LD-85-024, Forwards Proposed Final Version of CESSAR Chapter 16 Sys 80 Tech Specs,Per 850502 Meeting W/Nrc.Completion of Review of Tech Specs Requested by End of June 1985 to Include Revs in Amend 10 to CESSAR: Difference between revisions

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{{#Wiki_filter:____              .    .    .
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C-E Powsr Systtms                  Tel. 203/688-1911                                  _
Combustion Engineering. Inc.        Telex: 99297                                        g 1000 Prospect Hill Road Windsor, Connecticut 06095 POWER                                                                                    i M SYSTEMS                                                                                      j
                                                                                                  ~
Docket No. STN 50-470F                                      May 16,1985 LD-85-024                        -
2 1
3 Hugh L. Thompson, Director                                                                    e Division of Licensing                                                                              -
U.S. Nuclear Regulatory Commission
* Washington, D.C.          20555
 
==Subject:==
CESSAR Technical Specifications                                            k n
 
==Reference:==
C-E Letter LD-85-020, A. E. Scherer to H. L. Thompson (NRC),
                      "CESSAR Technical Specifications", April 26, 1985                              -
Dear Mr. Thompson.
In response to C-E's request in the above Reference to finalize NRC review efforts associated with CESSAR Chapter 16 Technical Specifications, a meeting                    -
was held with the NRC Staff on May 2,1985 to discuss the expected scope of the                    m review. In response to that meeting, this letter forwards a proposed final                    g version of the CESSAR Chapter 16 System 80" Technical Specifications (Enclosure 1).            _
In order to facilitate a timely review and approval by the Staff, and to                      5 minimize the impact of the review on Staff priorities and resources, C-E is                    4 providing, as Enclosure 2, a marked-up copy of the Palo Verde Technical                              -
Specifications (the first System 80 plant to be licensed) to identify differences from the proposed CESSAR Technical Specifications. C-E has evaluated each difference and has determined that most of them should require only minimal review by the NRC Staff. This approach is described in detail in                    _
the Attachment, " Comparison of Palo Verde and System 80 Technical                              -
Specifications". We request that the NRC attempt to complete this review as                          ,
close to the end of June 1985 as possible so that revisions can be included in                    _1 Amendment 10 to CESSAR.                                                                            1 4
If you have any questions on this subject, please feel free to call me or Mrs.                    d R. O. Hoogewerff of my staff at (203) 285-5217.                                                  5 4'
Very truly yours, COMBUSTION ENGINEERING, INC.                      1 Director                          g              l Nuclear Licensing              g                y AES:las                                                                                1              =
Attachment (1)                                                                        g Enclosures (2)                    8505210305    850516 DR                                                -
A    ADOCK 05000470                            -
i PDR
 
ATTACHMENT TO LD-85-024 COMPARISON OF PALO VERDE AND SYTEM 80' TECHNICAL SPECIFICATIONS Summa ry l      C-E conducted a review of both the System 80 and Palo Verde Technical i      Specifications in order to determine the degree of compatibility between the two documents. The results indicated that the Systen 80 Technical Specifications were very nearly identical to Palo Verde's with respect to j      technical content, as was expected; however, a number of editorial or format-related differences were identified. In an effort to expedite NRC review time, i      most of these differences were resolved by revising the Sytem 80 Technical
.      Specifications to be consistent with Palo Verde Technical Specifications. The t      remaining differences are identified in Enclosure (2), a marked-up copy of the Palo Verde Technical Specifications. C-E believes that this approach will i      enable the Staff to base their review on the existing, approved Palo Verde
.      Technical Specifications.
Detailed Review Process First, the System 80 and Palo Verde Technical Specifications were compared on a l    word-by-word basi s. The differences were identified by circling words in the
:    Palo Verde Technical Specifications that did not appear in System 80 Technical Specifications and by adding words to Palo Verde's Technical Specifications that appear in System 80's but not Palo Verde's Technical Specifications. The differences were then classified as either:
(1) Editorial / format differences (denoted by an "ed." in the left-hand margin).
(2) Balance-of-plant differences which are outside System 80's scope (denoted by a "B0P" in the lef t-hand me.rgin).
(
$      (3) Generic differences which imply that the System 80 Technical Specifications I          are consistent with CESSAR FSAR and/or the Standard Technical Specifications (denoted by a  "G" in the left-hand margin).
(4) Plant specific differences (denoted by a "PS" in the left-hand margin).
1
[      Whenever practicable, editorial / format differences between the two documents p      were eliminated by modifying System 80 Technical Specifications to conform to Palo Verde's for ease of review. Actual technical review should only be
[      required for a small subset of the plant specific differences.
The results of this effort include two packages.      Enclosure (1) provides a 1      revision of the System 80 Technical Specifications, where changes that were made to be consistent with Palo Verde are indicated by a "C" in the left-hand i      ma rgi n. Also, for ease of NRC review, a marked-up copy of the Palo Verde Technical Spec 1 fictions that identify all remaining differences between the
?      documents is provided as Enclosure (2).
lir L
me                                            _    _ _ _                            _
 
e LD-85-024 Attachment-Page 2 of 2 In summary, it is C-E's belief that the Staff will find technical review of the
                    . System 80 Technical Specifications'to be minimal based on 'the small number of
                    -technical differences between System 80 and Palo Verde Technical Specifications. ' Such a review can be facilitated by a side-by-side comparison of the proposed System 80 Technical Specifications [ Enclosure (1)] and the Palo Verde Technical Specifications with System 80 differences identified [ Enclosure
                    -(2)].
 
r-ENCLOSURE 1 TO LO-85-024 CESSAR Chapter 16, System 80 Technical Specification Revisions
 
s CHAPTER 16 SYSTEM 80 ,FSAR CESSAR l                                                                                    TECHNICAL SPECIFICATIONS I
C t
f I
Chapter 16 is completely replaced in Anendment No. 9                                                                                      Due to the effort required to revise the format from the original submittal, previous amendments were submitted as interim documents. Therefore, revision lines are not shown on the revised pages.
i Amendment Number 9 i
February 27. 1984
                ..____.,_--,.--.,-.,,-.___n,,,,,__,,_,,~._                , , , , , , . . , , - , _ _ . _ , , _ . , . . . , _ , , . . , , , _ , , . . , . _ - . . . _ , , . _ . . . . . , , , , _ . , , _ _ . , _ , . . . . , , , -    - - , . . -
 
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INDEX DEFINITIONS SECTION                                                                            PAGE 1.1 ACTION . . . . . . . . . . . . . . . . . . . . . . . . . . . .                1-1 1.2 AXIAL SHAPE INDEX . . . . . . . . . . . . . . . . . . . . . .                  1-1 1.3 AZIMUTHAL POWER TILT . . . . . . . . . . . . . . . . . . . . .                1-1 1.4 CHANNEL CALIBRATION . . . . . . . . . . . . . . . . . . . . .                  1-1 1.5 CHANNEL CHECK . . . . . . . . . . . . . . . . . . . . . . . .                  1        1.6 CHANNEL FUNCTIONAL TEST . . . . . . . . . . . . . . . . . . .                  1-1 1.7 CONTAINMENT INTEGRITY . . . . . . . . . . . . . . . . . . . .                  1-2 1.8 CONTROLLED LEAKAGE . . . . . . . . . . . . . . . . . . . . . .                1-2 1.9 CORE ALTERATION . . . . . . . . . . . . . . . . . . . . . . .                  1-2 1.10 DOSE EQUIVALENT I-131 . . . . . . . . . . . . . . . . . . . .                1-2 1.11 E-AVERAGE DISINTEGRATION ENERGY  ...............1-2 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME . . . . . . . . . . .                1-2 1.13 FREQUENCY NOTATION . . . . . . . . . . . . . . . . . . . . . .                1-2 1.14 IDENTIFIED LEAKAGE . . . . . . . . . . . . . . . . . . . . . .                1-2 1.15 OPERABLE - OPERABILITY . . . . . . . . . . . . . . . . . . . .                1-3 1.16 OPERATIONAL MODE - MODE . . . . . . . . . . . . . . . . . . .                1-3 1.17 PHYSICS TESTS . . . . . . . . . . . . . . . . . . . . . . . .                1-3 1.18 PLANAR RADIAL PEAKING FACTOR - Fxy . . . . . . . . . . . . . .                1-3 1.19 PRESSURE BOUNDARY LEAKAGE . . . . . . . . . . . . . . . . . .                1-3 1.20 RATED THERMAL POWER . . . . . . . . . . . . . . . . . . . . . 1-3 1.21 REACTOR TRIP SYSTEM RESPONSE TIME ..............1-3 1.22 REPORTABLE OGGUARENGE. W          E, ,y , , , , , , , , , , , , , , , , , , 1,4 1.23 SHIELD BUILDING INTEGRITY . . . . . . . . . . . . . . . . . .                1-4 1.24 SHUTDOWN MARGIN . . . . . . . . . . . . . . . . . . . . . . .                1-4 1.25 SOFTWARE . . . . . . . . . . . . . . . . . . . . . . . . . . .                1-4 1.26 STAGGERED TEST BASIS . . . . . . . . . . . . . . . . . . . . .                1-4 1.27 THERMAL POWER ........................1-4 1.28 UNIDENTIFIED LEAKAGE , . . . . . . . . . . . . . . . . . . . .                1-4 Amendment Number 9 February 27, 1984 I
 
P INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION                                                                                                                                                    PAGE 2.1 SAFETY LIMITS                                .
2.1.1                      REACTOR CORE . . . . . . . . . . . . . . . . . . . . . .                                                                        2-1 2.1.1.1                    DNBR . . . . . . . . . . . . . . . . . . . . . . . . . .                                                                        2-1 2.1.1.2                    PEAK LINEAR HEAT RATE . . . . . . . . . . . . . . . . . .                                                                      2-1 2.1.2                      REACTOR COOLANT SYSTEM PRESSURE . . . . . . . . . . . . .                                                                      2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1                      REACTOR TRIP SETPOINTS . . . . . . . . . . . . . . . . .                                                                        2-2 2.2.2                      CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS . . . .                                                                        2-2 BASES SECTION                                                                                                                                                    PAGE 2.1 SAFETY LIMITS 2.1.1                      REACTOR CORE . . . . . . . . . . . . . . . . . . . . . . . B 2-1 2.1.2                      REACTOR COOLANT SYSTEM PRESSURE . . . . . . . . . . . . . . B 2-2 I
2.2 LIMITING SAFETY SYSTEM SETTINGS l  2.2.1                      REACTOR TRIP SETPOINTS . . . . . . . . . . . . . . . . . . B 2-3 i
2.2.2                      CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS . . . . . B 2-7 f
l Amendment Number 9 February 27, 1984
 
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                                                    PAGE
            ~3/4.0 APPLICABILITY , . . . . . . . . . . . . . . . . . . . . . .                                          3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 ~BORATION CONTROL SHUTDOWN MARGIN - Tcold > 210*F . . . . . . . . . . . .                                3/4 1-1 SHUTDOWN MARGIN - T cold 32 10'F . . . . . . . . . . . .                  3/4 1-3 MODERATOR TEMPERATURE COEFFICIENT . . . . . . . . . . .                                3/41-4 MINIMUM TEMPERATURE FOR CRITICALITY . . . . . . . . . .                                  3/4 1-6 3/4.1.2      BORATION SYSTEMS FLOW PATHS - SHUTDOWN . . . . . . . . . . . . . . . . .                                3/4 1-7 FLOW PATHS - OPERATING . . . . . . . . . . . . . . . .                                  3/4 1-8 CHARGING PUMPS - SHUTDOWN . . . . . . . . . . . . . . .                                3/4 1-9 CHARGING PUMPS - OPERATING . . . . . . . . . . . . . .                                  3/4 1-10 BORATED WATER SOURCES - SHUTDOWN . . . . . . . . . . .                                  3/4 1-11 BORATED WATER SOURCES - OPERATING . . . . . . . . . . .                                3/4 1-13 BORON DILUTION ALARMS . . . . . . . . . . . . . . . . .                                3/4 1-15 3/4.1.3        MOVA8LE CONTROL ASSEMBLIES CEA POSITION . . . . . . . . . . . . . . . . . . . . .                                  3/4 1-22 POSITION INDICATOR CHANNELS - OPERATING . . . . . . . .                                  3/4 1-25 POSITION INDICATOR CHANNELS - SHUTOOWN . . . . . . . .                                  3/4 1-26 CEA DROP TIME . . . . . . . . . . . . . . . . . . . . .                                  3/4 1-27 SHUTDOWN CEA INSERTION LIMIT . . . . . . . . . . . . .                                  3/4 1-28 REGULATING CEA INSERTION LIMITS . . . . . . . . . . . .                                  3/4 1-29 l
Amendment Number 9        ;
February 27,1984 l
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i
 
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                  PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1    LINEAR HEAT RATE . . . . . . . . . . . . . . . . . . . .      3/4 2-1 3/4.2.2    PLANAR RADIAL PEAKING FACTORS . . . . . . . . . . . . . . 3/4 2-2 3/4.2.3    AZIMUTHAL POWER TILT . . . . . . . . . . . . . . . . . .      3/4 2-3 3/4.2.4    DNBR MARGIN . . . . . . . . . . . . . . . . . . . . . . . 3/4 2-5 3.4.2.5    RCS FLOW RATE . . . . . . . . . . . . . . . . . . . . . . 3/4 2-9 3/4.2.6    REACTOR COOLANT COLD LEG TEMPERATURE . . . . . . . . . .      3/4 2-10 3/4.2.7    AXIAL SHAPE INDEX . . . . . . . . . . . . . . . . . . . . 3/4 2-12 3/4.2.8    PRESSURIZER PRESSURE . . . . . . . . . . . . . . . . . .      3/4 2-13 3/4.3 INSTRUMENTAT M 3/4.3.1    REACTOR PROTECTIVE INSTRUMENTATION . . . . . . . . . . .      3/4 3-1 3/4.3.2
* ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . 3/4 3-16 1/4.3.3    MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION . . . . . . . . .      3/4 3-36 INCORE DETECTORS . . . . . . . . . . . . . . . . . . .      3/4 3-37 SEISMIC INSTRUMENTATION . . . . . . . . . . . . . . . . 3/4 3-38 METEOROLOGICAL INSTRUMENTATION . . . . . . . . . . . .      3/4 3-39 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION . . . . . . . .      3/4 3-40 POST-ACCIDENT MONITORING INSTRUMENTATION . . . . . . .      3/4 3-43 CHLORINE DETECTION SYSTEMS . . . . . . . . . . . . . .      3/4 3-46 FIRE DETECTION INSTRUMENTATION . . . . . . . . . . . .      3/4 3-47 LOOSE-PART DETECTION INSTRUMENTATION . . . . . . . . .      3/4 3-48 3/4.3.4    TURBINE OVERSPEED PROTECTION . . . . . . . . . . . . . .      3/4 3-49 Amendment Number 9 IV                          February 27, 1984
 
INDEX LIMITING CONDITIONS FOR OPEP.ATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                                    PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1    REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION . . . . . . . . . . . . . .                        3/4 4-1 HOT STANDBY . . . . . . . . . . . . . . . . . . . . . .                        3/4 4-2 HOT SHUTDOWN . . . . . . . . . . . . . . . . . . . . .                        3/4 4-3 COLD SHUTDOWN - LOOPS FILLED . . . . . . . . . . . . .                        3/4 4-5 COLD SHUTDOWN - LOOPS NOT FILLED . . . . . . . . . . .                        3/4 4-6 3/4.4.2    SAFETY VALVES SHUTDOWN . . . . . . . . . . . . . . . . . . . . . . .                        3/4 4-7 OPERATING . . .      . g . ,,,,    . . . . . . . . . . . .                  3/4 4-8 g 3/4.4.3      PRESSURIZER g .% m u ed.y.$P/ M ., . . . . . . . . . .                          3/4 4-9 3/4.4.4    STEAM GENERATORS . . . . . . . . . . . . . . . . . . . .                        3/4 4-10 3/4.4.5    REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS . . . . . . . . . . . . . . .                        3/4 4-11 OPERATIONAL LEAKAGE . . . . . . . . . . . . . . . . . .                        3/4 4-12 3/4.4.6    CHEMISTRY , . . . . . . . . . . . . . . . . . . . . . . .                        3/4 4-13 3/4.4.7    SPECIFIC ACTIVITY    . . . . . . . . . . . . . . . . . . .                      3/4 4-16 3/4.4.8    PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM . . . . . . . . . . . . . . . .                        3/4 4-20 h                PRESSURIZERgsp1u7/ qcptpc,Wp ,u,H jff . . . . . . . . .                        3/4 4-23 OVERPRESSURE PROTECTION SYSTEMS . . . . . . . . . . . .                        3/4 4-24 3/4.4.9    STRUCTURAL INTEGRITY . . . . . . . . . . . . . . . . . .                        3/4 4-25
    @  3/4.480 (Ithcied. CDCLA&f $V?.1EH JENG . . . . .. .                                ...
3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1    SAFETY INJECTION TANKS . . . . . . . . . . . . . . . . .                        3/4 5-1 3/4.5.2    ECCS SUBSYSTEMS - HOT STANDilY, STARTUP AND POWER OPERATION. 3/4 5-3 3/4.5.3    ECCS SUBSYSTEMS - DOT SHUTDOWN pD,Hff,5?@ . . . . .                              3/4 5-6 3/4.5.4    REFUELING WATER TANK . . . . . . . . . . . . . . . . . .                        3/4 5-7 Amendment Number 9 y
February 27,1984 i
 
