACTUATION INSTRUMENTATION :.
i, g VALVE GROUPS MINIMUM NUMBER APPLICABLE i s OPERATED BY OPERABLE CHANNELS OPERATIONAL 1, TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b)(c) CONDITION ACTION {j h}
c: i:
M 1. PRIMARY CONTAINMENT ISOLATION. '
" a. Reactor Vessel Water Level -
- 1. Low (Level 3) 2, 6, 10, 2 1,2,3 20 (2B21-N680 A, B, C, D) 11, 12
- 2. Low-Low (Level 2) 5, #,
- 2 1,2,3 20 (2B21-N682 A, B, C, D) '
- 3. Low-Low-Low (Level 1) 1 2 1,2,3 20 (2B21-N681 A, B, C, D)
- b. Drywell Pressure - High 2, 6, 7, 10, 2 1,2,3 20 l (2C71-N650 A, B, C, D) 12, #,
- ua c. Main Steam Line 20 1. Radiation - High 1, 12, #, (d) 2 1,2,3 21 ua (2D11-K603 A, B, C, D) b
- 2. Pressure - Low 1 2 -
1 22 (2B21-N015 A, B, C, D)
- 3. Flow - High 1, # 2/line 1, 2, 3 21 (2B21-N686 A, B, C, D)
(2B21-N687 A, B, C, D)
(2B21-N688 A, B, C, D)
(2B21-N689 A, B, C, D)
- d. Main Steam Line Tunnel Temperature - High 1 2/line(* 1, 2, 3 21 (2B21-N623 A, B, C, D)
(2B21-N624 A, B, C, D)
(2B21-N625 A, B, C, D) <
(2B21-N626 A, B, C, D)
- e. Condenser Vacuum - Lot' 1 2 l', 2,( ,3( } 23 (2B21-N056 A, B, C, D) .
- f. Turbine Building Area Temperature - High 1 2
) 1,2,3 21 (2U61-R001, 2U61-R002, 2U61-R003, 2U61-R004) ,
, 'n k f~)l C b ,
TABLE 3.3.2-1 (C ntinued)
ISOLATION ACTUATION INSTRUMENTATION .
$! VALVE GROUPS MINIMUM NUMBER APPLICABLE E! OPERATED BY OPERABLE CHANNELS OPERATIONAL i-7 TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b)(c) CONDITION ACTION.
5 y 2. SECONDARY CONTAINMENT ISOLATION w
- a. Reactor Building Exhaust Radiation - High 6, 10, 12,
- 2 1,2,3,5 and** 24 (2D11-K609 A, B, C, D) t,
- b. Drywell Pressure - High 2, 6, 7, 10, 2 1,2,3 24 (2C71-N650 A, B, C, D) 12, #, *
- c. Reactor Vessel Water 4
Level - Low Low (Level 2) 5,#,* 2 1, 2, 3 24 (2B21-N682 A, B, C, D) s
- d. Refueling Floor Exhaust Y Radiation - High , 6, 10, 12, #,
- 2 1,2,3,5 and** 24 i w (2D11-K611 A, B, C, D) i'
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
~
- a. A Flow - High (2G31-N603 A, B) 5 1 1,2,3 25
- b. Area Temperature - High 5 1 1,2,3 25 (2G31-N662 A, D, E, H, J, M) l
- c. Area Ventilation A Temp. - High 5 1 1,2,3 25 (2G31-N663 A, D, E, H, J, M; 2G31-N661 A, D, E, H, J, M; 2G31-N662 A, D, E, H, J, M),
- d. GLCS Initiation (NA) 5(8) NA 1,2,3 25
- e. Reactor Vessel Water Level - Low Low 5, # ,
- 2 1,2,3 25
{ (Level 2) (2B21-N682 A, B, C, D) i i
O O~
TABLE 3.3.2-1 (Crntinued)
O ISOLATION ACTUATION INSTRUMENTATION s
9 VALVE GROUPS HINIMUM NUMBER APPLICABLE E: OPERATED BY OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION SIGNAL (a) PER TRIP SYSTEM (b)(c) _ CONDITION ACTION
- 4. HIGil PRESSURE COOLANT INJECTION SYSTEM ISOLATION
- b. HPCI Steam Supply Pressure -
Low (2E41-N658 A,B,C,D) 3,8 2 1,2,3 26 l
- d. IIPCI Pipe Penetration Room u Temperature - High (2E41-N671 A, B) 3 1 1,2,3 26 l
- e. Suppression Pool Area Ambient Temperature-liigh (2E51-N666 C, D ) 3 1 1,2,3 26
- g. Suppression Pool Area Temperature Timer Relays (2E41-M603 A, B) 3(g 1 1,2,3 26
- h. Emergency Area Cooler Temperature-Ifigh (2E41-N670 A, B) 3 1 1,2,3 26 l
- i. Drywell Pressure-High (2E11-N694 C, D) 8 1 1,2,3 26
- j. Logic Power Monitor (2E41-K1) NA(h) 1 1, 2, 3 27
TABLE 3.3.2-1 ntinue.Q ISOLATION ACTUATION INSTRUMENTATION i:
VALVE GROUPS HINIMUM NUMBER APPLICABLE i OPERATED BY OPERABLE CHANNELS OPERATIONAL
@ TRIP FUNCTION SIGNAL (aL PER TRIP SYSTEM (b)(c) CONDITION ACTION
- i 5. REACTOR CORE ISOLATION h COOLING SYSTEM ISOLATION l E i y a. RCIC Steam Line Flow-High 4 1 1,2,3 26 i y (2E51-N657 A,B)
- b. RCIC Steam Supply Pressure -
Low (2E51-N658 A, D, C, D) 4, 9 2 1,2,3 26 J
- c. RCIC Turbine Exhaust Diaphragn Pressure - High 4 2 1,2,3 26 (2E51-N685 A, B, C, D) u d. Emergency Area Cooler Temperature -
e High (2E51-N661 A, B) 4 1 1,2,3 26 w
[ e. Suppression Pool Area Ambient ;
- Temperature-High 4 1 1,2,3 26 (2E51-N666 A, B ) !