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                                    PAGE C-E ATMOSPHERIC TYPE CONTAINMENT 3/4.6 CONTAINMENT SYSTEMS 3/4.6.!        PRIMARY CONTAINMENT CONTAINMENT INTEGRITY . . . . . . . . . . . . . . . . .                  3/4 6-1 CONTAINMENT LEAKAGE . . . . . . . . . . . . . . . . . .                  3/4 6-1 CONTAINMENT AIR LOCKS . . . . . . . . . . . . . . . . .                  3/4 6-1 CONTAINMENT ISOLATION VALVE AND CHANNEL WELD PRESSURIZATION SYSTEM . . . . . . . . . . . . . . .                  3/4 6-1 l                    INTERNAL PRESSURE . . . . . . . . . . . . . . . . . . .                  3/4 6-1 AIR TEMPERATURE . . . . . . . . . . . . . . . . . . . .                  3/4 6-1 i
CONTAINMENT STRUCTURAL INTEGRITY . . . . . . . . . . .                  3/4 6-1 CONTAINMENT VENTILATION SYSTEM . . . . . . . . . . . .                  3/4 6-1 3/4.6.2        DEPRESSURIZATION AND COOLING SYSTEMS l
CONTAINMENT SPRAY SYSTEM . . . . . . . . . . . . . . .                  3/4 6-2 IODINE REMOVAL SYSTEM . . . . . . . . . . . . . . . . .                  3/4 6-3 CONTAINMENT COOLING SYSTEM . . . . . . . . . . . . . .                  3/4 6-5 3/4.6.3        CONTAINMENT ISOLATION VALVES . . . . . . . . . . . . . .                  3/4 6-6 3/4.6.4        COMBUSTIBLE GAS CONTROL HYDROGEN ANALYZERS . . . . . . . . . . . . . . . . . .                  3/4 6-10 ELECTRIC HYDROGEN RECOMBINERS . . . . . . . . . . . . .                  3/4 6-10 HYOR0 GEN MIXING SYSTEMS . . . . . . . . . . . . . . . .                3/4 6-10 PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM . . . . . .                  3/4 6-10 VACUUM RELIEF VALVES . . . . . . . . . . . . . . . . .                  3/4 6-10 3/4.6.5        IODINE CLEANUP SYSTEM . . . . . . . . . . . . . . . . . .                  3/4 6-11 Amendment Number 9 February 27,1984 VI l
 
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                    PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1    TURBINE CYCLE SAFETY VALVES . . . . . . . . . . . . . . . . . . . . .      3/4 7-1 EMERGENCY FEEDWATER SYSTEM . . . . . . . . . . . . . .      3/4 7-4 CONDENSATE STORAGE TANK . . . .                              3f47,5
: g. g g.g.      ,,
ACTIVITY  ...........              . . . . . . . . . . . 3/4 7-6 b            MAIN STEAM LINE ISOLATION val p    . . . . . . . . . . . 3/4 7-8 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION . . . . .        3/4 7-9 3/4.7.3 COMPONENT COOLING WATER SYSTEM . . . . . . . . . . . . .          3/4 7-10 3/4.7.4 SERVICE WATER SYSTEM . . . . . . . . . . . . . . . . . .          3/4 7-10 3/4.7.5 EMEAGGWHHMMM-#GNG U.@.A@. h@,5fN.k . . . . . . .                  3/4 7-10 3/4.7.6 FLOOD PROTECTION . . g . . . . . . . . . . . . . . . .            3/4 7-10 3/4.7.7 CONTROL ROOM EMERGENCYaCLEANUP SYSTEM . . . . . . . . .            3/4 7-10 l 3/4.7.8 ECCS PUMP ROOM EXHAUST AIR CLEANUP SYSTEM . . . . . . . .          3/4 7-10 l 3/4.7.9 SNUBBERS . . . . . . . . . . . . . . . . . . . . . . . .          3/4 7-11 3/4.7.10 SEALED SOURCE CONTAMIN ATION , . . . . . . . . . . . . . .        3/4 7-14 3/4.7.11 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM . . . . . . . . . . . . .      3/4 7-15 SPRAY AND/0R SPRIhKLER SYSTEM . . . . . . . . . . . . .      3/4 7-15 CO SYSTEMS . . . . . . . . . . . . . . . . . . . . . .      3/4 7-15 2
HALON SYSTEM . . . . . . . . . . . . . . . . . . . . .      3/4 7-15 FIRE HOSE STATIONS . . . . . . . . . . . . . . . . . .      3/4 7-15 l              YARD FIRE HYDRANTS AND HYDRANT HOSE HOUSES . . . . . .      3/4 7-15 3/4.7.12 FIRE BARRIER PENETRATIONS . . . . . . . . . . . . . . . .        3/4 7-16 3/4.7.13 AREA TEMPERATURE MONITORING . . . . . . . . . . . . . . .          3/4 7-17 l
3/4.7.14 SHUTDOWN COOLING SYSTEM . . . . . . . . . . . . . . . . .        3/4 7-18 i
Amendment Number 9 February 27,1984  .
VII                                            l l
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                                                                                                                      PAGE
                                        @      3/4.8 ELECTRICAL                                                                                SYSTEMS 3/4.8.1                  A.C. SOURCES OPERATING . . . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 8-1 SHUTDOWN . . . . . . . . . . . . . . . . . . . . . . .
3/4 8-1 3/4.8.2                  D.C. SOURCES OPERATING . . . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 8-2 SHUTDOWN . . . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 8-2 3/4.8.3                  ONSITE POWER DISTRIBUTION SYSTEMS OPERATING . . . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 8-3 SHUTDOWN . . . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 8-3 3/4.8.4                  ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES . . . . . . . . . . . . . . . . .                                                                                        3/4 8-4 l                                                                          MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES . . . . . . . . . . . . . . . . . . .                                                                                        3/4 8-4 3/4.9 REFUELING OPERATIONS 3/4.9.1                  BORON CONCENTRATION . . . . . . . . . . . . . . . . . . .                                                                                        3/4 9-1 3/4.9.2                  INSTRUMENTATION . . . . . . . . .' . . . . . . . . . . . .                                                                                        3/449-2 j                                              3/4.9.3                  DECAY TIME . . . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 9-3 3/4.9.4                  CONTAINMENT BUILDING PENETRATIONS . . . . . . . . . . . .                                                                                        3/4 9-4 l                                              3/4.9.5                  COMMUNICATIONS . . . . . . . . . . . . . . . . . . . . .                                                                                          3/4 9 4 3/4.9.6                  REFUELING MACHINE                                                                                      . . . . . . . . . . . . . . . . . . . 3/4 9-4 3/4.9.7                  CRANE TRAVEL - SPENT FUEL STORAGE P0OL BUILDING . . . . .                                                                                        3/4 9-4 3/4.9.8                  SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL . . . . . . . . . . . . . . . . . . .                                                                                          3/4 9-5 l                                                                          LOW WATER LEVEL . . . . . . . . . . . . . . . . . . . .                                                                                        3/4 9-6 3/4.9.9                  CONTAINMENT PURGE VALVE ISOLATION SYSTEM . . . . . . . .                                                                                          3/4 9-7 3/4.9.10 EATERLEVEL-REACTORVESSEL . . . . . . . . . . . . . .                                                                                                              3/4 9-8 F3/4.9.11 UATERLEVEL-STORAGEPOOL . . . . . . . . . . . . . . .                                                                                                                  3/4 9-9 3/4.9.12 STORAGE P00L AIR CLEANUP [ M E.H. . . . . . . . . . . . .                                                                                                          3/4 9 10 i
                                      '  g(ueL
                                          . , .6 MA%6HELtES                                                                                                                                                          Amendment Number 9 l                                                                                                                                                                                        y!!I                        February 27,1984 l
 
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                  PAGE 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTOOWN MARGIN . . . . . . . . . . . . . . . . . . . . .      3/4 10-1 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS . . . . . . . . 3/4 10-2 3/4.10.3 REACTOR COOLANT LOOPS . . . . . . . g .        .g g g.j Qa g? -
    @ 3/4.10.4 CEA POSITION h REGULATING CEA                        j INSERTION  3/4 10-4LIMIT 5 J g 3/4.10.5 SAFETY-fWf6ftM-TANMS M*le"!H. ($PtPWo?( % gf . 3/4 10-6 3/4.10.6 SAFETY INJECTION TANK [MEGGWRE , . . . . . . . . . . . .        3/4 10-7 3/4.10 ~}. SAF6T/ IMRCfl0N TAWK Pace.oR6 . . . . . .        . . .      3/4 (0- 8 1
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1 Amendment Number 9 gx l                                                                        February 27. 1984 l
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INDEX BASES SECTION                                                                    PAGE 3/4.0 APPLICABILITY . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1    BORATION CONTROL . . . . . . . . . . . . . . . . . . . . . B 3/4 1-1 3/4.1.2    BORATION SYSTEMS . . . . . . . . . . . . . . . . . . . .      8 3/4 1-2      ,
3/4.1.3    MOVABLE CONTROL ASSEMBLIES .. . . . . . . . . . . . . . . . B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1    LINEAR HEAT RATE . . . . . . . . . . . . . . . . . . . . . B 3/4 2-1 l 6 3/4.2.2    N PEAKING FACTORS . . . . . . . . . . . . . . . . . . B 3/4 2-2
;  3/4.2.3    AZIMUTHAL POWER TILT . . . . . . . . . . . . . . . . . .      8 3/4 2-2 3/4.2.4    DN8R MARGIN . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-3 l  3/4.2.5      RCS FLOW RATE . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-3 3/4.2.6      REACTOR COOLANT COLD LEG TEMPERATURE . . . . . . . . . .      8 3/4.2-4    .
3/4.2.7    AXIAL SHAPE INDEX . . . . . . . . . . . . . . . . . . . .      8 3/4 2-4
!  3/4.2.8      PRESSURIZER PRESSURE . . . . . . . . . . . . . . . . . . . B 3/4 2-4 l
l  3/4.3 INSTRUMENTATION l  3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY l
FEATURES ACTUATION SYSTEM INSTRUMENTATION . . . . . . . 8 3/4 3-1 3/4.3.3    MONITORING INSTRUMENTATION . . . . . . . . . . . . . . . . B 3/4 3-1 3/4.3.4      TUR81NE OVERSPEED PROTECTION . . . . . . . . . . . . . .      8 3/4 3-2 3/4.4 REACTOR COOLANT SYSTEM l
l  3/4.4.1      REACTOR COOLANT LOOPS AND COOLANT CIRCULATION . . . . . . 8 3/4 4-1
\
l  3/4.4.2      SAFETY VALVES . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-2
(  3/4.4.3      PRESSURIZER . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-2 l  3/4.4.4      STEAM GENERATORS . . . . . . . . . . . . . . . . . . . . . B 3/4 4 2 l
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Amndment Number 9    '
X February 27,1984 l
 
INDEX BASES SECTION                                                                  PAGE 3/4.4.5    REACTOR COOLANT SYSTEM LEAKAGE . . . . . . . . . . . . . . B 3/4 4-2          <
3/4.4.6    CHEMISTRY , . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.7    SPECIFIC ACTIVITY . . . . . . . . . . . . . . . . . . . . . B 3/4 4-3 3/4.4.8    PRESSURE / TEMPERATURE LIMITS . . . . . . . . . . . . . . . . B 3/4 4-4 3/4.4.9    STRUCTURAL INTEGRITY . . . . . . . . . . . . . . . . . . . B 3/4 4-6
      @ 3lq.q.go    (2GncdoL CDCLAnf $1SfGH VGNU 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1    SAFETY INJECTION TANKS . . . . . . . . . . . . . . . . . 8 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUS $YSTEMS . . . . . . . . . . . . . . . . . B 3/4 5-2 3/4.5.4 REFUELING WATER TANK . . . . . . . . . . . . . . . . . . . I 3/4 5-3 C-E - ATMOSPHERIC TYPE CONTAINMENT l
3/4.6 CONTAINMENT SYSTEMS 3/4.6.1    PRIMARY CONTAINMENT . . . . . . . . . . . . . . . . . . .      8 3/4 6 1 3/4.6.2    DEPRESSURIZATION AND COOLING SYSTEMS . . . . . . . . . . . B 3/4 6-1 3/4.6.3    CONTAINMENT ISOLATION VALVES . . . . . . . . . . . . . . 8 3/4 6-1 3/4.6.4    COMBUSTIBLE GAS CONTROL . . . . . . . . . . . . . . . . . . B 3/4 6-1 1
3/4.6.5    IODINE CLEANUP SYSTEMS . . . . . . . . . . . . . . . . . 8 3/4 6-1            ,
3/4.7 PLANT 5) STEMS 3/4.7.1    TUP 81NE CYCLE . . . . . . . . . . . . . . . . . . . . . . 8 3/4 7-1 3/4.7.2    STEAM GENERATOR PRES $URE/ TEMPERATURE LIMITATICN . . . . . 8 3/4 7 2 3/4.7.3    COMP 0NENT COOLING WATER SYSTEM . . . . . . . . . . . . . . B 3/4 7 2 3/4.7.4    SERVICE WATER SYSTEM . . . . . . . . . . . . . . . . . . 8 3/4 7-2 3/4.7.5    EMHetWet-t00t-tMt-90N0*.W.".iAT9 8.N.9 .*.9 . . . . . . . I 3/4 7 2 3/4.7.6    FLOOO PROTECTION . . . . . . . . . . . . . . . . . . . . 8 3/4 7 2 3/4.7.7    CONTROL ROOM ESSENTIAL FILTRATION SYSTEM . . . . . . . .      8 3/4 7-3      ,
3/4.7.8    ESF PUMP ROOM AIR EXHAUST CLEANUP $YSTEM . . . . . . . . . B 3/4 7-3 3/4.7.9    $NUBBERS . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7 3          I Amendment Number 9  .
XI                          February 27,1964 r
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INDEX BASES SECTION                                                                  PAGE 3/4.7.10 SEALED SOURCE CONTAMINATION . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.11 FIRE SUPPRESSION SYSTEMS . . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.12 FIRE RATED ASSEMBLIES . . . . . . . . . . . . . . . . . . . B 3/4 7-3 3/4.7.13 AREA TEMPERATURE MONITORING . . . . . . . . . . . . . . . . B 3/4 7-3
  @ 3/4.7.14 SHUTDOWN COOLING yY.5ff M . . . . . . . . . . . . . . . . . B 3/4 7-3          '
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2. and 3/4.8.3 A.C. SOURCES, D.C. SOURCES AND                            l ONSITE POWER 0!$TRIBUT!0N SYSTEMS . . . . . . . . . . 8 3/4 8-1 3/4.8.4    ELECTRICAL EQUIPMENT PROTECTIVE DEVICES . . . . . . . . . 8 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.2    INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.3 DECAY TIME . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 9-1 3/4.9.4 CONTAINMENT PENETRATIONS . . . . . . . . . . . . . . . . 8 3/4 9-1 3/4.9.5 CONRJNICATIONS . . . . . . . . . . . . . . . . . . . . . 8 3/4 9-1 3/4.9.6 REFUELING MACHINE . . . . . . . . . . . . . . . . . . . . 8 3/4 9-1 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILO!NG . . . . . .'8 3/4 9-1 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION . . . . . . . . . B 3/4 9-1 3/4.9.9 CONTAINMENT PURGE VALVE !$0LAT!0N SYSTEM . . . . . . . . . B 3/4 9-2 3/4.9.10 and 3/4.9.11        WATER LEVEL - REACTOR VESSEL AND STORAGE POOL . . . . . . . . . . . . . . . . . . . . . . B 3/4 9 2 3/4.9.12 STORAGE POOL AIR CLEANUP SYSTEM . . . . . . . . . . . . .      8 3/4 9 2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUT 00WN MARGIN . . . . . . . . . . . . . . . . . . . . . . B 3/4 10 1      1 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION AND POWER O!$TRIBUT!0N LIMITS . . . . . . . . . 8 3/4 10 1 l
3/4.10.3 REACTOR COOLANT LOOPS . . . . . . . . . . . . . . . . . . . 8 3/4 10 1 Amendment Number 9
)
Fouruary 27,1984 XII
 
INDEX l      BASES SECTION                                        y    y    eo      , PAGE REGULATING CEA INSERTION LIMITSg . . . . . 8 3/4 10-1 (Q  4.10.4 CEA POSITION l      3/4.10.5 SAFETY INJECTION TANKS . . . . . . . . . . . . . . . . . . B 3/4 10-1 3/4.10.VSAFETYINJECTIONTANKPRESSURE . . . . . . . . . . . . . . B 3/4 10-2 Hsulnurs StH9tttMut& AND SttssuRe
    @ N4,ga.S ('ed. C A lf R A L i r y i
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I Amendment Number 9 February 31,1984 XI!!                                      '
I l
 
INDEX DESIGN FEATURES SECTION                                                                PAGE 5.1 SITE . . . . . . . . . . . . . . . . . . . . . . . . . . . . .      5-1 5.2 CONTAINMENT . . . . . . . . . . . . . . . . . . . . . . . . .      5-1 5.3 REACTOR CORE 5.3.1      FUEL AS$EMBLIES . . . . . . . . . . . . . . . . . . . . . 5-1 i
5.3.2      CONTROL ELEMENT ASSEMBLIES . . . . . . . . . . . . . . . 5-1 l
5.4 REACTOR COOLANT SYSTEM 5.4.1      DESIGN PRES $URE AND TEMPERATURE . . . . . . . . . . . . . 5-1 5.4.2      VOLUME . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.5 METEOROLOGICAL TOWER LOCATION . . . . . . . . . . . . . . . .      5-1 l
l  5.6 FUEL STORAGE l  5.6.1 ,    CRITICAL ITY , . . . . . . . . . . . . . . . . . . . . . . 5-1 5.6.3      DRAINAGE . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.6.4      CAPACITY . . . . . . . . . . . . . . . . . . . . . . . . 5-1
                                                          ~~
5.7 COMPMENT CYCLIC OR TRAN51ENT LIMITS . . . . . . . . . . . . .      5-2 r
I l
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Amendment Number 9 !
February 27,1964  !
XIV                                          l f
 
c                                                                                          -
i INDEX ADMINISTRATIVE CONTROLS lifrilEE                                                              fAE 6.1 RESP 0N5!BILITY , . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2 ORGANIZATION . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 l
6.3 UNIT STAFF QUALIFICATION . . . . . . . . . . . . . . . . . . . 6-1 6.4 TRAINING . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.5 RivitW AND AUDIT . . . . . . . . . . . . . . . . . . . . . . . 6-1 GVGNf
              @          6.6 REPORTABLE OGGWMNGE ACTION    .................6-1 6.7 SAFETY LIMIT VIOLATION . . . . . . . . . . . . . . . . . . . . 6-1 6.8 PROCT 00RES AND PROGRMS . . . . . . . . . . . . . . . . . . . 6-1 6.9 REPORTING REQUIREMENTS . . . . . . . . . . . . . . . . . . . . 6-1 6.10 RECORD RETENTION . . . . . . . . . . . . . . . . . . . . . . . 61 6.11 M01AT10N PROTECTION PR06MM . . . . . . . . . . . . . . . . . 61 6.12 HIGH MDIATION AttA .....................                        61 XV                  Amendment Number 9 February 27,1984 s                                                                              .
 
l l
!                                              .l.!El LIST OF FIGURES l                                                                                PAGE 3.1-1      ALLOWASLE MTC MODES 1 AND 2 . . . . . . . . . . . . . . .      3/4 1-5 3.1-2      MINIMUM 80 RATED WATER VOLUMES . . . . . . . . . . . . . . 3/4 1-12 l      3.1-3      CEA INSERTION LIMITS V5. THERMAL POWER ........            . 3/4 1-31 (Q 3.2-1      DN8R MARGIN OPERATING LIMIT BASED ON COL 55    i'.Y        . 3/4 2-7 l
3.2-2      DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR $(COL 550UTOFSERVICE)    . . . . . . . . . . 3/4 2-8 l      3.2-3      REACTOR C00LANT COLD LEG TEMPERATURE VS. CORE POWER LEVEL . 3/4 2-11 3.4-1      D0SE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT 0F RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY > 1.0 uCi/ GRAM 005E l                    EQUIVALENT 1-131 . . . . . . . . . . . . . . . . . . .      3/4 4-19 3.4-2      REACTOR C00LANT SYSTEM PRES $URE TEMPERATURE LIMITATIONS . 3/4 4-21 XVI                  Amen eent Number 9 February 27, IMI
 
INDEX LIST OF TA8LES PAGE 1.1        FREQUENCY NOTATION . . . . . . . . . . . . . . . . . . .        1-5 1.2      OPERATIONAL MODES . . . . . . . . . . . . . . . . . . . .        1-6 2.2-1      REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS .      2-3 2.2-2    CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS . . . .        2-7 3.1-1    PONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR K,ff > 0.98 FOR MODES 3, 4. AND 5 . . . . . . . . . . .      3/4 1-17 3.1-2    MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.98 g K,ff > 0.97 FOR MODES 3, 4. AND 5 . . . . . . .      3/4 1-18 3.1-3    MONITORING FREQUENCIES FOR BACKUP B0RON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.97 g K,ff > 0.96 FOR MODES 3, 4. AND 5 . . . . . . .        3/4 1-19 3.1-4    MONITORING FREQUENCIES FOR BACKUP B0RON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.96 g K,ff > 0.95 FOR MODES 3, 4. AND 5 . . . . . . .        3/4 1 20 3.1-9    MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.95 g K,ff FOR MODES 3, 4. AND 5 . . . . . . . . . . .      3/4 1-21 3.3-1      REACTOR PROTECTIVE INSTRUMENTATION . . . . . . . . . . .        3/4 3 3 3.3-2      REACTOR PROTECTIVE INSTRUMENTATION Rr.5PONSE TIMES . . . .      3/4 3 9 4.3 1      REACTOR PROTECTIVE INSTRUMENTATION $URVE!LLANCE REQUIREPENTS . . . . . . . . . . . . . . . . . . . . .        3/4 3 12 3.3 3      ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION . . . . . . . . . . . . . . . . . . . .      3/4 3 17 3.3 4      ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES . . . . . . . . . . . . . .      3/4 J.24 3.3 5      ENGINEERED SAFETY FEATURES RESPONSE TIMES . . . . . . . .      3/4 3 27 4.3 2      ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . .      3/4 3 30 3.3 6      REMOTE SHUTDOWN SYSTEM INSTRUMENTATION . . . . . . . . .        3/4 3 41 4.3 3      REMOTE SHUTDOWN SYSTEM INSTRUMENTAfl0N SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . .        3/4 3 42 3.3 7      P0$f ACCIDENT MONITORING INSTRUMENTATION . . . . . . . .        3/4 3 44 XVII    Amendment Number 9 February 27,1984 i
 
INDEX LIST OF TA8LES PAGE 4.3-4    POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . .                            3/4 3-45 3.4-1    REACTOR COOLANT SYSTEM CHEMISTRY , . . . . . . . . . . .                            3/4 4 14 4.4-1    REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVi!LLANCE REQUIREFENTS . . . . . . . . . . . . . . . . . . . . .                            3/4 4-15 4.4-2    PRIMARY COMANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS Pr0GRM , . . . . . . . . . . . . . . . . . .                            3/4 4-18 4.4-3    REACTOR VESSEL MATERIAL SURVE!LLANCE PROGRAM WITH0RAWAL SCHEDULE . . . . . . . . . . . . . . . . . .                          3/4 4-22 3.6 1    CONTAINMENT ISOLATION VALVE ACTUATION TIMES . . . . . . .                            3/4 6-7 3.7-1      STEM LINE SAFET) VgVgggg gy . . . . .                                              3/4 7-2
  @  3.7-2    MAXIMUM ALLOWABLtAVARIABLE OVERPOWER MW TRIP SETP0 INT WITH INOPERABLE STEM LINE SAFETY VALVES DURING TWO LOOP OPERATION WITH FOUR PUMPS OPERATING . . . . . . .                            3/4 7-3  ,
4.7-1      SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRM . . . . . . . . . . . . . . . . .                            3/4 7-7 g  3r7-3e -  SMETY-RELATED-HYDRAULIC-SNU66ERS . . . . . . . . .~ r . .                          3/4-7=W 3r7-h      SA FE T Y-Aft Af f D-MECHAMtC At-- S E00"RS . . .-. . . . . . . .                  3/4 7-13 5.7-1    COMPONENT CYCLE OR TRANSIENT LIMITS . . . . . . . . . . .                            53 5.7 2    COMP 0MMf-ttttt-0R-fRAM$ttMf-t-fttHS                                                55
  @              twv.nettca, Wn't wate trM WiM XVI!!                                        Amendment Number 9 February 21, 1984
 
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D6FINIftoNS r
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1.0        DEFINITIONS The defined tems of this section appear in capitalized type and are applic-able throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.
AXIAL SHAPE INDEX 1.2 The AXIAL SHAPE INDEX shall be the power generated in the lower half of the core less the power generated in the upper half of the core divided by the sum of these powers.
AZIMUTHAL POWER T!LT - T, 1.3 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symetric fuel assemblies.
CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CAL 18-RATION shall encompass the entire channel including the sensor and alarm and/
or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This detemination shall include, where possible, comparison of the channel indication anJ/or status with other indi-cations and/or status derived from indeperdent instrument channels measuring the same parameter.
CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall bei
: a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERA 81LITV including alam and/or trip functions,
: b. Sistablechannels-theinjectionofasimulatedsignalintothe sensor to verify OPERABILITY including alam and/or trip functions.
: c. Olgital computer channels      the exercising of the digital computer hardwareusingdiagnosticprogramsandtheinjectionofsimulated process data into the channel to verify OPERABILITY including alam l
and/or trip functions.
Amendment Number 9
  @ -a,    ,f,,g g 11                        February 21,1984
 
                                                      ~                                        ,
(                - - - - - -
The CHANNEL FUNCTIONAL TEST shall irclude adjustment, as necessary, of
{
I the alare, interlock and/or trip setpoints such that the setpoints are            '
l I        Q in the required range and accuracy.
                                                                    .  . . . . _ _ _ . _ _  ]
 
CONTAINMENT INTEGRITY 1.7 See Applicant's SAR.
CONTROLLED LEAKAGE                                        -
1.8 Not Applicable.
CORE ALTERATION 1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspensica of CORE ALTERATION shall not preclude completion of movement of a compenent to a safe conservative position.
DOSE EQUIVALENT l-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/
gram) which alone would produce the same thyroid dose as the quantity and sotopic mixture of I-131,1-132, I-133,1-134 and 1-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table !!! of T!D-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
E,- AVERAGE O!SINTEGRATION ENERGY 1.11Eshallbetheaverage(weightedinproportiontotheconcentrationof each radionuclide in the reactor coolant at the tirne of sam of the average beta and game energies per disintegration        (pling))of in MEV        the sum for iso-topes, other than iodines, with half Itves greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
ENGINEERED SAFETY FEATURE RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of perfoming its safety function (i.e.,thevalvestraveltotheirrequiredpositions,pumpdischarge pressuresreachtheirrequiredvalues,etc.). Times shall include diesel generator starting and sequence loading delays where applicable.
FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
IDENTIFIEDt.EAKAGE 1.14 10ENTIFIED LEAKAGE shall bet
: a. Leahane into closed systems other than reactor coolant pump control-led b'eedoff flow, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or Amendment Number 9 g                          February 27,1984 L                                                                                        ,
 
                          - - - - - . - - _ u m __
j                                                                                            ll
: b. Leakage into the containment atmosphere froa sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE B0UND-ARY LEAKAGE, or
: c. Reactor coolant system leakage through a steam generator to the secondary system.
OPERABLE - OPERABILITY
(
1.15 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of perfoming its specified function (s),
and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE - MODE 1.16 An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and cold leg reactor coolant temperature specified in Table 1.2.
PHYSICS TESTS 1.17 PHYSICS TESTS shall be those tests perfomed to measure the fundamental nuclear characteristics of.the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) othemise approved by the Comission.
PLANAR RADIAL PEAKING FACTOR - Fxy 1.18 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane aver-age power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.
PRESSURE BOUNDARY LEAKAGE 1.19 PRESSURE B0UNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.
RATED THERMAL POWER 1.20 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3800 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.21 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from
(  when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.
l                                                                      Amendment Number 9 February 27, 1984 1-3 f
9
                                                          >I _
 
Et&f REPORTABLE OGGUARENGE-0    1'22    '-^ ' '^^ ^ ^ **"
A REPCRfABLG GUEtW Shu bt GZd7Vid&'"5(Mfik2 Gs Sec.tien. So 33 tv Iocf(2 ('
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S'o(.
SHIELD BUILDING INTEGRITY 1.23 See Applicant's SAR.
SHUTDOWN MARGIN                                                                              I 1.24 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:                                                                '
: a. No change in part length control element assembly position, and
: b. All full length control element assemblies (shutdown and regulating)
[                    are fully inserted except for the single assembly of highest reactiv-ity worth which is assumed to be fully withdrawn.
SOFTWARE 1.25 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation and proce-dures.
STAGGERED TEST BASIS 1.26 A STAGGERED TEST BASIS shall consist of:
: a. A test schedule for n systems, subsystems, trains or other designat-l                    ed components obtained by dividing the specified test interval into n equal subintervals, and
: b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.
THERMAL POWER
: l.        1.27 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
UNIDENTIFIED LEAKAGE 1.28 UNIDENTIFIED LEAKAGE shall be all leakage which does not constitute either IDENTIFIED LEAKAGE or reactor coolant pump controlled bleedoff flow, i
Amendment Number 9 ary 27, N 1-4                                                f r                                                                                                      l n  .
I
 
4 TABLE 1.1 FREQUENCY NOTATION i
Notation                                                            Frequency S                                            At least once per 12 hours.
D                                            At least once per 24 hours.
W                                            At least once per 7 days.
M                                            At least once per 31 days.
Q                                            At least once per 92 days.
SA                                            At least once per 184 days.
R                                            At least once per 18 months.
S/U                                          Prior to each reactor startup'.
N.A.                                          Not applicable.
i Amendment Number 9 1-5                                February 27, 1984
 
1 TABLE 1.2 OPERATIONAL MODES REACTIVITY            % RATED            COLD LEG CONDITION, K                          TEMPERATURE, T MODE                          eff    THERMAL POWER                    cold
: 1. POWER OPERATION          > 0.99              > 5%                > 500'F
: 2. STARTUP                  > 0.99              < 5%                > 500*F
: 3. HOT STANDBY              < 0.99                  0                > 350*F
: 4. HOT SHUTDOWN              < 0.99                  0 350*F > Tcold > 210*F
: 5. COLD SHUTDOWN.            < 0.99                  0                < 210'F
: 6. REFUELING ***            < 0.95                  0                < 135'F
* Excluding decay heat
**  See further temperature restrictions in LCO #3.1.1.4.
*** Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.
Amendment Number 9 February 27, 1984 1-6
                                                            ,y v  -    ++-ey w      e  rm--
 
TABLE 2.2-1 (Continued)
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATION t
(1)  Trip may be manually bypassed above 10-4% f RATED THERMAL POWER: bypass shall be automatically removed when THERMAL POWER is less than or equal to 10~g% of RATED THERMAL POWER.                                        '
(2) In MODES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(3) In MODES 3-6, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint m          is reached.
Oi (4) % of the distance between steam generator upper and lower level wide range instrument nozzles.
(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpcint includes              !
measurement, calculational and processor uncertainties, and dynamic allowances. Trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is r than or e M heatf6)is te    not usekual to 1% of RATED THERMAL POWER.
(7) r#LOOR_fLthe_ minimum _value of the trip _ setpcHt)
W is the maximum rate of decrease of the trip setpoint. There are no restrictions on the rate at which    ;
1he setpoint can increase.
      @ j"TAND is the amount by which the trip setpoint is below the input signal unless limited by the rate or the    *
,        h h art % d KO73 M M OM-(8) The setpoint may be altered to disable trip function during testing pursuant to Specification 3.10.3.        ;
e o.                                                                                                                  :
* ?+                                                                                                                    i t%                                                                                                                    i a
n,.
 