- g. Suppression Pool Area Temperature Timer Relays (2E51-M602 A, B) 4(g) 1 1,2,3 26
- h. Drywell Pressure - High (2E11-N694 A, B) 9 1 1,2,3 26
- i. Logic Power Monitor (2E51-K1) NA(h) 1 1,2,3 27
- 6. SHUTDOWN COOLING SYSTEM ISOLATION
- a. Reactor Vessel Water Level-Low (Leveli 6, 10, 11, 2 2 3, 4, 5 26 3)(2B21-N680 A, B, C, D) 12
- b. Reactor Steam Dome Pressure-High 11 1 1,2,3 28 (2B31-N679 A, D)
,-.L TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION ACTION ACTION 20 -
Be in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
ACTION 21 -
Be in at least STARTUP with the main steam line isolation valves l.
closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT SHUTDOWN within 6
, hours and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
i ACTION 22 -
Be in at least STARTUP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
I
' ACTION 23 -
Be in at least STARTUP with the Group 1 isolation valves closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or in at least HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 24 -
Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas treatment system operating within one hour.
. ACTION 25 -
Isolate the reactor water cleanup system.
ACTION 26 Close the affected system isolation valves and declare the affected system inoperable.
! ACTION 27 -
Verify power availability to the bus at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or close the affected system isolation valves and declare the affected system inoperable.
ACTION 28 -
Close the shutdown cooling supply and reactor vessel head spray ,
isolation valves unless reactor steam dome pressure < 145 psig. I NOTES
} # Actuates operation of the main control room environmental control j system in the pressurization mode of operation.
,
- Actuates the standby gas treatment system.
- When handling irradiated fuel in the secondary containment.
- a. See Specification 3.6.3, Table 3.6.3-1 for valves in each valve group.
- b. A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for
~
3 required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.
- c. With a design providing only one channel per trip system, an inoperable channel need not be placed in the tripped condition where this would cause the Trip Function to occur. In these cases, the inoperable channel shall be restored to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or the ACTION required by Table 3.3.2-1 for that Trip Function shall be taken.
- d. Trips the' mechanical vacuum pumps.
- f. May be bypassed with all turbine stop valves closed. !
f i -
- h. Alarm only.
- i. Adjustable up to 60 minutes.
HATCH - UNIT 2 3/4 3-15
O O O TA8LE 3.3.2-2 l'
h ISOLATION AC'lUATION INSTRUMENTATION-SETPOINTS a .
& ALLOWABLE
% TRIP FUNCTION TRIP SETPOINT VALUE i H N 1. PRIMARY CONTAINMENT ISOLATION
. 1. Low (Level 3) > 8.5 inches * > 8.5 inches *
- 2. Low Low (Level 2) 5 -55 inches
- i -55 inches *
- 3. Low Low Low (Level 1) [-121.5 inches * [-121.5 inches *
- b. Drywell Pressure - High < 1.85 psig i 1.85 psig I
- c. Main Steam Line w 1. Radiation - High 5 3 x full power background < 3 x full power I
s w 2. Pressure - Low 'l 825 psig 1 825 psig ,
i 3. Flow - High 5 138% rated flow $ 138% rated flow I os
- d. Main Steam Line Tunnel i 4 Temperature - High < 194 F < 194 F L j e. Condenser Vacuum - Low 1 7" Hg vacuum 1 7" Hg vacuum t
- f. Turbine Building Area Temp.-High < 200 F $ 200 F ,
- 2. SECONDARY CONTAINMENT ISOLATION I
I
- a. Reactor Building Exhaust !
j Radiation - High 5 60 mr/hr i 60 mr/hr l 3
- b. Drywell Pressure - High 1 1.85 psig i 1.85 psig
- c. Reactor Vessel Water Level - Low Low (Level 2) 1 -55 inches
- 1 -55 inches
- l 1
1 d. Refueling Floor Exhaust ;
i . Radiation - High < 20 mr/hr i 20 mr/hr I
- See Bases Figure B 3/4 3-1.
g i
i
( O % i V (V ;
TABLE 3.3.2-2 (Continued) !i f ii g ISOLATION ACTUATION INSTRUMENTATION SETPOINTS
?
E s ALLOWABLE' TRIP FUNCTION TRIP SETM) INT VALUE ,,
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. A Flow - High $ 79 gpm $ 79 gpm
.b. Area Temperature-High $ 124 F $ 124*F
- c. Area Ventilation 3 Temperature - High $ 67 F $ 67 F
- d. SLCS Initiation NA NA w e. Reactor Vessel Water Level-Low Low -> -55 inches * ~> -55 inches
- D (Level 2) l
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION -
u
- a. HPCI Steam Line Flow-High $ 307% of rated flow 1 307% of rated flow l.
- b. HPCI Steam Supply Pressure - Low > 100 psig > 100 psig
- d. HPCI Pipe Penetration Room
- Temperature - High $ 169*F $ 169 F
- e. Suppression Pool Area Ambient Temperature-High 1 169 F $ 169 F i
- f. Suppression Pool Area AT - High < 42.5 F < 42.5"F l
- g. Suppression Pool Area Temperature Timer Relays NA NA
- h. Emergency Area Cooler Temperature -
High $ 169 F $ 169 F
- i. Drywell Pressure - High 5 1.85 psig 1 1.85 psig
- j. Logic Power Bus Monitors NA NA
- See Ba'es s Figure B 3/4 3-1.
l3 TABLE 3.3.2-2 (Continued)
- M -
4: ISOLATION ACTUATION INSTRUMENTATION SETPOINTS S
e ,
H .
y ALLOWABLE !
TRIP FUNCTION '
TRIP SETPOINT VALUE I I
- 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION
- a. RCIC Steam Line Flow - liigh i 312% of rated flow $ 312% of rated flow l j i
- b. RCIC Steam Supply Pressure - Low 1 60 psig 1 60 psig I
u, d. Emergency Area Cooler Temperature-High i 169 F 1 169 F 20 m e. Suppression Pool Area Ambient Temperature i 169 F $ 169 F ,
J, liigh i co
- f. Suppression Pool Area AT - High 5 42.5 F $ 42.5 F
- g. Suppression Pool Area Temperature Timer Relays NA NA
- h. Drywell Pressure - High 1 1.85 psig $ 1.85 psig l 1 i. Logic Power Monitor NA NA
- 6. SHUTDOWN COOLING SYSTEM ISOLATION
- a. Reactor Vessel Water Level - Low 1 8.5 inches
- 1 8.5 inches *
(Level 3)
- b. Reactor Steam Dome Pressure - High 5 145 psig $ 145 psig s
! *See Bases Figure B 3/4 3-1.
A
(_,/ TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
- 1. PRIMARY CONTAINMENT ISOLATION
- 1. Low (Level 3) < 13*
- 2. Low Low (Level 2) 513*
- 3. Low Low Low (Level 1), except MSIVs 5 13*
- b. Drywell Pressure - High 5 132
- 1. Radiation - High*** < 1.0**
,e 2. Pressure - Low < 13*
- 3. Flow - High 31.0**
- 4. Reactor Vessel Water Level - Low Low Low 5 1.0**
(Level 1) I
- d. Main Steam Line Tunnel Temperature - High 5 13*
Condenser Vacuum - Low O, e. NA
- f. Turbine Building Area Temperature - High NA
- 2. SECONDARY CONTAINMENT ISOLATION
- a. Reactor Building Exhaust Radiation - High*** 5 13*
- b. Drywell Pressure - High 5 13*
- c. Reactor Vessel Water Level - Low Low (Level 2) 5 13* l
- d. Refueling Floor Exhaust Radiation - High*** -< 13*
l l
- The isolation actuation instrucentation response time shall be measured and recorded as a part of the ISOLATION SYSTEM RESPONSE TIME. Response time specified is diesel generator start delay time assumed in accident analysis.
- Isolation actuation instrumentation response time.
- Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first l electronic component in the channel.
- Times to be added to valve movement times shown in Tables 3.6.3-1, 3.6.5.2-1
()
and 3.9.5.2-1 to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
i HATCH - UNIT 2 3/4 3-19
~ ~ ^ ~ " ~ ^
k TABLE 3.3.2-3 (Continued)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION l
- a. A Flow - High 1 13*
- b. Area Temperature - High 1 13*
- c. Area Ventilation Temperature AT - High 1 13*
- d. SLCS Initiation NA
- e. Reactor Vessel Water Level-Low Low (Level 2) i 13* !
i
- 4. HIGH PRESSURE COOLANT INJECTION SYSTEM ISOLATION
- a. HPCI Steam Line Flow-High 1 13*
, b. HPCI Steam Supply Pressure - Low 1 13*
, d. HPCI Pipe Penetration Room Temperature -
High NA
- e. Suppression Pool Area Ambient Temp. - High NA I f. Suppression Pool Area AT - High NA
- g. Suppression Pool Area Temp. Timer Relays NA A h. Emergency Area Cooler Temperature - High - NA
! \j i. Drywell Pressure - High 1 13*
l j. Logic Power Monitor NA j 5. REACTOR CORE ISOLATION COOLING SYSTEM ISOLATION l a. RCIC Steam Line Flow - High - NA
- b. RCIC Steam Supply Pressure - Lov NA
- c. RCIC Turbine Exhaust Diaphragn; Pressure - High NA
- d. Emergency Area Cooler Temperature - High NA
- e. Suppression Pool Area Ambient Temp. - High NA
- f. Suppression Pool Area AT - High NA
- g. Suppression Pool Area Temperature Timer Relays NA
- h. Drywell Pressure - High 1 13*
- i. Logic Power Monitor NA
- 6. SHUTDOWN COOLING SYSTEM ISOLATION
- a. Reactor Vessel Water Level - Low (Level 3) NA l
- b. Reactor Steam Dome Pressure - High NA O
HATCH - UNIT 2 3/4 3-20
q p . i
\ .
TABLE 4.3.2-1
]
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENE {'
W g '
CHANNEL OPERATIONAL y: CHANNEL. FUNCTIONAL CHANNEL -
CONDITIONS IN WHICH TRIP FUNCTION CHECK TEST ' CALIBRATION. SURVEILLANCE REQUIRED
- 1. PRIMARY CONTAINMENT ISOLATION
- 1. Low (Level 3) S M R 1, 2, 3 .
- 2. Low Lcw (Level 2) S M R 1, 2, 3
- 3. Low Low Low'(Level 1) S H R -
1,2,3
! b. Drywell Pressure - High S H R 1, 2, 3 l i
- c. Main Steam Line !
- 1. Radiation - High D W(*} R 1, 2, 3 ;
- 2. Pressure - Low NA M Q 1 ;
w 3. Flow - High S M R 1, 2, 3 l N l w d. Main Steam Line Tunnel l
~
- /, Temperature - High S M R 1, 2, 3 I
- e. Condenser Vacuum - Low NA M Q 1, 2#, 3#
- f. Turbine Building Area Temp. -
liigh NA M R 1, 2, 3
- 2. SECONDARY CONTAINMENT ISOLATION i
! a. Reactor Building Exhaust !
Radiation - High D M(* R- 1, 2, 3, 5 and
- ft
- b. Drywell Pressure - High S M R 1,2,3 ;
- c. Reactor Vessel Water Level - l l Low Low (Level 2) S M R 1, 2, 3 1
- d. Refueling Floor Exhaust I
] Radiation.- High D M(*) Q 1, 2, 3, 5 and
- l
- When handling irradiated fuel in the secondary containment. ]
! #May be bypassed with all turbine stop valves closed. I aInstrinnent alignment uing a standard current source.
O O O 1 l
i TABLE 4.3.2-1 (Continued) I g ISOLATION ACTUATION INSTRUMENTATION SURVEILIANCE REQUIREMENTS s :
O CHANNEL OPERATIONAL E: CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH g
$ TRIP FUNCTION CHECK TEST CALIBRATIO_N_ SURVEILLANCE REQUIRED
- 3. REACTOR WATER CLEANUP SYSTEM ISOLATION
- a. A Flow - High D H R 1, 2, 3
- b. Area Temperature - High S M R 1, 2, 3 l ;
i t
- c. Area Ventilation A g i
Temperature - liigh S M R 1,2,3 1
- d. SLCS Initiation NA R NA 1,2,3 it e. Reactor Vessel Water Level - S M R 1,2,3 Low Low (Level 2)
Y y 4. IIIGli PRESSURE COOLANT INJECTION SYSTEM ISOLATION l a. IIPCI Steam Line Flow-liigh S M R 1,2,3 !
- b. IIPCI Steam Supply Pressure-Low S M R 1, 2, 3
- c. IIPCI Turbine Exhaust S M R 1,2,3
- Diaphragm Pressure - High
- d. IIPCI Pipe Penetration Room Temperature - High S M R 1, 2, 3 l
- e. Suppression Pool Area Ambient
- Temp. - liigh S M R 1, 2, 3 l.