Y                                        -
(9) Percent of RATED THERMAL POWER.jWith one or more inoperable main steap-line-safety-valves,_refekto blEL.mihm_3.1.1.1 ror_Ine_maxisun_ variable _ overpower _tdp seteo_13f.
(19 filONG is the maximum value of the trip setpoint).
nmt is the maximum rate of increase of the trip setpoint. There are no restrictions on the rate at which h      4 he setpoint can decrease.
8AND is the amount by which the trip setpoint is above the input signal unless limited by the rate or the ceITing.
(lli % of the distance,between steam generator upper and lower level narrow range instrument nozzles.
    '?
cn n
4a 0l=
 
l TABLE 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS
: 1. TYPE I ADDRESSABLE CONSTANTS Point ID      Program                                                    Allowable Number        Label                  Description                          Value 60            FC1    Core coolant mass flow rate calibra-  See Applicant's SAR tion constant 61          FC2    Core coolant mass flow rate calibra-  See Applicant's SAR tion constant 62        CEANOP    CEAC/RSPT inoperable flag              See Applicant's SAR 63          TR      Azimuthal tilt allowance              See Applicant's SAR 64          TPC    Thermal power calibration constant    See Applicant's SAR 65        KCAL    Neutron flux power calibration          See Applicant's SAR constant 66        DNBRPT    DNBR pretrip setpoint                  See Applicant's SAR 67        LPDPT    Local power density pretrip setpoint    See Applicant's SAR Amendment Number 9 2-7                              February 27, 1984  l I
 
E~
THIS PAGE INTENTIONALLY BLANK l
 
w i
AL SPECIFI BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
                                                                                                                                                          .      4 l
Amendment Number 9 i                                                                                                                                    February 27, 1984 l
i t
 
l i
e-THIS PAGE INTENTIONALLY BLANK
 
NOTE The BASES contained in the succeeding pages sununarizes the reasons for the specifications of Section 2.0, but, in accordance with 10 CFR 50.36, are not a part of the Technical Specifications.
4 i
1 i
i t
Amendment Number 9 B 2-N                                    February 27, 1984 I
 
O THIS PAGE INTENTIONALLY BLANK L
 
2.1        SAFETY LIMITS BASES I
2.1.1        REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel clad-ding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature, and (2) maintaining the dynamically adjusted peak linear heat rate of the fuel at or less than 21.0 kw/ft to prevent fuel centerline melting in any fuel rod.
First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly greater than the coolant saturation temperature.
The  upperboiling" nucleate    boundary  (of the DNB).      nucleate At this point,boiling regime there is      is termed a sharp        " departure reduction of the from heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location of DN8 for axially unifom and non-uniform heat flux distributions. The local DN8 ratio (DN8R), defined as the ratio of the predicted DN8 heat flux at a particular core location to the actual heat flux at that location, is indicative of the margin to DN8. The minimum value of DN% during nomal operation and design basis anticipated operational occurrence, is limited to 1.231 based upon a statistical combina-tion of CE-1 CHF correlation and engineering factor uncertainties and is established as a Safety Limit.
Second, operation with a peak linear heat rate below that which would cause fuel centerline melting maintains fuel rod and cladding integrity. Above this peak linear heat rate level (i.e., with some melting in the center), fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods. Volume changes which accom-pany the solid to liquid phase change are significant and require accomoda-tion. Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting. Because cf the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit. To account for fuel rod dynamics (lags), the directly indicated linear heat rate is dynamically adjusted by the CPC program.
Limiting safety system settings for the Low DN8R, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and limiting conditions for operation on DNBR and kw/ft margin
  ,  are specified such that there is a high degree of confidence that the speci-fled acceptable fuel design limits are not exceeded during normal operation
  ;  and design basis anticipated operational occurrences.
B 2-1                      Amendment No. 9 February 27, 1984 1
 
2.1.2        REACTCR COOLANT SYSTEM PRESSURE The restriction of this safety limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel, piping, and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant Components which pennits a maximum transient pressure of 110% (2750 psia) of the design pressure.' The Reactor Coolant System valves and fittings, are designed to either Section III of the ASME Code or ANSI B 31.7, Class I, which permits a maximum transient pressure of 110%(2750 psia) of component design pressure. See Applicant's FSAR for specific Code, Standard Editions, and Addenda. The safety limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
i Amendment Number 9 February 27, 1984
 
i                                                                                                                                                              . ,
2.2                    LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1                          REACTOR TRIP SETPOINTS I
The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal aaaration an g    design basis anticipated operational occurrencetg'Tnfrequent inciden , and to assist the Engineered Safety Features Actuation system in mitigating the
;                  consequences of limiting faults. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable
;                  on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analysis.
l                    The DNBR - Low and Local Power Density - High are digitally generated trip 4
setpoints based on Safety Limits of 1.231 and 21.0 kw/ft, respectively. Since l                    these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to trips generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.
To maintain the margins of safety assumed in the safety analyses, the calcula-i tions of the trip variables for the DNBR - Low and Local Power Density - High trips include the measurement, calculational and processor uncertainties and i                  dynamic allowances as defined in System 80 applicable system descriptions and
;                    safety analyses.
Manual Reactor Trip The manual Reactor Trip is a redundant channel to the automatic protective                                                                    '
instrumentation channels and provides manual reactor trip capability.
Variable Overpower Trip                                                                                                                        !
l                  A reactor trip on Variable Overpower is provided to protect the reactor core 3
during rapid positive reactivity addition excursions. This trip function will trip the reactor when the indicated neutron flux power exceeds either a rate i                  limited setpoint at a great enough rate or reaches a preset ceiling. The flux
!                  signal used is the average of three linear subchannel flux signals originating in each nuclear instrument safety channel. These trip setpoints are provided in Table 2.2-1.
Locarithmic Power Level-High i
,                  The Loparithmic Power Level - High trip is provided to protect the integrity
;                  of fue cladding and the Reactor Coolant System pressure boundary in the event                                                                  i of an unplanned criticality from a shutdown condition. A reactor trip is
~
initiated by the Logarithmic Power Level - High trip unless this trip is manu-
;                  ally bypassed by the operator. The opera
!                  when the THERMAL POWER level is above%10~                                gor may of RATED    THERMAL      manually            bypass POWER; this this trip          i byg%ofRATEDTHERMALPOWER.ss 10                                                            is automatically removed when the THERMAL PO B 2-3                                      g g t 'l        M 8                )
2 i                                                                                                                                                                  l l
 
i Pressurizer Pressure-High                                                                                                                                              l The . Pressurizer Pressure-High trip, in conjunction with the pressurizer safety                                                                                        I valves and main steam safety valves, provides reactor coolant system protection i                            against overpressurization in the event of loss of load without reactor trip.
_                      This trip's setpoint is below the nominal lift setting of the pressurizer safety valves and its operation minimizes the undesirable operation of the pressurizer safety valves, t
  !                          Pressurizer Pressure-Low 1
l The Pressurizer Pressure-Low trip is provided to trip the reactor and to i
assist the Engineered Safety Features System in the event of a decrease in Reactor Coolant System inventory and in the event of an increase in heat removal by the secondary system. During normal operation this trip's setpoint i                        may be manually decreased to a minimum value of 100 psia as pressurizer                                                                                                  i
:                @          pressure is reduced during plant shutdowns, provided the margin between theprej                                                                                        ,
setpoint increases automatically as pressurizer pressure incrFaTes until the7                                                                                          ;
trip setpoint is reached. The operator may manually bypass this trip when pressurizer pressure is below 400 psia; this bypass is automatically removed                                                                                  M    g  (
when the pressurizer pressure increases to 500 psia.                                                                                                                    "
;                          Containment Pressure-High
;                          The Containment Pressure-High trip p ovides assurance that a reactor trip is                                                                                            ,
initiated in the event of containment building pressurization due to a pipe                                                                                              '
break inside the containment building. The setpoint for this trip is identi-cal to the safety injection setpoint.
Steam Generator Pressure I N The Steam Generator Pressure-Low trip provides protection in the event of an increase in heat removal by the sec6ndary system ar.d subsequent cooldown of                                                                                            ,
the reactor coolant. The setpoint is sufficiently below the full load operat-                                                                                            ,
ing , "nt so as not to interfere with normal operation, but still high enough                                                                                            !
to p          ide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator                                                                                                  i j                          pressure is reduced during plant shutdowns, provided the margin be1 ween the                                                                                            :
steam generator pressure and this trip's setpoint is maintained ai d 00 psi; this setpoint increases automatically as steam generator p essure increases T
              @            until the nonnal pressure trip setpoint is reached.                                                                                                        JLwfkam      !
d4 aguaf.
}                          Steam Generator Le' vel-Low                                                                                                                                        *  .)
The Steam Generator Level-Low trip provides protection against a loss uf feed-4 water flow incident and assures that the design pressure of the Reactor Cool-l                          ant System will not be exceeded due to a decrease in heat removal by the                                                                                                ;
!                          secondary system. This specified setpoint provides allowance that there will i              @ be sufficient water inventory in the steam generator at the time of the trip l                          to provide a margin at least 10 minutes before emergency feedwater is required.
l                                                                                                                                                                b c
Amendment Number 9 B24          February 27, 1984
      - , . . _.___.__m-                  ._,.._,.,___._,,,_....___-.-__.-,__..m._.-                                                                  - _ . _ _ . - _ . . . , _ _ _ - -
 
Local Power Pensity-High The Local Power Density-High trip is provided to prevent the linear beat rrte (kw/ft) in the limiting fuel rod in the core from exceeding the fuel desfgn limit in the event of any design basis anticipated operational occurrence.                j The local power density is calculated in the Reactor Protective System utiliz-            )
ing the following information:
: a. Nuclear flux power and axial power distribution from the excore flux                '
monitoring system;                                                                  l l
: b. Radial peaking factors from the position measurement for the CEAs;
: c. AT power from reactor coolant temperatures and coolant flow measurements.
The local power density (LPD), the trip variable, calculated by the CPC incor-          '
porates uncertainties and dynamic compensation routines. These uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the            ;
actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a vio-lation of the peak Linear Heat Rate Safety Limit. CPC uncertainties related to peak LPD are the same types used for DNBR calculation. Dynamic compensa-
  . tion for peak LPD is provided for the effects of core fuel centerline tempera-ture delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
DN8R-Low The DNBR - Low trip is provided to prevent the DNBR in the limiting coolant channel in the core from exceeding the fuel design limit in the event of a design bases anticipated operational occurrence. The DN8R - Low trip incorpor-ates a low pressurizer pressure floor of (*) psia. At this pressure a DN8R -
Low trip will automatically occur. The DN8R is calculated in the CPC utiliz-ing the following infonnation:
: a. Nuclear flux power and axial power distribution from the excore neutron flux monitoring system; i
: b. Reactor Coolant System pressure from pressurizer pressure measurement;
  ,                                    DeMa.'T Q c. Differential temperature (g) power fron, reactor coolant temperature and coolant flow measurements,                                                        3
: d. Radial peaking factors from the position measurement for the CEAs; i
';    e. Reactor coolant mass flow rate from reactor coolant pump speed;
}    f. Core inlet temperature from reactor coolant cold leg temperature measure-y            ments.
l
* See Applicant's SAR.
Amendment Number 9 B 2-5                        February 27, 1984
 
123I i
!                  The DNSR, the trip variable, calculated by th CPC, incorporates various
!                  uncertainties and dynamic compensation routine to assure a trip is initiated prior to violation of fuel design limits. Thes! uncertainties and dynamic compensation routines ensure that a reactor trip occurs when the calculated i
h    core DN8R is sufficiently greater than i '"-''t:'-- ''Mt such that the decrease in calculated core DN8R after the trip will not result in a violation i                  of the DN8R safety limit. CPC uncertainties related to DN8R cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer
!                  equipment processing uncertainties. Dynamic compensation is provided in the i                  CFC calculations for the effects of coolant transport delays, core heat flux
{                  delays (relative to changes in core power), sensor time delays, and protection j                  system equipment time delays.
l                  The DN8R algorithm used in the CPC is valid only within the limits indicated                                                                                                              ;
below and operation outside of these limits will result in a CPC initiated j                  trip.
Parameter                                                                    Limiting Value
: a.        RCS Cold Leg Temperature-Low 4
: b.        RCS Cold Leg Temperature-High
: c.        Axial Shape Index-Positive                                                                      .
i                  d.        Axial Shape Index-Negative l                  e.        Pressurizer Pressure-Low
,                  f.        Pressurizer Pressure-High                                                                          See Applicant's SAR l                g.          Integrated Radial Peaking Factor-Low l                h.          Integrated Radial Peaking Factor-High
:                  1.        Quality Margin-Low                                                                                                                                                              .
t Steam Generator Level-High f.
The Steam Generator Level-High trip provides protection in the event of excess l                feedwater flow. The setpoint for this trip is identical to the main steam isolation setpoint.
4 Reactor Coolant Flow-Low                                                                                                                                                                  '
4 i
The Reactor Coolant Flow-Lcw trip provides protection against a reactor i                coolant pump sheared shaft event and a two-pump opposite loop flow coastdown j
event. A trip is initiated when the pressure differential across the primary side of either steam generator decreases below a variatie setpoint. This j                variable setpoint stays a set amount below the se44 pr9ssure differential j-unless limited by a set maximum decrease rate or a set _ minimum value. The                                                                                                                !
snecified Qinear      leat rate      setpoint      ensures or DNBR                  that alimits safety          reactor          9a4*  tripn-occurt), to prevent violation on M            " aaaditions. M.
:                Pressurizer Pressure-High (SPS)
A% %L sy W.
!                The Supplementary Protection System (SPS) augments reactor protection against i
overpressurization by utilizing a separate and diverse trip logic from the Reactor Protection System for initiation of reactor trip. The SPS will initiate a reactor trip when pressurizer pressure exceeds a predetermined value.
I, 1
4 B 2-6                                Amendment Number 9 4
February 27, 1984
\.
y..      ,    .          _-,,.~..,---,-,,y-,,-          %%-,,,,,e-            ynm..,.-,,.----,-g---,                              _.--.,__v      _ery-.~---,,,,,mm_-._              , - - - , - - -
 
e.tRs tenf6ctgN ChLituAf M C) 2.2-2      APC- ADDRESSARLE CONSTANTS g
The Core Protection Calculator (CPC) addressable constants are p ovided to allow calibration of the CPC system to more accurate indications :::5 :: ::1-power level end RCS flowrate,: d ':nr: =d :=r:
d:txter :i;=?: fr axial flux shape,j radial peaking factors and CEA devia-tion penalties. Other CPC addressable constants allow penalization of the calculated DNBR and LPD values based on measurement uncertainties or inoper-able equipment. Administrative controls on changes and periodic checking of addressable constant values (see also Technical Specifications 3.3.1 and 6.8.1) ensure that inadvertent misloading of addressable constants into the CPC's is unlikely.
r 1
;                                              8 2-7                Amendment Number 9 February 27, 1984 I
 
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CHNICAL SPEC SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l
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l Amendment Number 9 February 27, 1984 u
 
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3/4        LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS 3/4.0        APPLICABILITY LIMITING CONDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding specifications is required during the OPERATIONAL MODES or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a specification shall exist when the requirements of the Limiting Condition for Operation and/or associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration _of the specified time intervals, completion of the ACTION requirements is not required.
  @                                                        G & rd 3.0.3 When a Limiting Condition for Operation ::::;t k n ti;ft:d, except as provided in the associated ACTION requirements, within one hour, action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
: 1. At least HOT STANDBY within the next 6 hours,
: 2. At least HOT SHUTDOWN within the following 6 hours, and
: 3. At least COLD SHUTDOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.
This specification is not applicable in MODE 5 or 6.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as requir-ed to comply with ACTION statements. Exceptions to these requirements are stated in the individual specifications.
SURVEILLANCE REQUIREMENTS 4.0.1    Surveillance Requirements shall be. applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:
Amendment Number 9 February 27, 1984 3/4 0-1
: a. A maximum allowable extension not to exceed 25% of the surveillance interval,-and
: b. The combined time interval for any 3 consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
4.0.3' Failure to perfom a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be perfomed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise specified.
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2 and 3 components shall be applicable as follows:
: a. Inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50 Section 50.55a(g), except where specific written relief has been granted by the Consnission pursuant to 10 CFR 50, Section 50.55a(g)
(6)(1).
: b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
ASME Boiler and Pressure Vessel Code and applicable          Required frequencies Addenda terminology for              for performing inservice inservice inspection and            inspection and testing testing activities                      activities Weekly                        At least once per 7 days Monthly                        At least once per 31 days Quarterly or every 3 months          At least once per 92 days Semiannually or every 6 months        At least once per 184 days Yearly or annually            At least once per 366 days
: c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for perfoming ' inservice inspection and testing activities.
Amendment Number 9 3/4 0-2                      February 27,1984 t.
 