- f. Suppression Pool Area AT -
Iligh S M R 1,2,3 l
- g. Suppression Pool Area Temp.
Timer Relays NA SA R 1,2,3
- h. Emergency Area Cooler Temp. -
liigh S M R 1,2,3 l
- i. Drywell Pressure - High S M R 1,2,3
- j. Logic Power Monitor NA R NA 1,2,3
O O O :
TABLE 4.3.2-1 (Continued) s ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4
- g. ,
)
i f
h.
e- CHANNEL.- OPERATIONAL
"' CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN WHICH TRIP FUNCTION ,
CHECK _
TEST CALIBRATION , SURVEILLANCE REQIITRED
'f
,: ;i.
, 5. REACTOR CORE ISOLATION , .
COOLING SYSTEM ISOLATION '
, a. RCIC Steam Line Flow-High S- M R 1, 2, 3
- b. RCIC Steam Supply Pressure- S' M R 1,2,3 Low
- d. Emergency Area Cooler S M R 1, 2, 3 l 1 Temperature - High R
- e. Suppression Pool Area S M R 1, 2, 3 l Ambient Temperature-High Y f. Suppression Pool Area AT - S M R 1, 2, 3 U High
- g. Suppression Pool Area Temp. Timer Relays NA SA R 7,2,3
- h. Drywell Pressure - High S M R 1, 2, 3 l
, i. Logic Power Monitor NA R NA 1, 2, 3 i
- 6. SilUTDOWN COOLING SYSTEM ISOLATION
- a. Reactor Vessel Water Level - S M R 3, 4, 5 Low (Level 3)
- b. Reactor St. cam Dome S M R 1,2,3 l Pressure - High
~
L
, ~ - . , - ~ . , ..-
O- O O i j TABLE 3.3.3-1 :
i
@ EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION '
h
@ MINIMUM NUMBER APPLICABLE i y OPERABLE CHANNELS OPERATIONAL !
t y TRIP FUNCTION PER TRIP SYSTEM CONDITIONS i
! I j 1. CORE SPRAY SYSTEM
] a. Reactor Vessel Water Level - Low Low Low (Level 1) 2 1,2,3,4,5 ,
(2B21-N691A,B,C,D)
! b. Drywell' Pressure - High (2E11-N694 A,B,C,D) 2 1,2,3 l c. Reactor Steam Dome Pressure - Low (Injection Permissive) ,,
i (2B21-N690A,B,C,D) 2 1,2,3,4,5 l .
, d. Logic Power Monitor (2E21-K1A,B) ,
1/ bus (,) 1,2,3,4,5
- 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM .
- a. Drywell Pressure - High (2E11-N694A,B,C,D) 2 1,2,3
- b. Reactor Vessel Water Level - Low Low Low (Level 1) 2 1,2,3,4*,5*
Y (2B21-N691A,B,C,D)
$$ c. Reactor Vessel Shroud Level (Level 0) - High (Drywell Spray l Permissive) (2B21-N685A, B) 1 1,2,3,4*,5*
l d. Reactor Steam Dome Pressure - Low (Injection Permissive) >
j (2B21-N690A,B,C,D) 2 1,2,3,4*,5* -
l e. Reactor Steam Dome Pressure - Low (Recirc. Discharge Valve Permissive) (2B21-N641B,C and 2B21-N690E,F) 2 1,2,3,4*,5* l
- f. RHR Pump Start - Time Delay Relay 1/ pump 1,2,3,4*,5*
i 1) Pump A (2E11-K70A, 2E11-K125B)
- 2) Pump B (2E11-K70B, 2E11-K125A)
- 3) Pump C (2E11-K75B)
t >
1
- Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.
i d
(a) Alarm only. When inoperable, verify power availability to the bus at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j or declare the system inoperable.
1 i
O O O TABLE 3.3.3-1 (Ccntinued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION x
8 MINIMUM NUMBER APPLICABLE h
H OPERABLE CHANNELS OPERATIGNAL g TRIP FUNCTION PER TRIP SYSTEM CONDITIONS # l n
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM s <
- a. Reactor Vessel Water Level - Low Low (Level 2) 2 1,2,3 (2B21-N692 A,B,C,D) .
- b. Drywell Pressure - High (2E11-N694 A,B,C,D) ,3
1, 2, 3
- d. Suppression Chamber Water Level-High (2E41-N662B,D) 2(b)(c) 2 1, 2, 3 l
i
! e. Logic Power Monitor (2E41-K1) I
) 1, 2, 3
- f. Reactor Vessel Water Level-High (Level 8) (2B21-N693 B,D) 2 1,2,3 l 4
- a. Drywell Pressure - High (Permissive) (2E11-N694A,B,C,D) 2 1,2,3 y b. Reactor Vessel Water Level - Low Low Low (Level 1)
N (2B21-N691 A,B,C,D) 2 1,2,3 ,
- d. Reactor Vessel Water Level-Low (Level 3) (Permissive) 1 1,2,3
- (2B21-N695A,B)
- c. Core Spray Pump Discharge Pressure - High (Permissive) l l (2E21-N655A,B; 2E21-N652A,B) 2 1,2,3 I l; RHR (LPCI MODE) Pump Discharge Pressure - High (Permissive) f.
(2E11-N655A,B,C,D; 2Z11-N656A,B,C,D) 2/loog) 1, 2, 3 l
- g. Control Power Monitor (2B21-K1A,B) 1/ bus 1, 2, 3 l
i 5. LOW LOW SET S/RV SYSTEM i a. Reactor Steam Dome Pressure - High (Permissive)
(2B21-N620A,B,C,D) 2 1, 2, 3 (a) Alarm only. When inoperable, verify power availability to the bus at least once per 12 I
hours or declare the system inoperable.
(b) Provides signal to HPCI pump suction valves only.
l (c) When either channel of the automatic transfer logic ie inoperable, align HPCI pump suction to the suppression pool.
1
il
O O O TABLE 3.3.3-2 H EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS
- i 9
E:
z y ALLOWABLE l g TRIP FUNCTION TRIP SETPOINT VALUE ;
i
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Water Level - Low Low Low (Level 1) 1 -121.5 inches
- 1 -121.5 inches
- l l
- b. Drywell Pressure - High i 1.85 psig i 1.85 psig Reactor Steam Dome Pressure - Low 1 422 psig** 1 422 psig**
- c. '
- d. Logic Power Monitor NA NA i 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM l
I
{h a. Drywell Pressure - High 1 1.85 psig i 1.85 psig 4
- b. Reactor Wessel Water Level - Low Low Low (Level 1) > -121.5 inches * > -121.5 inches *
! Y '
- c. Reactor Vessel Shroud Level (Level 0) - High 5 -207 inches * ~5 -207 inches *
! S$ d. Reactor Steam Dome Pressure-Low ,- 422 psig** 1 422 psig**
- e. Reactor Steam Dome Pressure-Low 1 325 psig 1 325 psig
) f. RHR Pump Start - Time Delay Relay
]
- 1) Pumps A, B and D 10 1 I seconds 10 1 I seconds J 2) Pump C 0.5 1 0.5 seconds 0.5 1 0.5 seconds
- g. Logic Power Monitor NA NA i
i, d
See Bases Figure B 3/4 3-1. !
l **This trip function shall be less than or equal to 500 psig. l i
s 4 i
,' I n
- O O O '
,i TABLE 3.3.3-2 (Continued) <
) 'h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS 4
g i
N ALIX)WABLE I w TRIP FUNCTION TRIP SETPOINT VAIJJE
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM 9 a. Reacter Vessel Water Level - Low Low (Level 2) 1 -55 inches
- 1 -55 inches *
- b. Drywell Pressure-High i 1.85 psig i 1.85 psig
- c. Condensate Storage Tank Level - Low 1 0 inches ** 1 0 inches ** i
- d. Sappression Chamber Water Level - High i 33.2 inches i 33.2 inches I i e. Logic Power Monitor NA NA l
]
- f. Reactor Vessel Water Level-High (Level 8)* i 56.5 inches i 56.5 inches 1
t', a. Drywell Pressure-High i 1.85 psig i 1.85 psig 8'
- b. Reactor Vessel Water Level - Low Low Low (Level 1) > -121.5 inches * > -121.5 inches
- i y c. ADS Timer < 120 secoads < 120 seconds
, y d. Reactor Vessel Water Level-Low (Level 3) [8.5 inches * [8.5 inches
- l e. Core Spray Pump Discharge Pressure - High 1 130 psig 1 130 psig
- f. RHR (LPCI MODE) Pump Discharge Pressure - High 1 105 psig 1 105 psig j g. Control Power Monitor NA NA i
1
- 5. LOW LOW SET S/RV SYSTEM
- a. Reactor Steam Dome Pressure - High i 1054 psig i 1054 psig l
i l
l
- See Bases Figure B 3/4 3-1.
l ** Equivalent to 10,000 gallons of water in the CST.
O O O TABLE 4.3.3-1
% EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS s
h CHANNEL OPERATIONAL
@ CHANNEL FUNCTIONAL CHANhEL CONDITIONS IN WHICH .
[j TRIP FUNCTI,0N CHECK TEST CALIBRATION SURVEILLANCE REQUIRED w i
- 1. CORE SPRAY SYSTEM
- a. Reactor Vessel Water Level - ,
Low Lcw Low (Level 1) S M R 1, 2, 3, 4, 5
- b. Drywell Pressure - High S M R 1, 2, 3
- c. Reactor Steam Dome Pressure - Low S M R 1,2,3,4,5 l
- d. Logic Power Monitor NA R NA 1,2,3,4,5 u, 2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM 2:
u, a. Drywell Pressure - High S M R 1, 2, 3 d, b. Reactor Vessel Water Level - l Lew Low Low (Level 1) S M R 1, 2, 3, 4* , 5* I
- c. Reactor Vessel Shroud Level (Level 0) - High S M R 1, 2, 3, 4*, 5* l
- d. Reactor Steam Lame Pressure - Low S M R 1, 2, 3, 4*, 5*
- e. Reactor Steam Dome Pressure - Low S M R 1, 2, 3, 4*, 5*
- f. RHR Pump Start-Time Delay Relay NA NA R 1, 2, 3, 4*, 5*
- g. Logic Power Monitor NA R NA 1, 2, 3, 4*, 5*
- Not applicable when two core spray system subsystems are OPERABLE per Specification 3.5.3.1.
e
.a
(~ . (~~'\
G b Q)
TABLE 4.3.3-1 (Continued) I h EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS O
g CHANNEL OPERATIONAL i s CI'ANNEL FUi.'CTIONAL CHANNEL CONDITIONS IN WHICH
] TRIP FUNCTIGN CHECK TEST CALIBRATION SURVEILLANCE REQUIRED #
- 3. HIGH PRESSURE COOLANT INJECTION SYSTEM
Low Low (Level 2) S M R 1, 2, 3
- b. Drywell Pressure-High S M R 1, 2, 3
- c. Condensate Storage Tank Level -
Low NA H Q 1, 2, 3
- d. Suppression Chamber Water Level - High S M R 1, 2, 3 l
- e. Logic Power Monitor NA R NA 1,2,3
- f. Reactor Vessel Water Level-High S M R 1,2,3 R
s~
(Level 8)
- a. Drywell Pressure-High S M R 1,2,3 I
l Low Low Low (Level 1) S H R 1, 2, 3 I
- c. ADS Timer NA NA R 1,2,3
- d. Reactor Vessel Water Level - Low S H R 1, 2, 3 (Level 3) ,
- e. Core Spray Pump Discharge Pressure - High S M R 1,2,3 l
Pressure - thgh S H R 1, 2, 3 i
- g. Control Power Monitor hA R NA 1, 2, 3
- 5. LOW LOW SET S/RV SYSTEM
- a. Reactor Steam Dome Pressure - l High S H P 1,2,3 l
O O O TABLE 3.3.4-1 I, M REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUtiiENTATION Y -
E MINIMUM NUMBER OF M OPERABLE CHANFELS y _ FUNCTIONAL UNITS PER TRIP SYSTEM- ,
- a. Reactor Vessel Water Level - Low Low (Level 2) 7. ! ,
(2B21-N692 A, B, C, D)
{
i.
t :
I i i a h i~ .
I 4
1 i w
! Y i W l
i i
i j
J .
l i
I l
4
! i l
w m , - -
~
i O O O >
TABLE 3.3.4-2 1
g
- N REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION ,
I i i E '
l M ALLOWABLE 4 w FUNCTIONAL UNITS TRIP SETPOINT VALUE a
t
! a. Reactor Vessel Water Level - Low Low (Level 2) > -55 inches * > -55 inches
- l 4
l i
LJ i I U1 l t
i 1 .