n_
: d. Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
: e. Nothing in the ASME Boiler and Pressure Vessel Code shall be con-strued to supersede the requirements of any Technical Specification.
t Amendment Number 9 February 27, 1984 3/4 0-3 i
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3/4.1          REACTIVITY CONTROL $YSTEMS 3/4.1.1          80 RAT!0N CONTROL SHUTDOWN MARGIN - T        GREATER THAN 210*F cold LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 6.0% delta k/k.
APPLICA8ILITY: MODES 1,2*, 3 and 4.
ACTION:
iI . '                  ritical)                          #
Comply with            -
ification 3.1.3.6 (b          the shutdown margin less than 6% delta                        K    <1 (i.e.. Modes 2. 3 and 4 subtritical) jWiately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal
  @      to 4000 ppm boron or equivalent until the required shutdown Lmargin is restored.
          %i% % seut00s3 HMIGlN Jus.bs~ 4.0% W W)
    $URVEILLANCE REQUIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal                        '
to 6.0% delta k/k:
: a. Within one hour after detection of in inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA is inovable or untrippable, the above required SHUT-DOWN MARGIN shall be verified acceptable with an increased allowance
              , for the withdrawn worth of the immovable or untrippable CEA(s),
: b. When in MODE 1 or MODE 2 with K                  greater then or equal to 1.0, at least once per 12 hours by veriffing that CEA group withdrawal is within the Transient Insertion Limits of Specification 3.1.3.6.
: c. When in MODE 2 with X                  less than 1.0, within 4 hours prior to achieving reactor criII{ality by verifying that the predicted critical CEA position is within the limits of Specification 3.1.3.6.
see special Test Deeption 3.10.1.
Amendment Number g 3/4 1-1                      February 27, 1984  l l
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: d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.
: e. When in MODES 3 or 4, at least once per 24 hours by consideration of at least the following factors:
: 1. Reactor coolant system boron concentration,
: 2. CEA position,
: 3. Reactor coolant system average temperature,
: 4. Fuel burnup based on gross thermal energy generation,
: 5. Xenon concentration, and
: 6. Samarium concentration.
4.1.1.1.2 The overall core reactivity balance shall be compared to predicted valuestodemonstrateagry              within + 1.0% delta k/k at least once per 31 Effective Full Power Dayq.          s comparTson shall consider at least those
@ factors stated in Specif' cation 4.1.1.1.1.e. above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core condi-tions prior to exceeding a fuel burnup of 60 Eff;;;iv. T.li F=;- 4: after each fuel loading.                                                EFP) i l
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'                                                                          Amendment Number 9 3/4 1-2                      February 27, 1984
 
4.-
                                                          ,_                                              l t
i SHUTDOWN MARGINST                LESS THAN OR EQUAL TO 210*F cold LIMITING CONDITION FOR OPERATION                                                                              l i
1 3.1.1.2o The SHUTDOWN MARGIN shall be greater than or equal to 4.0% delta k/k.
1 APPLICABILITY: ; MODE 5.
8
                                                                                                                \                l
                  . ACTION:'
With the SHUTDOWN MARGIN less than 4.0% delta k/k, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 4000 ppm boron or equivalent until the required                                      j SHUTDOWN MARGIN is restored.                        ,                                                        i 1                                        e SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to_be greater than or equal to 4.0% delta k/k:          j                                                                                t
: a.      Within one hour'after detection of an inoperable CEA(s) and at least                          l l
once per 12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA 'is immovable or untrippable, the above required          1                1
                                  , SHUTDOWN MARGIN shall be increased by an amount at least equal to            '
the withdrawn wdcth of the immovable or untrippable CEA(s).                              .f
: b.      ' At leastionce per 24 hours by consideration of the following factors:
                                    +,,
3,
: 1. Reactor coolant system boron concentration, l-                                  2. CEA position,
      't
: 3.      Reactor coolant system average temperature,
: 4.      Fuel burnup based on gross thermal energy generation,
                  ).                5. Xenon concentration, and
: 6.      Samarium concentration.
4'.1. . .                  11 be determined to be eoual to or less_than 0 9A a+ has$
                                                                            ~
oice per 24 ll rs
    '@              low leve
                                                              %w+ar level is drained below the pressurizer ment tap, by performinghivity_ balance considering the listed in 4.1.1.2.lb.
: t.      A j
              -                                                                                      \
N l
                                                                                                                    \
                            ~
Amendment Number 9 '        Y 3/4 1-3                  February 27,1984 l                                        s
_ -- % -g
 
y
                      - MODERATOR TEMPERATURE COEFFICIENT f              \    LIMITING CONDITION FOR OPERATION u
a    3.1.1.3 The moderator temperature coefficient (MTC) shall be within the area of Acceptable Operation on Figure 3.1-1.
      ~l                APPLICABILITY: MODES 1 and 2*#.
  ,y.                  ACTION:
With the moderator temperature coefficient outside the area of Acceptable Operation on Figure 3.1-1, be in at least HOT STANDBY within 6 hours.
I SURVEILLANCE REQUIREMENTS 4.1.1. 3.1 The MTC shall be determined to be within its limits by confimatory measurements. MTC measured values shall be extrapolated and/or compensated to                                    !
pemit direct comparison with the above limits.                                                                    '
4.1.1. 3. 2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:
>3;                          a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading.
g      b. At any THERMAL POWER, within 7 EFPD after reaching a core average exposure of 40 Effective Full P = = hy: burnup into the current cycle.                  EE P P_D
: c. At any THERMAL POWER, within 7 EFPD after reaching a core average exposure equivalent to 2/3 of the expected end-of-cycle core average burnup.
L l
* With K,ff greater than or equal to 1.0.
                        # See Special Exception 3.10.2.
Amendment Number 9 February 27, 1984 3/4 1-4
 
I SEE APPLICANT'S SAR i
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FIGURE 3.1-1                                                              {
ALLOWABLE M1C MODES 1 AflD 2                                                      l I
Amendment Number 9 3/4 1-5                                            February 27, 1984
 
1 i
MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant System lowest operating loop temperature T shall be greater than or equal to 552*F.                                                    cold APPLICABILITY: MODES 1 and 2*f.
ACTION:
less than
@ With a Reactor 552*F,          Coolant System operating loop temperature restore T                                                  e    Ttowithinitslimitwithin15 STANDBY within thS Next 15 minutes.
SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant System temperature (Tcold) shall be detemined to be greater than or equal to 552*F:
: a. Within 15 minutes prior to achieving reactor criticality, and
: b. At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T cold is less than 557'F.
* See Special Test Exception 3.10.5                                                                l
  # With K,ff greater than or equal to 1.0.                                                          l Amendment Number 9 February 27, 1984
                                                            . - . . - -    . - - . ..-.-. - - -.      E
 
3/4.1.2                      B0 RATION SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.1      As a minimum, one of the following boron injection flow paths shall be OPERABLE:
: a.            IfonlythespentfuelpoolinSpecification1                                                is OPERABLE, g                                      or a flow path from the spent fuel pool via                          gravitya)3.1.2.55 fee      )d connection and a charging pump to the Reactor Coolant System.
: b.            If only the refueling water tank in Specification D3.1.2.55bh)is OPERABLE, a flow path from the refueling water tank via either a charging pump or a high pressure safety injection pump or a low pressure safety injection pump to the Reactor Coolant System.
APPLICABILITY: MODES 5 and 6.
ACTION:
With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes,                                                                                . ,
SURVEILLANCE REQUIREMENTS 4
4.1.2.1      At least one of the above required flow paths shall be demonstrated OPERABLE:
At least once per 31 days be verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
4 Amendment Number 9 3/4 1-7                              February 27, 1984
                                                                                      ,v- - - -
                                                                                                -m---  , - --,-e .. -,,-.--n    ,            .  , - , , <
 
FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:
: a. A gravity feed flow path from either the Refueling Water Tank or the Spent Fuel Pool through CH-536 (RWT Gravity Feed Isolation Valve) and a charging pump to the Resctcr Coolant System,
: b. A gravity feed flow path from the Refueling Water Tank through
.            CH-327 (RWT Gravity Feed / Safety Injection System Isolation Valve) and a charging pump to the Reactor Ccolant System,
: c. A flow path from either the Refueling Water Tank or the Spent Fuel Pool through CH-161 (Boric Acid Filter Isolation Valve) or CH-164 (Boric Acid Filter Bypass Valve), utilizing gravity feed and a charging pump to the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least kOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 6%
ak/k at 210*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 kours.
SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:
: a. At least once per 31 days be verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
: b. At least once per 18 months when the Reactor Coolant System is at          I nomal operating pressure by verifying that the flow path required by Specification 3.1.2.2 delivers : tet:1 of at least 40 ; F to the Reactor Coolant System.                                  A      -      m g                                                    for Io                3M q*3PJ * **nt" O Amendment Number 9 3/4 1-8                    F9bruary 27, 1984  l
 
l I
    ' CHARGING PUMP - SHUTDOWN                                                                  1 LIMITING CON 0! TION FOR OPERATION 3.1. 2. 3 At least one charging pump
* or one high pressure safety injection              ,
pump or one low pressure safety injection pump in the boron injection flow                l path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and            '
capable of being powered from an OPERA 8LE emergency power source.
APPLICA8ILITY: MODES 5 and 6.
ACTION:                                                                                    ;
With no charging pump or high pressure safety injection pump or low pressure              !
safety injection pump OPERA 8LE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
I SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by Specification 4.0.5.                                          s l
l
* Whenever the reactor coolant level is below the botton: of the pressurizer I
in Mode 5. one and only one charging pump shall be OPERABLE by verifying at least once per 7 days that power is removed from the remaining charg-ing pumps.
i i
i                                                                                                !
i Amendment Number 9 f f                                                                              February 27,1984 l                                            3/4 1-9 l
 
CHARGING PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERNBLE.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With only one charging pump OPERABLE, restore at least two charging pumps to OPERA 8LE status within 72 hours or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 6% delta k/k at 210'F within the next 6 hours; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
SURVEILLANCE REQUIREMENTS 4.1.2.4 No additional Surveillance Requirements other than those required by Specification 4.0.5.
Amendment Number 9 3/4 1-10                                February 27, 1984
 
L A                                                                                                ,
BORATED WATER SOURCES - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:
: a. The spent fuel pool with:
: 1. A minimum borated water volume as specified in Figure 3.1-2, and
: 2. A boron concentration of between 4000 and 4400 ppm boron, and
: 3. A solution temperature between 60'F and 180'F.
: b. The refueling water tank with:
: 1. A minimum contained borated water volume as specified in Figure 3.1-2,
: 2. A boron concentration of between 4000 and 4400 ppm boron, and
: 3. A solution temperature between 60'F and 120*F.
APPLICABILITY: MODES 5 and 6.
ACTION:
With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one borated water
      . source is restored to OPERABLE status.
SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:
: a. At least once per 7 days by:
: 1. Verifying the boron concentration of the water, and
: 2. Verifying the contained borated water volume of the refueling water tank or the spent fuel pool.
(        @M
: b. At least once per 24 hours by verifying the      4temperature when it b              is the source of borated water and the outside air temperature is outside the 60*F to 120*F range.
: c. At least once per 24 hours by verifying the spent fuel pool tempera-            ,
ture when it is the source of borated water and irradiated fuel is            !
present in the pool.                                                            l Amendment Number 9 3/4 1-11                    February 27, 1984 l
 
SEE APPLICANT'S SAR t
1 P
FIGURE 3.1-2 MINIMUM BORATED WATER VOLUMES i
j Amendment Number 9 f                                                                                                                                                            February 27,1984 3/4 1-12
 
BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 Each of the following borated water sources shall be OPERABLE:
: a.              The spent fuel pool with:
w,_
: 1.          A minimum borated water volume as specified in Figure 3.1-2, and
: 2.          A boron concentration of between 4000 and 4400 ppm boron, and
: 3.          A solution temperature between 60*F and 180?F.
: b.              The refueling water tank with:
  ,                                        1.          A minimum contained borated water volume as specified in Figure 3.1-?, and
: 2.          A boron concentration of between 4000 and 4400 ppm boron, and
: 3.          A solution temperature between 60'F and 120*F.
APPLICABILITY: MODES 1, 2, 3 and 4.
AC1 ION:
: a.            With the above required spent fuel pool inoperable, restore the pool to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 6% ak/k at 210*F; restore the above required spent fuel pool to OPERABLE status within the next 7 days or be in COLD SHUT-DOWN within the next 30 hours.
: b.            With the refueling water tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS I
4.1.2.6 Each of the above required borated water sources shall be demonstrat-ed OPERABLE:                                                                                                                                                                      I
: a.            At least one per 7 days by:
: 1.            Verifying the boron concentration in the water, and
: 2.            Verifying the contained borated water volume in the water
,                                                      source.
I Amendment Number 9 February 27, 1984 i                                                                                                                                                                                                      l
    .m -----w,----.          ....,,_,,,_.,,wyr.,_,,.r,            ...m.m_..
                                                                              . . . , , , _ - - ,- - . , , ... , . . .._. ..,, - ., , . - _ .-,_,,.__,. , . , , -      re,  --,__.--.3. _ , -_,,.,-_.y
 
F MIM
    @ b. At least once per 24 hours by verifying theb temperature when the outside air temperature is outside the 60*F to 120*F range,
: c. At least once per 24 hours by verifying the spent fuel pool tempera-h    ture when Q is the source of borated water anaJirradiated fuel is present in the pool.                      g Amendment Number 9 3/4 1-14                    February 27,1984
 
BORON DILUTION ALARMS LIMITING CONDITION FOR OPERATION 3.1.? 7 Both startup cb!au M@ :::t-aallux alams shall be OPERABLF, nd g      t to alarm when the ratio of the measured flux to the flux at K g less than or equal to 4.7.
c fied APPLICABILITY: MODES        3**, 4, 5, and 6.
C A_CTION:
: a. With one startup channel high neutron flux alam not OPERABLE:
: 1. Determine the RCS boron concentration when entering MODE 3, 4, 5 or 6 or at the time the alam is detemined to be inoperable.
From that time, the RCS boron concentration shall be determined at the applicable monitoring frequency in Table 3.1-1,3.1-2, 3.1-3, 3.1-4, or 3.1-5; depending upon the degree of suberiti-cality available by either boronometer or RCS sampling *.
: b. With both startup channel high neutron flux alarms not OPERABLE:          A
: 1.      ermine the RO Doron wnwnirsi.lon by 50iH bwenomette-an RCSlampQng* when entering MODE 3, 4 orTor at the ti              o alams are determined to be inoperable. From that              he (C                RCS boron concenbattion shall be determined a          applicable monitoring frequency in' Table 3.1-1, 3.L-2d 1-3, 3.1-4, or l                          3.1-5; depending upon the deg?eegf sub' criticality available by both baronometer and RCS sampli'ng* ^!fsone of the methods of determining the RCS boron' concentration ibnot available, imediately suspend'all operations involving CORQLTERATIONS i
or positi s-reactivity changes until at least one additional
;                          metho    or detecting a boron dilution is restored to OPERA 8LE s    s.
l p teur Lu>ll .j
: 2. When in MODE 5 with the RCS dnt:d r te the centerline of g                    the hotleg or MODE 6 suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one startup channel high neutron flux alam is restored to OPERABLE i                          status,
: c. The provisions of Specification 3.0.3 are not applicable.
* With one or more reactor coolant pumps (RCP) operating, the sample should be obtained from the hot leg. With no RCP cperating, the sample should be obtained from the discharge line of the low pressure safety injection (LPSI)pumpoperatingintheshutdowncoolingmode.
        **      This specification is not 1pplicabirduring thr_ firs.t20~iETEftes af te)
Q
                                                        ~ ~ ~ ~
entering MODE __31rrom MODE 2r ny    3, ju 4tu, yis .yruJru, ftpx. h UN*, N M*f M14 40% a Waufr ANAvh                                                Amendment Number 9 February 27,1984 3/4 1-15
 
O
                                          /
IDetermine the RCS boron concentration by either boronmeter and i RCS sampling ** or by independent collection and analysis of two (
RCS samples when entering Mode 3, 4, or 5 or at the time both                i From that time, the        i falarmsaredeterminedtobeinoperable.
RCS boron concentration shall be determined at the applicable                  !
                                / monitoring frequency in Tables 3.1-1 through 3.1-5, as applicable,.
I by either baronmeter and RCS sampling ** or by collection and analysis of two independent RCS samples. If redundant determina-tion of RCS t'oron concentration cannot be accomplished immediately, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the method for determining and confirmingj RCS boron concentration is restored.                                          '
I
 
SURVEILLANCE REQUIREMENTS 4.1.2.7 Each startup channel high neutron flux alarm shall be demonstrated OPERA 8LE by performance of:
: a. A CHANNEL CHECK M
: 1. (Once per      hours.
: 2.          15 minutes arter settina tne alarm setpoints)
: b. A CHANNEL FUNCTIONAL TEST every 31 days of cumulative operation during shutdown.                                                          -
@  q au              wnq m e a u p %
                ) 9,ut,- W N tlk w A A(Jer~ irf a.
b) QL4 a e.enk=Utd raatfer,shuddeum,5 Mih 1. M q u m h p w Mhd.4 Th SfW fy M V10DG 3-(
1 Amendment Number 9 3/4 1-16                      February 27, 1984
 