1
- See Bases Figure B 3/4 3-1.
1 i
O O O '
TABLE 4.3.4-1 5
- h, REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E !
- E CHANNEL
! w CHANNEL FUNCTIONAL CHANNEL !
l FUNCTIONAL UNITS CHECK TEST -C%LIBRATION I i
i
! a. Reactor Ve :e1 Water Level - Low Low S M R ,
(Level 2) -
l l 1
i !
l l
1 i
, u t w i
I
]
a :
i l I
b l
i
3 O (d (V3 ~
/
h '
! $ TABLE 3.3.6.4-1
?
E POST-ACCIDENT MONITORING INSTRUMENTATION 4
w MINIMUM CHANNELS
, INSTRUMENT OPERABLE
- 1. Reactor Vessel Pressure (2B21-R623 A, B) 2
! 3. Suppression Chamber Water Level (2T48-R622 A, B) 2
p 6. Drywell Pressure (2T48-R608, 2T48-R609) 2'
- 8. Post-LOCA Gamma Radiation (2D11-K622 A, B, C, D) 2 j 9. Dryweil H 2
-0 2
Analyzer (2P33-R601 A, B) 2 i
10.a) Safety / Relief Valve Position Primary Indicator (2B21-N301 A-H and K-M)
- b) Safety / Relief Valve Position Secondary Indicator (2B21-N004 A-H and K-M) * '
j 1
l *If either the primary or secondary indication is inoperable, the torus temperature will be monitored at least once per shift to observe any unexplained temperature increases which might be indicative
- of an open SRV. With both the primary and secondary monitoring channels of an SRV inoperable, either verify that the S/RV is closed through monitoring the backup low low set logic position indicators (2B21-N302 A-H and K-M) at least once per shift or restore sufficient inoperable ,
channels such that no more than ene SRV has both primary and secondary channels inoperable within 7 days or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
J i
i i
l
~~ ~
- l
'% l REACTOR COOLANT SYSTEM REACTOR STEAM DOME LIMITING CONDITION FOR OPERATION 3.4.6.2 The pressure in the reactor steam dome shall be less than 1054 l psig.
APPLICABILITY: CONDITION 1* and 2*.
ACTION:
With the reactor steam dome pressure exceeding 1054 psig, reduce the pressure to less than 1054 psig within 15 minutes or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
y SURVEILLANCE REQUIREMENTS 3 4.4.6.2 The reactor steam dome pressure shall be verified to be less (V than 1054 psig at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- Not applicable during anticipated transients.
1 I
i i HATCH - UNIT 2 3/4 4-18
(
p]
\ CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)
- 3. Verifying a system flow rate of 4000 +0, -1000 cfm during system operation when tested in accordance with ANSI N510-1975.
- c. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accord-ance with Regulatory Position C.6.b of Regulatory Guide 1.52, Re-fision 1, July 1976, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 1, July 1976.
- d. At least once per 18 months by:
- 1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is < 6 inches Water Gauge while operating the filter train at a flow rate of 4000 +0, -1000 cfm.
s 2. Verifying that the filcer train starts and isolation k'_,) dampers open on.each of the following test signals:
- a. Drywall pressure-high,
- b. High rsdiation on the;
- 1) Refueling floor,
- 2) Reactor building. .
- c. Reactor Vessel Water Level-Low Low (Level 2). l
- 3. Verifying that the heaters dissipate 18.5 + 1.5 KW when tested in accordance with ANSI N510-1975.
(}/
s-HATCH - UNIT 2 3/4 6-41 t
80' T NOTE: SCALE IN INCHES ABOVE VESSEL ZERO G WATER LEVEL NOMENCLATUR1 HEIGHT A80VE 800 -- VESSEL ZERO NO. (INCHES) READING INSTRUMENT (8) 573.5 +56.5 BARTON 750 - ~ (7) 559 +42 G E/MAC' VESSEL _ (4) 549 +32 G E/MAC FLANGE , (3) 525.5 +8.5 ' BARTON (2) 462 -55 BARTON 700 -
(1) 305.5 -121.5 BARTON (0) - 310.0 -207 BARTON 650 - -
~~ ~
LINE f
. 600 --
577
+6 - ' "
+60 - . 56.5 60 -
460 - - .
8 55 (8) ! 8. (7) 42 HI ALARM 550 -r 54 G t4)
- 525.5(3) .yRi k (4) '32 LO ALARM 8.5 LOW (LEVEL 3) [
BOTTOM OF STE AM (3) 517 INSTRUMENT (31 }- 8.5
-DRYER SK8RT ~~
CONTRIBUTETO ADS
- FEED -- 484.5 5 - - SCRAM WATER ~
% -55 LOW LOW (LEVEL 2) g _ CORE INsTIATE HPCI,
- 482(2) SPRAY #
450 - - RCIC TRIP RE.
CIRC. PUMPS O 400 --_395,5{1) . -121.5 LOW LOW LOW (LEVEL 11 367 -150 --. . INITIATE RHR, C.S.,
START DIESEL AND 350 -6 352.56 o CONTRIBUTE TO A.D.S.
2/3 CORE CLOSE MSIV's
-207 HEIGHT -r 310(0)
PERMISSIVE 300 --
(LEVEL 0) ACTIVE FUEL 250 --
-317 - - 200 -= 208.56 REC lRC RECIRC
- 178Ji6 DISCHARGE SUCTION - 161.C -- NOZZLE NOZZLE 150 --
100 --
m 50 --
0--
9380-2 BASES FIGURE B 3/4 3-1
. RE CTOR VESSEL WATER [EVNLS B 3/4 3-6
- -- -..= - .
- , - , - , -na e6 - & am, -
- s A,,,_- , - - S.. -,_,, 3 m ,_wraAm m_w,a s ,14xwg4. s AM,,_ga 4, a g, m,,A e
I i
i APPENDIX 1 SIGNIFICANT HAZARDS REVIEW i
4
- O I
i 0
("j 10 CFR 50.92 Evaluation for the Proposed Changes to the Technical Specifica-(_ / tions as a Result of the Inrtallation of the Analog Transmitter Trip System for Edwin I. Hatch Nuclear Plant-Unit 2 Georgia Power Company (GPC) has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed Technical Specifications changes due to the installation of the analog transmitter trip system (ATTS). ATTS replaces the pressure, level, and temperature switches in the reacter protection system and emorgency core cooling system (ECCS) with analog seasor/ trip unit combinations. The system is designed to improve sensor intelligence and reliability, while providing continued monitoring of critical parameters and performing the basic logic function. Since the ATTS instrumentation is superior to the mechanical switches currently used at Plant Hatch and demonstrates less drift, certain surveillance requirements may be extended without significantly increasing tne probability or consequences of an accident previously evaluated. GPC proposes to change the surveillance requirements for the ATTS instrumentation to once per shift for channel checks, once per month for channel functional tests, and once per operating cycle for channel calibrations. These proposed surveillance requirements were previously approved on a generic basis for ATTS equipment by the Nuclear Regulatory Commission (NRC) in NED0-21617-A. Additional changes to the nomenclature used in the Technical Specifications are included for clarification and consistency with this proposed change. GPC has reviewed these proposed changes and considers them not to involve a significant 7s hazards consideration, because they will not significantly increase the g
i probability or consequences of an accident previously evaluated, create the possibility of a new or different accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
The results of this ch'ange are clearly within design and all acceptance criteria; therefore, this change is consistent with Item (vi) of the
" Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983, issue of the Federal Register. The' changes in nomenclature are administrative in nature and consistent with Item (vi) of the above referenced page of the Federtl Register.
l 0
\
i I
l v) i l Al-1
/O 10 CFR 50.92 Evaluation for the Proposed Changes to the Technical ksl i
m -Specifications Reactor Vessel Vater Level - Low Low (Level 2) Trip Setpoint for Edwin I. Hatch Nuclear Plant-Unit 2ca>
Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed reactor vessel water _ level 2 trip setpoint Technical Specifications change. This proposed change lowers the level 2 trip setpoint/ allowable value from -38 in. to -55 in., increasing plant operability by decreasing the number of trips due to normal operational transients. Thus, this change will not involve a significant hazards consideration, because it will not significantly increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
The proposed reactor vessel water level 2 trip setpoint/ allowable value i
revision is well within design and all other acceptance criteria.
Consequently, this change is consistent with Item (vi) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983, issue of the Federal Register and will not result in a significant hazards consideration.
. [h
\/-
r i
O
- a. - See subsection 4B.1 (page 4-31) for discussion of proposed changes.
1 Al-2 t.
t* - - * * - ~ v- T-- ---#'- *-" -e--- --~r-+- e e w-- - - - ww~,- - - - - ----'---e-r- * +--e+'*
- e- -W -v- =, ~ '*
~
/' 10 CFR 50.92 Evaluation for the Proposed Changes to the Technical k_ /N Specifications due to the Deletion of the High Drywell Pressure Signal for Residual Heat Removal (Shutdown Cooling Mode), R,eactor Pressure Vessel (RPV)
Head Spray Valves, and Reactor Water Cleanup Isolation for Edwin I. Hatch Nuclear Plant-Unit 2cas Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they
. relate to the proposed changes to the Technical Specifications due to the deletion of the high drywell pressure signal for residual heat removal (RHR)
(shutdown cooling mode), RPV head spray valves, and RWCU isolation. The purpose of this change is to stop small steam leaks in the drywell from preventing operation of the RHR and RWCU systems during the shutdown cooling mode, thereby prohibiting an acceptable normal shutdown procedure. GPC has reviewed this change and considers it not to involve a significant hazards consideration, because it will not significantly increate the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
Although this change removes a trip signal, it was shown that the use of high drywell pressure as an isolation signal has negligible effect in preventing coolant losses due to an RHR or RWCU pipe break. Therefore, this change does not affect the Appendix K calculation results, and the requirements of 10 CFR 100 are still met. This change is consistent with Item (vi) of the 7-'s " Examples of Amendments that are Considered Not Likely to Involve Significant
( ,) Considerations" listed on page 14870 of the April 6, 1983, issue of the Federal Register and will not result in a significant hazards consideration.
Thus, the results of this change are clearly within all acceptance criteria.
O
\
~
- a. See subsection 4B.2 (page 4-4) for discussion of proposed changes.
Al-3
'y 10 CFR 50.92 Evaluation for the Proposed Changes to the Technical s/ Specifications.due to the Lowered Water Level Trip Setpoint for Isolation of Reactor Weter Cleanup and Secondary Containment, and Starting of Standby Gas '
Treatman't System for Edwin I. Hatch Nuclear Plant-Unit 2ca>
Georgia-Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed changes to the Technical Specifications due to the lowered water level trip setpoint fer isolation of RWCU and secondary containment, and starting of the standby gas treatment system (SGTS). This change proposes to lower the water level trip setpoint for isolation cf RWCU and secondary containment, and startup of the standby gas treatment system (SGTS) from level 3 to level 2. A reactor scram from normal power
(<50-percent rated).usually results in a reactor vessel water level transient due to a void collapse that causes RWCU isolation at level 3. This usually results in the dropping of the cleanup filter cake, added radwaste
. processing, and loss of ability to remove water from the reactor vessel immediately after scram. These problems may be avoided by lowering RWCU isolation to level 2. Lowering the SGTS actuation and secondary containment isolation from level 3 to level 2 reduces the potential for spurious isolations. GPC has reviewed these changes and does not consider them to involve a significant hazards consideration, because they will not significantly increase the probability or consequences of an accident from any accident previously evaluated, create the possibility of a new or different accident from any accident previously evaluated, or involve a N significant reduction in a margin of safety.
f Since the above changes reduce operational problems associated with RWCU and secondary containment isolation and because the ECCS analysis design basis already assumes that the SGTS initiates at level 2, these changes are well within the design and all other acceptance criteria. Consequently, this change is consistent with Item (vi) of the " Example of Amendments that are
~
Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983, issue of the Federal Register and will not result in a significant hazards consideration. .
.I
- a. See subsection 4B.3 (page 4-5) for discussion of proposed changes.
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. - ..- . . . . . . ~ . - - . .