F TABLE 3.1-1: MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR SY5ILM 80 tKitNUED FUEL CYCLE FOR K ,, > 0.98.FOR MODES 3, 4, AND 5 TIME PERIOD IN HOURS AT WHICH MONITORING IS REQUIRED OPERATIONAL        '
WITH THE FOLLOWING NUMBER OF CHARGING PUMPS OPERATING MODE O                      1                      2                      3 3                              12.0                      1.0                      OPERATION NOT ALLOWED
( Hot Standby )
4                              12.0                      1.0                      OPERATION NOT ALLOWED
( Hot Shutdown )
* 5                                8.0                      1.0                      OPERATION NOT ALLOWED
{    ( Cold Shutdown )
5                                                        OPERATION NOT ALLOWED
( RCS PARTIALLY DRAINED FOR SYSTEM REPAIRS )
6*                                24.0                    8.0                    4.0                    2.0
( Refueling )
A conservative value of an initial boron concentration is assumed which is bounded by the Technical Specification 3.9.1. Furthermore, during refueling the LPSI pumps should be used for any makeup operation.
If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be Lsed.
w.
EI Q= "
                            ~
N2 S
sl s
 
TABLE 3.1-2: MONITORING FREQUENCIES FOR BACKUP BOP.0N DILUTION DtitCTION FOR SYSTEM 80 EAltNUED FUEL CYCLE FOR 0.98 2 Kg > 0.97 FOR PODES 3, 4, AND 5 l                                                                                                                    TIME PERIOD IN HOURS AT WHICH MONITORING IS REQUIRED i                                                                          OPERATIONAL                              WITH THE FOLLOWIN C RUMBER OF CHARGING PUMPS OPERATING M0'JE 4
0                        1                    2                    3 3                              12.0                      2.5                    1.0                    0.5 1                                                                            ( Hot Standby )                                          ,
4                              12.0                      2.5                    1.0                    0.5
( Hot Shutdown )                                                        .
R
* 5                                8.0                      2.5                    1.0                    0.5
( Cold Shutdown )
{
5                                8.0                      0.5                        OPERATION NOT ALLOWED (RCSPARTIALLYDRAINgD 1
FOR SYSTEM REPAIRS )
i 6*                                24.0                      8.0                    4.0                  2.0
( Refueling )
The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and plant condition.
                  ;p;a-
* A conservative value of an initial boron concentration is assumed which is bounded by the Technical g5                                                              Specification 3.9.1. Furthennore, during refueling the LPSI pumps should be used for any makeup operation.
gE                                                                If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used.
4! -
                ?g d=,
4
 
TABLE 3.1-3: MONITORING FREQUENCIES FOR BACKUP BORON DILUTION                          i DETECTTGIITOR SYSTEM 80 ExitmiED FUEL CYCLE FOR 0.97 2 K    > 0.% FOR PWDES 3, 4, AND 5 TIME PERIOD IN HOURS AT WHICH MONITORING IS REQUIRED OPERATIONAL                    -
WITH THE FOLLO6(DilDIUMBER OF CHARGING PUMPS OPERATING MODE O                        1                    2                3 3                  .          12.0                      3.5                  1.5              1.0
( Hot Standby )
4                              12.0                      3.5                  1.5              1.0
( Hot Shutdown )
    $          ( Cold Shutdown )
(RCSPARIALLYDRAIN{D FOR SYSTEM REPAIRS )
6*                                24.0                      8.0                  4.0
                                                                                                                            ^
2.0
( Refueling )
The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and plant condition.
mg                    A conservative value of an initial boron concentration is assumed which is bounded by the Technical (g                    Specification 3.9.1. Furthermore, during refueling the LPSI pumps should be used for any makeup operation.
gg                    If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used, w :s vo NE a
P.
 
TA8tE 3.1-4: MONITORING FREQUENCIES FOR BACKUP BORON DILUTION UtILLII(M POR 5T51tR 80 txitimED FUEL CYCLE tuu 0.% > Kg > 0.95 FOR EDES 3, 4, AE 5 TIME PERIOD IN HOURS AT leflCH MONITORING IS REQUIRED OPERATIONAL                                      WITH THE FOLLOWTlilDium ER OF CHARGING PUMPS OPERATING MODE O                        1                      2                              3 3                                  12.0                      5.0                    2.0                            1.0
( Hot Standby )
4                                  12.0                      5.0                    2.0                            1.0
( Hot Shutdown )
          $                            5                                  8.0                      5.0                    2.0                            1.0 A,          ( Cold Shutdown )
o 5                                  8.0                      1.5                      OPERATION NOT ALLOWED (RCSPARTIALLYDRAIN(D FOR SYSTEM REPAIRS )
6*                    '
24.0                      8.0                    4.0                          2.0
( Refueling )
* The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and plant condition.
              ,y
* A conservative value of an initial boron conceptration is assumed which is bounded by the Technical 33                Specificattein 3.9.1. Furthermore, during refueling the LPSI pumps should be used for any makeup operation.
2 R.              If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used.
              %2 w=
ro M3 e
e
 
l l
l l
TABLE 3.1-5: MINITORING FREQUDeCIES FOR BACKUP BORON DILUTION UtItcTIM tou SY3Itn W txituutD FUEL CYCLE tout K        5 0.95 tout ruuuth 3, 4, AM 5 y
1 l
TIE PERIOD IN HOURS AT 18tICH MONITORING IS REQUIRED OPOtATIONAL                              WITH THE FOLLO 6 ER OF CHAAGING PUMPS OPERATING N00E O                            1              2                        3 3                                12.0                          6.0            3.0                      1.5
( Hot Stan ey )
l 4                                  12.0                          6.0            3.0                      1.5
( Hot Shutdown )
u 1                              5                                  8.0                          6.0            3.0                      1.5 7                ( Cold Shutdown )
m
                          ~
5                                  8.0                          2.0                OPERATION NOT ALLOWED l
(RCSPARTIALLYDRAIy)
FOR SYSTEM REPAIRS l
l 6*                                24.0                          8.0            4.0                  2.0
( Refueling )
* The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and plant condition.
* A conservative value of an initial boron concentration is assumed which is bounded by the Technical
                                      ,i ;t.
S- 4        Specification 3.9.1. Furthermore, daring refueling the LPSI pumps should be used for any makeup operation.
If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used.
m.
su E
a f.4
 
3/4.1.3        MOVA8LE CONTROL ASSEMBLIES CEA POSITION LIMITING CON 0! TION FOR OPERATION 3.1.3.1 All full length (shutdown and regulating) CEAs, and all part length CEAs which are inserted in the core, shal          be OPERA 8LE with each CEA of a given group positioned within 6.6 inches (indicated position) of all other CEAs in its group.
APPLICA8!LITY: MODES 1* and 2*.
ACTION:
: a. With one or more full length CEAs inoperable due to being immovable as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within I hour or be in at least H0T STAN08Y within 6 hours.
: b. With more than one full length or part length CEA inoperable or misaligned      from any)other (indicated position                CEAHOT
                                    , be in at least  in itsSTAND group 8Y by within more 6than 19 inches hours.
With one full length or part length CEA misaligned from any ot (EA in its group by more than 19 inches, operation in MODES '              2 m      ontinue, provided that within one hour the misali            A is eithe
: 1. Restoree t 0PERA8LE status within its a ove specified alignment (requ rements, or
: 2. Declared inoperable      d      SHUTDOWN MARGIN requirement of Specification 3.1.1.1        satisfied. After declaring the CEA g                  inoperable, opera the requiremen in MOSES 1 and 2 may continue pursuant to of Specificatto      1.3.6 provided:
a)    Wit    one hour the remainder ofsthe CEAs in the group w    the inoperable CEA shall be 11tgned to within 6.6 nches of the inoperable CEA while maintaining the allow-able CEA sequence and insertion limits'thown on Figure            i 3.1-3, the THERMAL POWER Level shall be r4 ricted pursu-          l ant to Specification 3.1.3.6 during subseque operation.          l
: b. The SHUTDOWN MARGIN requirement of Specification        1.1.1 is determined at least once per 12 hours.
Otherwise, be in at least H0T STAND 8Y within 6 hours.
l 5ee 5pecial Test Exceptions 3.10.2 and 3.10.4.
3/4 1-22                  Amendment Number 9 February 27, 1984
 
A f
c.
!                    %      With one or more full-length or part-length CEAs misaligned from any other CEAs in its group by more than 6.6 inches '"' " '' - --
.                            :-":1 t: 19 '-9 , operation in MODES 1 and 2 may continue, provid-l                            ed that within one hour the misaligned CEA(s) is either:
: 1.      Restored to OPERABLE status within its above specified align-ment requirements, or
: 2.      Declared inoperable and the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. After declaring the CEA
;                                    inoperable, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6 provided:
1 a)            Within one hour the remainder of the CEAs in the group with the inoperable CEA shall be aligned to within 6.6                                              1 inches of the inoperable CEA while maintaining the allow-                                          '
able CEA sequence and insertion limits shown on Figure 3.1-3; the THERMAL POWER' level shall be restricted pursu-
,                                                  ant to Specification 3.1.3.6 during subsequent operation.
b)            TheSHUTDONNMARGINrequirementofSpecification3.1.1.1                                                  i 4
3 is determined at least once per 12 hours.
Otherwise, be in at least HOT STANDBY within 6 hours.                                                                      !
r                      el
:                      %    With one full length CEA inoperable due to causes other than addressed by ACTION a. above, and inserted beyond the Long Term Steady State                                                        !
Insertion Limits (Figure 3.1-3) but within its above specified                                                            ;
alignment requirements, operation in MODES 1 and 2 may continue                                                            i pursuant to the requirements of Specification 3'.1.3.6.
b-444ttant_ full length CEA inoperable due to causes other +h*D                                                                  I 4                            dressed by n iivn .. d r:t en ure that                                      s          )withinits                        l
: 2) either fully with-l
              @              above specified ali nment drawn or, if                              n th CEA g o        5, wit                                Tenn Steady l'                  on Limits ( igure 3. -
Sta                                                        . Then operation                              1              i
                              ~n d 2 may continue.                                                              ____.____
l e.
;                      g. With one part length CEA inoperable and inserted in the core,
;                            operation may continue provided the alignment of the inoperable PLCEA is maintained within 6.6 inches (indicated position) of all other PLCEAs in its group.
SURVEILLANCE REQUIREMENTS
.                4.1.3.1.1 The position of each full length and part len determined to be within 6.6 inches (indicated position)ofgth                            all CEA othershall        CEAsbein i                its group at least once per 12 hours, except during time intervals when one
<              CEAC is inoperable or when both CEACs are inoperable; then verify the individ-
;                ual CEA positions at least once per 4 hours.
1 l                                                                                                                          Amendment Number 9 February 27, 1984 l
t
 
4.1.3.1.2 Each full length CEA not fully inserted and each part length CEA which is inserted in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 31 days.
Amendment Number 9 3/4 1-24                  February 27,1984
 
POSITION INDICATOR CHANNELS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.3.2 At least two of the following three CEA position indicator channels shall be OPERABLE for each CEA:
: a. CEA Reed Switch Position Transmitter (RSPT 1) indication with the capability of determining the absolute CEA positions within 5.2 inches,
: b. CEA Reed Switch Position Transmitter (RSPT 2) indication with the capability of determining the absolute CEA positions within 5.2 inches, and
: c. The CEA pulse counting position indicator channel.
APPLICABILITY: MODES 1 and 2.
ACTION:
With a maximum of one CEA per CEA group having only one of the above required CEA position indicator channels OPERABLE, within 6 hours either:
: a. Restore the inoperable position indicator channel to OPERARLE status, or
: b. Be in at least HOT STANDBY, or
: c. Position the CEA group (s) with the inoperable position indicator (s) at its fully withdrawn position while maintaining the requirements of Specification 3.1.3.1 and 3.1.3.6. Operation may then continue provided the CEA group (s) with the inoperable position indicator (s) is maintained fully withdrawn, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2, and each CFA in the group (s) is verified fully withdrawn at least once per 12 hours thereafter by its " Full Out"* limit.
SURVEILLANCE REQUIREMENTS 4.1.3.2 Each of the above required position indicator channels shall be detemined to be OPERABLE by verifying that for the same CEA, the position indicator channels agree within 5.2 inches of each other at least once per 12 hours, s
* CEA's are fully withdrawn (" Full Out") when withdrawn to at least 144.75 inches (193 steps).
3/4 1-25                Amendment Number 9 February 27, 1984
 
POSITION INDICATOR CHANNEL - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 At least one CEA Reed Switch Position Transmitter indicator channel shall be OPERA 8LE for each shutdown, regulating or part length CEA not fully inserted.
APPLICA81LITY: MODES 3*, 4* and 5*.
ACTION:
With less than the above required position indicator channel (s) OPERA 8LE, inmediately open the reactor trip breakers.
SURVEILLANCE REQUIREMENTS 4.1.3.3 The above required CEA Reed Switch Position Transmitter indicator.
channel (s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.
* W1tn the reactor trip breakers in the closed position.
Amendment Number 9 February 27,1984 3/4 1-26
 
CEA DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 Theindividualfulllength(shutdownandregulating)CEAdroptime, from a fully withdrawn position, shall be less than or equal to 4.0 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90 percent insertion position with:
                                                        .L
        @      a. Tcold greater than or equal to 55(*F, and
: b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2.
ACTION:
: a. With the drop time of any full length CEA determined to exceed the above limit, restore the CEA drop time to within the above limit prior to proceeding to_M00E_ Land.2.
                . WitY'tWe' TEA-droptimeswithinlimitsbutdate I      full reactor coolant' flows-operation          e  provided THERMAL    ;
POWER is restricted to les Lthen W equa to the maximum THERMAL t
(      POWER level allowahle-f6Fthe reactor coo a'nt pump combination l                    peratyt-tMe' time of CEA drop time determinat1A l        SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full length CEAs shall be demonstrated through measurement prior to reactor criticality:
: a. For all CEAs following each removal and reinstallation of the reactor vessel head,
: b. For specifically affected individuals CEAs following any maintenance on or modification.to the CEA drive system which could affect the drop time of those specific CEAs, and                                ,
: c. At least once per 18 months.
l l
Amendment Number 9 February 27,1984
 
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2.0        SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS                        i 2.1        SAFETY LIMITS
                                                                                          \
2.1.1          REACTOR CORE 2.1.1.1          DNBR The calculated DNBR of the. reactor core shall be maintained > l.231.
APPLICABILITY: MODES 1 and 2.                        ,'
3
{
L ACTION:
* 1 l
Whenever the calculated DNBR of the reactor core has decreased to less than 1.231, be in HOT STANDBY within I hour and comply with the requirements of
[
        .,              Specification 6.7.1.
3 2.1.1.2      ,  PEAK LINEAR HEAT RATE The ' peak linear heat rate (adjusted for fuel rod dynamics) of the fuel shall be mairittined < 21.0 kw/ft.
            ,.          APPL.ICABILITY: MODES 1 and 2.
ACTION:
                                              \
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21.0 kw/ft, be in HOT STANDBY within I hour and comply with the, requirements of Specification 6.7.1.
2.1.2          REACTOR COOLANT SYSTEM PRESSURE The Reactor Coolant System pressure shall not exceed 2750 psia.
l                        APPLICABILITY: MODES 1, 2, 3, 4. and 5.                                  '
I                                  L                        '
i
      -t'                MODES 1 and 2 Whenever thi Reactor Coolant System pressure has exceeded 2750 psia, be in HOT f
l                        STANDBY with the Reactor Coolant System pressure within its limit within 1 hour and comply with the requirements of Specification 6.7.1.
MODES 3, 4, and 5 Whenever the Reactor Coolant System pressure has exceeded 2750 psia reduce the Reactor Coolant System pressure to within its limit within'5 minutes and comply with the requirements of Specification 6.7.1.
2-1  Amendment Number 9 February 27, 1984
 
2.2        LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.
APPLICABILITY: As shown for each channel in Table 3.3-1.
ACTION:
With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specifica-tion 3.3.1 until the channel is restorea to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
l CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS 2.2.2 Core Protection Calculator Addressable Constants shall be in accordance with Table 2.2-2.
I APPLICABILITY: As shown for Core Protection Calculators in Table 3.3-1.
ACTION:
                                                                                                                          )
With a Core Protection Calculator Addressable Constant less conservative than the value shown in the Allowable Value column of Table 2.2-2, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status.
l l
f Amendment Number 9 February 27, 1984
 
TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT                            TRIP SETPOINT                          ALLOWABLE VALUES I. TRIP GENERATION A. Process          .
: 1. Pressurizer Pressure - High                  See Applicant's SAR                      See Applicant's SAR
: 2. Pressurizer Pressure - Low (2)                See Applicant's SAR                      See Applicant's SAR
: 3. Steam Generator Level - Low (4)              See Applicant's SAR                      See Applicant's SAR
: 4. Steam Generator Level - High (11)            See Applicant's SAR                      See Applicant's SAR
    ,'."            5. Steam Generator Pressure - Low (3)            See Applicant's SAR                      See Applicant's SAR
: 6. Containment Pressure - High                  See Applicant's SAR                      See Applicant's SAR
: 7. Reactor Coolant Flow - Low (8)                See Applicant's SAR                      See Applicant's SAR See Applicant's SAR                      See Applicant's SAR
'              g          a b
Floor (7)
Rate (7)                                  See Applicant's SAR                      See Applicant's SAR c  Band (7)                                  See Applicant's SAR                      See Applicant's SAR
: 8. Local Power Density - High (5)                See Applicant's SAR                      See Applicant's SAR
: 9. DNBR - Low (5)                                See Applicant's SAR                    See Applicant's SAR yP      B. Excore Neutron Flux
,      a.g
: 1. Variable Overpower Trip (9)                    See Applicant's SAR                    See Applicant's SAR
      !f E    g          a  ceiling'(10)                              See Applicant's SAR                    See Applicant's SAR y2                      Rate (10)                                  See Applicant's SAR                    See Applicant's SAR i      m h                c  Band (10)                                  See Applicant's SAR                    See Applicant's SAR
{%.
 
TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT                              TRIP SETPOINT              ALLOWABLE VALUES
: 2. Logarithmic Power Level - High (1)            See Applicant's SAR          See Applicant's SAR a) Startup and Operating                      See Applicant's SAR          See Applicant's SAR b) Shutdown                                    See Applicant's SAR          See Applicant's SAR C. Core Protection Calculator System                                                                        ,
: 1. CEA Calculators                                  Not Applicable              Not Applicable
: 2. Core Protection Calculators                      Not Applicable              Not Applicable D. Supplementary Protection System
  $                  1. Pressurizer Pressure - High                    See Applicant's SAR          See Applicant's SAR II. RPS LOGIC A. Matrix Logic                                          Not Applicable              Not Applicable B. Initiation Logic                                      Not Applicable              No't Applicable III. RPS ACTUATION DEVICES A. Reactor Trip Breakers                                Not Applicable              Not Applicable B. Manual Trip                                          Not Applicable              Not Applicable E{
w.
C Q.
  *a 1
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SHUTDOWN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown CEAs shall be withdrawn to at least 144.75 inches (193 steps).
APPLICABILITY: MODES 1 and 2*f.
ACTION:
With a maximum of one shutdown CEA withdrawn to less than 144.75 inches (193 steps), except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either:
: a. Withdraw the CEA to at least 144.75 inches (193 steps), or
: b. Declare the CEA inoperable and apply Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to at least 144.75 inches (193 steps).
: a. Within 15 minutes prior to withdrawal of any CEAs in regulating groups during an approach to reactor criticality, and
: b. At least once per 12 hours thereafter.                                    j
* See Special Test Exception 3.10.2
# With K,ff greater than or equal to 1.0.
Amendment Number 9 !
3/4 1-28                    February 27, 1984 l
1
 
b REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION 4
3.1.3.6 TheregulatingCEAgroupsshallbeligtedtothewithdrawalsequence, specified overlap, and to the insertion limits                          shown on Figure 3.1-3**, with CEA insertion between the Long Term Steady State Insertion Limits and the Transient Insertion Limits restricted to:
: a. Less than or equal to 4 hours per 24 hour interval,
: b. Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day interval, and
: c. Less than or equal to 14 Effective Full Power Days per 18 Effective Full Power Month.
APPLICA8ILITJ,: MODES 1* and 2*#.
ACTION:
: a. With the regulating CEA groups inserted beyond the Transient Inser-tion Limits, except for surveillance testing pursuant to Specifica-tion 4.1.3.1.2, within two hours either:
: 1. Restore the regulating CEA groups to within the limits, or
: 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using the above figure.
: b. With the regulating CEA groups inserted between the Long Tenn Steady-State Insertion Limits and the Transient Insertion Limits for intervals greater than 4 hours per 24 hour interval, operation may proceed provided either:
: 1. The Short Tem Steady State Insertion Limits of Figure 3.1-3 are not exceeded, or
* See special Test Exceptions 3.10.2 and 3.10.4.
              #            With K,ff greater than or equal to 1.0.
                        ~
              **          CEA's are considered ~ fully withdrawn in accordance with Figure 3.1-3 when withdrawn to at least 144.75 inches (193 steps).
              ##          Following a reactor power cutback in which (1) Regulating Group 5 is dropped or (2) Regulating Groups 4 and 5 are dropped and for cases (1)
                        .and (2) should the remaining Regulating Groups (Group 1, 2, 3, and 4) be sequentially inserted, the Transient Insertion Limit of Figure 3.1-3 can be exceeded for up to 2 hours. Also for cases (1) and (2), the f~                      specified overlap between Regulating Groups 3, 4 and 5 can be exceeded for up to 2 hours.
Amendment Number 9 3/4 1-29                        ,
February 27, 1984
 
1
: 2. Any subsequent increase in THERMAL POWER is restricted to less than or equal to 5% of RATED THERMAL POWER per hour.
: c. With the regulating CEA group inserted between the Long Tern Steady State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFPD per 30 EFPD interval or greater than 14 EFPD per 18 Effcetive Full Power Months, either:
: 1. Restore the regulating groups to wi. thin the Long Term Steady i
State Insertion Limits within two hours, or
: 2. Be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits at least once per 12 hours except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulat-ed times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at least once per 24 hours.
i 5
  =
Amendment Number 9 3/4 1-30                February 27, 1984
 
l r
SEE APPLICANT'S SAR FIGURE 3.1-3 CEA INSERTION LIMITS VS. THERMAL POWER l
Amendment Number 9 i February 27, 1984  j 3/4 1-31                                  l
 
THIS PAGE INTENTIONALLY BLANK l
 
1 i
3/4.2                          POWER DISTRIBUTION LIMITS 3/4.2.1                          LINEAR HEAT RATE l
l                  LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed 14.0 kw/ft.
APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
With the linear heat rate exceeding its limits, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on kw/ft; or (2) when the COLSS is not being used, any OPERABLE
!                  Local Power Density channel exceeding the linear heat rate limit, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:
: a.              Restore the linear heat. rate to within its limits within one hour, or
: b.              Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.2.1.1                  The provisions of Specification 4.0.4 are not applicable.
4.2.1.2 The linear heat rate shall be determined to be within its limits when THERMAL POWER is above 21% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System
!                  (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours that the linear heat rate, as indicated on all OPERABLE Local Power Density channels, is less than or equal to 14.0 kw/ft.
4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power-operating limit based on 14.0 kw/ft.
[
Amendment Number 9 February 27,1984 3/4 2-1
 
PLANAR
  .h  3/4.2.2      y  RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIAL PEAKING FAC                  ) shall be less than or equaltothePLANARRADIALPEAKINGFACTORS(F@{0            ) use Limit Supervisory System (COLSS) and in the C e Protection Calculators (CPC).
APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.*
ACTION:
With a F*xy exceeding a correspondingcF xy, within 6 hours either:
: a. Adjust the CPC addressable constants to increase the multiplier applied to planar radigl peaking by a factor equivalent to greater than or equal to F'" /F andrestrictsubsequentopeatignsothata margintotheCOLS57upEating 100% is maintained; or              limits of at least [( xy/FxY) - 1.0] x
: b. Adjust the affected PLANAR RADIAL PEAKING FACTORSc (F .) used in the COLSS and CPC to a value greater than or equal to thFmeasured PLANARRADIALPEAKINGFACTORS(Q)or
: c. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.2.1    The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 The measured PLANAR RADIAL PEAKING FACTORS (F*            obtained by using the incore detection system, shall bg determined to be*Te)ss than or equa the PLANAR RADIAL PEAKING FACTORS (Fxy), used in the COLSS and CPC at the following intervals:
: a. After each fuel loading with THERMAL POWER greater than 40% but prior to operation above 70% of RATED THERMAL POWER, and
: b. At least once pcr 31 Effective Full Power Days.
* See Special Test Exception 3.10.2.
Amendment Number 9 February 27, 1984 3/4 2-2
 
l l
3/4.2.3        AZIMUTHAL POWER TILT - T q 1
l LIMITING CONDITION FOR OPERATION                                                              i l
Po M R 3.2.3 The AZIMUTHAL POWER TILT (T          shall not exceed the AZIMUTHAL3TILT O',, ALLOWANCE used in the Core Protectio)n Calculators (CP ).
APPLICABILITY: MODE I above 20% of RATED THERMAL POWER.*
ACTION:                                Jc44%xA8 % M                                    %
6wdAL PesJet act N
: a. With the measured AZIMUTHAL POWER TILT detg              to exceed the      h*
who used in the CPCs, but            .10, either        the power tilt te withi- the v:12: ::d i- the C&G o adjust the AZIMUTHAL TILT ALLOWANCE used in the CPCato a value greater than or equ            to the 3
measured g                                  Aqg                    (tuYR
: b. With the measured AZIMUTHAL POWER TILT determined to exceed 0.10 then:
: 1. Due to misalignment of either a part length or full length CEA, within 30 minutes verify that the Core Operating Limit Super-visory System (COLSS) (when COLSS is being used to monitor the core power distribution per Specifications 4.2.1 and 4.2.4) is detecting the CEA misalignment.
: 2. Verify that the AZIMUTHAL POWER TILT is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and verify that the variable overpower trip setpoint has been reduced as appropriate within the next 4 hours.
: 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER. Subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that the AZIMUTHAL POWER TILT is verified within its limit at least once per hour for'12 hours or until verified acceptable at 95% or greater RATED THERMAL POWER.
SURVEILLANCE RE0tIIREMENTS 4.2.3.1  The provisions of Specification 4.0.4 are not applicable.
4.2.3.2 The AZIMUTHAL POWER TILT shall be detemined to be within its limit above 20% of RATED THERMAL POWER as follows:
* See Special Test Exception 3.10.2.
Amendment Number 9 3/4 2-3
: a. Continuously monitoring the tilt with COLSS when the COLSS is OPERABLE.
: b. Calculating the tilt at least once per 12 hours when the COLSS is inoperable,
: c. Verifying at least once per 31 days, that the COLSS Aximuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT greater than the AZIMUTHAL POWER TILT Allowance used in the CPCs.
EFPD (2[) d. Using the incore detectors at least once per 31 days Cffective Tell Rower-Geys to independently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.
l l
l l
l  -
Amendment Number 9 l                                              3/4 2-4                      February 27, 1984 i
l t..
 
                                                              . _ _                      _________ _ .__                      _ . . = _ . .        _                  -                ._                  _  _    _  _          ___
3/4.2.4        DNBR MARGIN LIMITING CONDITION FOR OPERATION 3.2.4 The DNBR margin shall be maintained by operating within the region of Acceptable Operation of Figure 3.2-1 or 3.2-2, as applicable.
4                                APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.
ACTION:
.: s With operation outside of the region of acceptable operation, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; or (2) when the COLSS is not being used, any OPERABLE Low DNBR channel below the DNBR limit, within 15 minutes initiate corrective action to restore either the DNBR core power operating limit or the DNBR to within the limits and either:
: a. Restore the DNBR core power operating limit or DNBR to within' its limits within one hour, or
: b. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.2.4.1    The provisions of Specification 4.0.4 are not applicable.
:                                4.2.4.2 The DNBR shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS)
.                                p:r ":,_._ :." i or, with t                                                      OLSS out of service, by verifying at least once per 2 hours that the DNBR a                                                      ndicated on all OPERABLE DNBR                                                              channels, is
            @ within the limit shown on, Figure 3.2-2.                                                                                                                                                    3
                                                                                                                                                                                                  -9
!                                4.2.4.3 At least once per 31 days, the COLSS Margin Alam shall be verified l-                              to actuate at a THERMAL POWER level less than or equal to the core power l
operating limit based on DNBR.
          -                      4.2.4.4 The following DNBR or equivalent penalty factors shall be verified to be included in the COLSS and CPC DNBR calcualtions at least once per 31 days:
EFP.D 4
1
      !                                                                                                                                                                                                          Amendment Number 9
;    I                                                                                                                                                                                                          February 27, 1984 l                                                                                                                  3/4 2-5 I
4
 
t GWD Burnup(FTU)                      DNBR Penalty (%)*
0-10                            0.5 10-20                              1.0 20-30                              2.0 30-40                              3.5 40-50                              5.5
* The penalty for each batch will be determined from the batch's maximum burnup assembly and applied to the batch's maximum radial power peak assembly. A single net penalty for COLSS and CPC will be determined from the penalties associated with each batch accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
4 1
1 t
Amendment Number 9 3/4 2-6                                      February 27, 1984
 
SEE APPLICANT'S SAR FIGURE 3.2-1 h DNBR MARGIN OPERATING LIMIT BASED ON COLSS
( cot.ss th) SGRusC6)
Amendment Number 9 3/4 2-7                February 27, 1984 i
 
SEE APPLICANT'S SAR FIGURE 3.2-2 a
DN8R MARGIh OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSSOUTOFSERVICE)
Amendment Number 9 3/4 2-8              February 27, 1984
 
3/4.2.5              RCS FLOW RATE LIMITING CONDITION FOR OPERATION 3.2.5 The actual Reacgor Coolant System total flow rate shall be greater than or equal to 164.0 x 10 lbm/hr.
APPLICABILITY: MODE 1.
ACTION:
With the actual Reactor Coolant System total flow rate determined to be less than the above limit, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined to be greater than its limit at least once per 12 hours.
1 Amendment Number 9
    -s
        ;                                            3/4 2-9                                                      February 27, 1984
    ~.-
 
3/4.2.6        REACTOR COOLANT COLD LEG TEMPERATURE                                                    i LIMITING CONDITION FOR OPERATION 3.2.6 The Reactor Coolant Cold Leg temperature shall be within the area of Acceptable Operation on Figure 3.2-3.
  @    APPLICABILITY: MODE 1        e 30% of Rated Therma                      [
M 3.
ACTION:
With the Reactor Coolant Cold Leg temperature exceeding its l'imit, restore the
                                                                                  'HERM L ? % ", te 1;;;
temperature      to EDwithin  its limit    within the2next    hours? heorr :.
r;i.;.kia M sfMbdy
  @    t      ^%  f T g F0wER    wiu.l..
SURVEILLANCE REQUIREMENTS 4.2.6 The Reactor Coolant Cold Leg temperature shall be determined to be within its limit at least once per 12 hours.
f See Spedal 'GSt Eve.aptiou. 3. to. q l
Amendment Number 9 t
3/4 2-10                            February 27, 1984
 
FIGURE 3.2 3 REACTOR COOLANT COLD LEG TEMPERATURE vs CORE POWER LEVEL 580        i      i      i    ,      ,
515  -
570                                                                  ,
w 570                        568                                        568 o
g                                    AREA OF ACCEPT ABLE OPERATION                          562  .
  $ 560 3
l555_/
  .                                                                552          ,
o 550  -
O 540  .
8      '
0                                              70    80 90      100 10    20    30    40    50    60 CORE POWER LEVEL, % OF RATED THERMAL POWER
 
s 3/4.2.7        AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:
: a. COLSS OPERABLE
                    -0.28 1 ASI 5 + 0.28 b .'  COLSS OUT CF SERVICE (CPC)
                    -0.20 1 ASI $ + 0.20 APPLICABILITY: MODE I above 20% of RATED THERMAL POWER
* ACTION:
q[)  With the core average AXIAL SHAPE INDEX ASI outside its above limits, restore the core average ASI to within its imits within 2 hours or reduce v- THERMAL POWER to less than 20% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.7 The core average AXIAL SHAPE INDEX shall be determined to be within its limit at least once per 12 hours using the COLSS or any OPERABLE Core Protec-tion Calculator channel.
* See Special Test Exception 3.10.2 Amendment Number 9 3/4 2-12      February 27, 1984
 
f
  ~.
3/4.2.8          PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION 3.2.8 The pressurizer pressure shall be maintained between 1815 psia and 2370 l                . psia.
h          APPLICABILITY: MODE 1 a.ed. d #
ACTION:
4                  With the pressurizer pressure outside its above limits, restore the pressure to within its limit within 2 hours or redse T:lE""AL P0"EP t: k;; th= Et c' OC            n                            'th4- the ae/t A Wrn ha fu at J2 sad HM MMb6'l "WATE % 0-sixk T"EP?^.L P^"E"l "1 m r
SURVEILLANCE REQUIREMENTS 4.2.8 The pressurizer pressure shall be determined to be within its limit at least once per 12 hours.
l
    $          See speciat fes1 Excaphu s.to.c
      ,                                                                                    Amendment Number 9 3/4 2-13                February 27, 1984
 