/x
/s.Y 10 CPR 50.92 Evaluation for tha Proposed Changes to the Technical
\~- Specifications due to the Deletion of Ambient Temperature Loops in the Leak Detection System Modification for Edwin I. Hatch Nuclear Plant-Unit 2ca>
Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed change to the Technical Specifications due to the deletion of ambient temperature-loops from the leak detection system. As part of the ATTS modification, the change proposes to use the hot leg of the differential temperature sensor for high ambient trip rather than using an independent trip element and trip device. This arrangement may cause slight changes in the sensitivity of the leak detection system, depending upon the heating, ventilation, and air-conditioning design, but it will not defeat the intended function of the system. The proposed Technical Specifications revisions will reference the trip unit loop from which the ambient temperature trip is taken in place of the existing ambient temperature trip instrument. GPC has reviewed this change and considers it not to involve a significant hazards consideration, because it will not significantly increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
The proposed revision to the leak detection system is well within design and all other acceptance criteria. Since the current single-failura criteria will be maintained by this new logic, this change is consistent with Item
(/)
s- (vi)'of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6,
- 1983, issue of the Federal Register and will not result in a significant hazards consideration.
l l
t
/\
N, -
' ' A. See subsection 4B.4 (page 4-6) for discussion of proposed changes.
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'N
/ 10 CFR 50.92 Evaluation for the Proposed Changes to the Technical k-! Specifications due to the Deletion of Drywell Pressure Sensors E11-N011A, B, C, D for Edwin I. Hatch Nuclear Plant-Unit 2ca>
Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed change to the Technical Specifications due to the deletion of drywell pressure sensors E11-N011A, B, C, D. This change proposes to make the drynell pressure sensor configuration consistent with water levels 2 and 3 sensors. Drywell pressure sensors E11 N010A, B, C, D may be used to provide signals for all four systems of the ECCS and still maintain single-failure criteria. The Technical Specifications revision involves changing the instrument numbers from E11-N010A, 9, C, D to E11-N694A, B, C, D. Thus, this change involves no reduction in safety. GPC has reviewed this proposed change and considers it not to involve a significant hazards consideration, beccuse it will not sf e;nificantly increase
-the probability or consequences of an accident previous 1f evaluated, create the possibility of a new or different accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
Although sensors E11-N011A, B, C, D are being deleted, this change involves no reduction in safety and is within all design and acceptance criteria.
Therefore, GPC considers this to be consistent with Item (vi) of the
" Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983, issue of
(}
\s the Federal Register and will not result in a significant hazards consideration.
f\
O
- a. See subsection 4B.5 (page 4-9) for discussion of proposed changes.
. Al-6
_ [' 10 CFR 50.92 Evaluation for the Proposed Trip Setpoint/ Allowable Value V Modifications to the Technical Specifications for Edwin I. Hatch Nuclear Plant-Unit 2ca>
Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed miscellaneous trip setpoint/ allowable value sodifications to the Technical Specifications. The purpose of this change is to update the Technical Specifications trip setpoints. Since the time that the original setpoints were calculated, a better methodology has been developed. This proposed change uses Regulatory Guide 1.105 methodology in updating the setpoints for the instruments being replaced with the new ATTS units. This change replaces the trip setpoints listed in the Technical Specifications which are the original analytical limits with the newly evaluated allowable values determined through Regulatory Guide 1.105 methodology. GPC has reviewed this change and considers it not to involve a significant hazards consideration, because it will not significantly increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
Since the new setpoints were determined using the criteria of Regulatory Guide 1.105; and the new ATTS instruments are more accurate and reliable, the setpoint changes are clearly within all design and acceptance criteria.
Consequently, GPC considers the setpoint changes to be consistent with both p Item (vi), as discussed above, and Item (vii) of the " Examples of Amendments
(,k that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6,1983, issue of the Federal Register and will not result in a significant hazards consideration.
I r
I i
- a. See subsection 4B.6 (page 4-10) for discussion of proposed changes, f Al-7 t
['j 10 CFR 50.92 Evaluation for the Proposed Changes to the Technical
\- ' -Specifications due to the Reactor Vessel Water Level - High (Level 8) Trip Instrumentation Modification for Edwin I. Hatch Nuclear Plant-Unit 2<a Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed changes to the Technical Specifications due to the reactor vessel water level - high (level 8) trip instrumentation. This change proposes to replace the current instruments with new ATTS units.
Since the ATTS instruments are superior to the mechanical switches currently used at Plant Hatch and demonstrate less drift, the setpoint/ allowable value will be lowered from 58 in. to 56.5 in., using the criteria of Regulatory Guide 1.105. GPC has reviewed this change and considers it not to involve a significant hazards consideration, because it will not significantly increase the probability or consequences of an accident previously evaluated, create the possibility of a new or different kind of accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
Since the new ATTS instrumentation is superior and demonstrates less drift and the new setpoints were calculated using the criteria of Regulatory Guide 1.105, the lowering of the reactor vessel water level 8 trip setpoint/ allowable value is well within the design and all other acceptance criteria. Consequently, this change is consistent with Item (vi) of the
" Examples of Amendments that are Considered Not Likely to Involve Significant
(~' Hazards Considerations" listed on page 14870 of the April 6,1983, issue of
\ -
the Federal Register and will not result in a significant hazards consideration.
a l
l l
l l
1 b
v
- a. See subsection 4B.7 (page 4-20) for discussion of proposed changes.
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['" 10 CFR 50.92 Evaluation for the Proposed Change to the Technical Specifications as a Result of the Elimination of the Reactor Pressure Permissive to the Bypass of the MSIV Closure Si VacuumforEdwinI.MatchNuclearPlant-Unit 2'gnalDuetoLowCondenser Georgia Power Company has reviewed the requirements of 10 CFR 50.92 as they relate to the proposed change to the Technical Specifications as a result of the climination of the reactor pressure permissive to the bypass of the main steam isolation valve (MSIV) closure signal due to low condenser vacuum. The change proposes to delete the reactor steam dome pressure permissive which allows the group 1 isolation valves to be bypassed on a low condenser vacuum isolation signal. .With the permissive deleted, the operator may open the valves from a hot pressurized condition before clearing a s'eram. GPC has reviewed this change and considers it not to involve a significant hazards consideration, because it will not significantly increase the probability or consequences of an accident previously evaluated, create the possibliity of a new or different accident from any accident previously evaluated, or involve a significant reduction in a margin of safety.
Consequently, this change is consistent with Item (vi) of the " Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983, issue of the Federal Register and will not result in a significant hazards consideration.
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By eliminating this permissive, the plant protection features and, therefore, plant safety are not compromised in any manner.
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l a. See subsection 4B.8 (page 4-21) for discussion of proposed changes.
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