THIS PAGE INTENTIONALLY BLANK l
I
 
s 3/4.3          INSTRUMENTATION                    ^
3/4.3.1              REACTOR PROTECTIVE INSTRUMENTATION 4
LIMITING CONDITION FOR OPERATION t
3.3.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table.3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in
                . Table 3.3-2.
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
i SURVEILLANCE REQUIREMENTS 4.3.1.1  Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the perfonnance of the CHANNEL _ CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at_ the frequencies shown in Table 4.3-1.
4.3.1.2 The logic for the bypasses shall be demonstrated OPERABLE prior to                        l each reactor startup unless performed during the preceding 92 days. The total bypass function shall be demonstrated OPERABLE at least once per 18 months
                ~ during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall-be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one channel per function such that all chan-nels are tested at least once every N times 18 months, where N is the total number of. redundant channels in a specific reactor trip function as shown in                      j the Total No. of Channels column of Table 3.3-1.
4.3.1.4 The isolation characteristics of each CEA isolation amplifier shall be verified at least once per 18 months during the shutdown per the following tests for the CEA position isolation amplifiers:
i*                      a. With 120 volts AC (60 Hz) applied for at least 30 seconds across the output, the reading on the input does not change by more than 0.015 volts DC with an applied input voltage of 5-10 volts DC.
: b. With 120 volts AC (60 Hz) applied for at least 30 seconds across the input, the reading on the output does not exceed 15 volts DC.
4.3.1.5 The Core Protection Calculators shall be determined OPERABLE at least
                ~once per 12 hours by verifying that less than three auto restarts have occur-red on each calculator during the past 12 hours. The auto restarts, Periodic
;                Test Restart (Code 30) and Nonnal System Load (Code 33), shall not be included
:          in this total.
:                                                                          Amendment Number 9 I                                                                          February 27, 1984 l                                                      3/4 3-1
 
4.3.1.6 The Core Protection Calculator System shall be subjected to a CHANNEL
                  ~
FUNCTIONAL TEST to verify OPERABILITY within 12 hours of receipt of a High CPC Cabinet Temperature Alarm.
4 4
Amendment Number 9 3/4 3-2                                                      February 27, 1984
  , _ . . _ _ _ ~ _ .          . _ . . _ _ _ _ . _    . _ - . _ - _ _ _ _            . - . - _ . -            - . - - - - - - - - - - - - - - - - -
 
1 TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION h
bPt MOF
                                                    /                                Total No.
{
Channels E imum        4 Channels / Applicable
                                                                                                                                            \
T RPS Functional Unit                            Of Channels    To Trip  Operable      Modes        Action I. TRIP GENERATION A. Process
: 1. Pressurizer Pressure - High                4              2      3          1, 2          2#, 3#
: 2. Pressurizer Pressure - Low                  4          2(b)        3        1, 2          27, 3#
R                      3. Stcam Generator Level - Low              4/SA          2/SG    3/SG        1, 2          2f, 3#
b                      4. Steam Generator Level - High            4/SG          2/SG    3/SG        1, 2          2#, 3#
b                      b. Steam Generator Pressure - Low          4/SG          2/SG    3/SG    1, 2, 3* , 4*    2#, 3#
: 6. Containment Pressure - High                4              2      3          1, 2          2f, 3#
: 7. Reactor Coolant Flow - Low              4/SG          2/SG    3/SG        1, 2          2f, 3#
: 8. Local Power Deneity - High                4          2 (c)(d)    3          1, 2          2#, 3#
: 9. DNBR - Low                                  4          2(c)(d)    3          1, 2          2#, 3#
y    B. Excore Neutron Flux
{g        1. Variable Overpower                          4              2      3          1, 2          2#, 3#
                                $l m"
: 2. Logarithmic Power Level - High
: a. Startup and Operating                    4          2(a)(d)    3          1, 2          2f, 3#
                                *!g            b. Shutdown                                4              2        3      3*, 4*, 5*      D%      a y,e                                                        4              0      2        3,4,5              4
 
TABLE 3.3-1  (Cont'd)
REACTOR PROTECTIVE INSTRUMENTATION "9                    - - - . - - - . . - - - -
k              d                  Minimum      k              k1 Total No.      Channels            Channels  Applicable RPS Functional Unit                  Of Channels      To Trip              Operable      Modes            Action C. Core Protection Calculator System
: 1. CEA Calculators                          2              1                    2 (e)      1, 2              6, 7
: 2. Core Protection Calcualtors              4          2 (c)(d)                      3    1, 2            2#, 3#, 7 N              D. Supplementary Protection System 5                  1. Pressurizer Pressure - High            4 (f)              2                    ky        1, 2                  8 II. RPS LOGIC A. Matrix Logic                                  6              1                        3    1, 2                  1 6              1                        3  3*, 4*,  5*              8 B. Initiation Logic                              4              2                        4    1, 2,                5 4              2                        4  3*, 4*,  5*              8 T :s$
yg    III RPS ACTUATION DEVICES
    ]R          A. Reactor Trip Breaker                        4 (f)              2                      4      1, 2,                5
    ."E  E 4 (f)              2                      4  3*, 4*, 5*              8 Gl          B. Manual Trip                                4 (f)              2                      4      1, 2                5
,  E ).                                                        4 (f)              2                      4  3*, 4*, 5*            .8 x
1
                                                                                                                                .'        i,jf '
s I
 
                                                                    ..    . . . ,    _ ~ ,. .            ..,.  .,        ,      , , -. - . . .  , , , . . . -
y i        O                            .
w k                                  -
E                                                                1                                  TABLE 3.3-1    (Cont'd)
D        '
TABLE NOTATION mi
                          -                            With the protective system trip breakers in the closed position and the e                                                    ' CEA drive system capable of CEA withdrawal, and fuel in the reactor                                          4
_                                                vessel.                                                                                                    l
                                                #      The provisions of Specification 3.0.4 are not applicable.
k                                                (a) Trip may be manually bypassed above 10'4% of RATED T ERMAL POWER; bypass
                  ,_7.
g                  shall be automatically removed when THERMAL POWER is s 10'4% of RATED THERMAL POWER.
A              g e p fo (b) Trip may be manually bypassed below 400 psi                      bypass shall be automatical-ly removed whenever pressurizer pressure 1 L
                                  @                                                                            cy:    500 naa psia.
M J #D E                                                (c) Trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass N                                    shall be automatically removed when THERMAL POWER IShl% of RATED THERMAL POWER.                                                        ,y w g apaf. It (d) Trip may be bypassed during testing pursuant to Special Test Exception
_                                                  3.10.3.
(e) See Special Test Exception 3.10.2.
            ,                                  (f) There are four channels, each of which is comprised of, or associated l
with, one of the four Reactor Trip breakers, arranged in a selective 2 out of 4 configuration (i.e., one-out-of-two taken twice).
It
  /                                                                                                ACTION STATEMENTS D
g                                                ACTION 1      -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoper-L                                    C                            able channel to OPERABLE status within 48 hours or be in at p                                                                  least systmgT      9 gTgY          rr. within the next 6 hoursg /er ofu h pM,A
_r_                                              ACTION 2      -
With the number of channels OPERABLE one less than the Total
*                      .                                            Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or B                                                                  tripped condition within 1 hour. If the inoperable channel is bypassed, the desirability of maintaining this channel in the y                                                                  bypassed con ti                  hall be reviewed in accordance with Specification .5.1,                . The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN.
w                                                                                                  l 1-                                                                                      (See Ap*cadI0obicd $ ttibiCSN  P
-~                                                                                                                                  Amendment Number 9 6
* February 27, 1984 E                                                                                                      3/4 3-5 m
[                                      **
uh
 
With a channel process measurement circuit that affects multiple functional units inoperable or in test,, bypass or trip all associated functional units as listed below:
Process Measurement Circuit          Functional Unit Bypassed / Tripped i  l'. Excore Nuclear Instrument -      Variable Overpower -
Linear Power                    Local Power Density - High (Subchannel or Linear)          DNBR - Low
: 2. Pressurizer Pressure - High    Pressurizer Pressure - High (Narrow Range)                  Local Power Density - High
            ,                                                DNBR - Low
: 3. Steam Generator Pressure -      Steam Generator Pressure - Low                            l Low SteamGenerator#1 Steam Generator,#2 Level                Level-Low  (ESF{
                                                                                                        - Low (ESF,    l
: 4. Steam Generator Level - Low    Steam Generator Level - Low (RPS)
(Wide Range)                    Steam Generator #1 Level - Low (ESF Steam Generator #2 Level - Low (ESF
: 5. Core Protection Calculator      Local Power Density - High DNBR - Low ACTION 3  -
With the number of channels OPERABLE One less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION
          )            may continue provided the following conditions are satisfied:
: a. Verify that one of the inoperable channels has been bypassed and place the other channel in the tripped condition within 1 hour, and
: b. All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below:
Process Measurement Circuit          Functional Unit Bypassed / Tripped
: 1. Excore Nuclear Instrument      Variable Overpower Linear Power                    Local Power Density - High (Subchannel or Linear)          DNBR - Low 1
: 2. Pressurizer Pressure - High    Pressurizer Pressure - High (Narrow Range)                  Local Power Density - High DNBR - Low
: 3. Steam Generator Pressure -      Steam Generator Pressure - Low Low SteamGenerator#1 Level-Low (ESF)
Steam Generator #2 Level - Low (ESF)
: 4. Steam Generator Level - Low    Steam Generator Level - Low (Wide Range)                    Steam Generator #1 Level - Low (ESF)
        .                                                    Steam Generator #2 Level - Low (ESF)
: 5. Core Protection Calculator      Local Power Density - High DNBR - Low Amendment Number 9 3/4 3-6                  February 27, 1984 9
 
6 STARTUP and/or POWER OPERATION may continue until the perfor-mance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied.
ACTION 4 -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5 -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, STARTUP and/or POWER I              OPERATION may continue provided the reactor trip breaker of the inoperable channel is placed in the tripped condition within 1 hour, otherwise, be in at least HOT STANDBY within 6 hours; however, the reactor trip breaker of the inoperable channel may be closed for up to one hour in order to perform surveillance testing per Specification 4.3.1.1.
ACTION 6 -
: a. With one CEAC inoperable, operation may continue for up to l                    7 days provided that at least once per 4 hours, each CEA is verified to be within 6.6 inches (indicated position) of all other CEAs in its group.
: b. With both CEACs inoperable, operation may continue provided that:                                                                  -
: 1. Within 1 hour the margins required by Specifications 3.2.1 and 3.2.4 are increased and maintained at a value equivalent to greater than or equal to 19% of RATED THERMAL POWER.
: 2. Within 4 hours:
a)  All full-length and part-length CEA groups are withdrawn to and subsequently maintained at the
                                " Full Out" position, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2 or for control when CEA group 5 may be insert.ed no further than 127.5 inches withdrawn, b)  The "RSPT/CEAC Inoperable" addressable constant in the CPCs is set to the inoperable status, c)  The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the " Standby" mode except during CEA group 5 motion permitted by a) above, when i                              the CEDMCS may be operated in either the " Manual l                              Group" or " Manual Individual" mode.
Amendment Number 9 February 27, 1984 3/4 3-7 1
: 3. At least once per 4 hours, all full length and part-length CEAs are verified fully withdrawn except l
during surveillance testing pursuant to Specification                                  ,
4.1.3.1.2 or during insertion of CEA group 5 as permitted by 2.a) above, then verify at least once per 4 hours that the inserted CEAs are aligned within l                                  6.6 inches (indicated position) of all other CEAs in its group.
ACTION 7 -
With three or more auto restarts (excluding auto restart codes 30 and 33) of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by perfonning a CHANNEL FUNCTIONAL TEST within the next 24 hours.
        @  ACTION 8 -
With the number of OPERABLE channels o Channels OPERABLE requirements restore ess than the Minimum 8
inoperable channel to OPERABLE status within 48 hours or operr tB affected reactor trip breake        within the next hour.      &
i l
l l
l i
                                                                                                                        /
i f'
Amendment Number 9 3/4 3-8 I - -  -        -
 
4
                                                                                    ' TABLE 3.3-2 i
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT                                                                            RESPONSE TIME I. TRIP GENERATION A.      Process j
i
: 1.      Pressurizer Pressure - High                                          5,    seconds
: 2.      Pressurizer Pressure - Low                                            <      seconds
: 3.      Steam Generator Level - Low                                          <      seconds
: 4.    . Steam Generator Level - High                                        3,    seconds u,
1                                Steam Generator Pressure - Low                                        <      seconds l              u,                        5.                                                                              ,
d2                                                                                                            seconds
!                                        6.      Containment Pressure - High                                          3,
: 7.  ' Reactor Coolant Flow - Low                                              5,    seconds
:                                        8.      Local Power Density - High***
1
: a. Neutron Flux Power from Excore Neutron Detectors              <      seconds
* i l                                                b. CEA Positions                                                  7      seconds **
: c. CEA Positions: CEAC Penalty Factor                            7      seconds **
2? !I                9.      DNBR - Low ***
i'                  T$                                                                                                  <    seconds
* E jf                        a. Neutron Flux Power from Excore Neutron Detectors 7      seconds **
Qs                          b. CEA Positions 7      secondsif na "                        c. Cold Leg Temperature
                    .'"Ei                        d. Hot Leg Temperature                                            7      secondsif
)
                    -. r                        e.    - Primary Coolant Pump Shaft Speed                              7      seconds #
X$, j'                      f. Reactor Coolant Pressure from Pressurizer                      7      seconds ###
: g. CEA Positions: CEAC Penalty Factor                            i      seconds **
1                      us Sas.O    ff  lica.tT4 fse#t forspel#c rup h.
 
      . _ _ ~. _      . _ . _    _ _ _      _    . _ _  . _          - . . . - . _ _    __    _ _._  . _.. ,..__ _ .      . _ . .      _ _ _ ..
i TABLE 3.3-2        (Cont'd)      .
REACTOR PROTECTIVE INSTRUMENTATIGN RESPONSE TIMES B. Excore Neutron Flux
: 1. Variable Overpower                                                                  <
_ seconds *
: 2. Logarithmic Power Level - High i                                        a.      Startup and Operating                                                      <
seconds
* b.
Shutdown                                                                    5            seconds
* C. Core Protection Calculator System m                          1. CEA Calculators                                                                    Not Applicable s
j    [                          2. Core Protection Calculators                                                        Not Applicable 1
4      o                      D. Supplementary Protection System
: 1. Pressurizer Pressure - High                                                        5            seconds II. RPS LOGIC A. Matrix Logic                                                                              Not Applicable 1
3 B. Initiation Logic                                                                          Not Applicable III. RPS ACTUATION DEVICES i                Ef
                !?s          A. Reactor Trip Breakers                                                                      Not Applicable i                Eg              Manual Trip Q3~          B.                                                                                            Not Applicable
                ~
i              ?j GW
!'              g, e
: 1. Su apGeaut>fmn y sge, ospna Hess
 
  +
TABLE 3.3-2  (Cont'd)
REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES
* Neutron detectors are exempt from repsonse time testing. The response time of the neutron flux signal portion of the channel shall be measured from the detector output or from the input of the first electronic component in the channel.
            **  The CEA position transmitters are exempt from response time testing. The response time shall be measured from the input to the CPC, CEAC or signal isolator.
            *** Response times are verified using CPC Response Time Test Software, and are for hardware delays only.
i    The pulse transmitters measuring pump speed are exempt from response time testing. The response time shall be measured from the pulse shaper input.
g      ##  Response time shall be measured from the output of the resistance temperature detector (sensor). RTD
    =          response time shall be measured at least once per 18 months. The measured response time (P T
                                                                                                              ) of the slowest w            RTD shall be less than or equal to 6.0 seconds.
    ~
            ### Response time shall be measured from the output of the pressure transmitter. The transmitter response time shall be less than or equal to (0.7) seconds.
T E&
4a*
    ~
      . 8-
    =4
 
i TABLE 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL    MODES IN WHICH CHANNEL          CHANNEL          FUNCTIONAL    SURVEILLANCE.
FUNCTIONAL UNIT                                  CHECK        CALIBRATION          TEST          REQUIRED I. TRIP GENERATION A. Process
: 1. Pressurizer Pressure - High        S                R                  M              1, 2
: 2. Pressurizer Pressure - Low          S                R                  M              1, 2
: 3. Steam Generator Level - Low        S                R                  M              1, 2 g                                  4. Steam Generator Level - High        S                R                  M              1, 2
  'r'                                5. Steam Generator Pressure - Low      S                R                  M        1, 2, 3*, 4*
M
: 6. Containment Pressure - High        S                R                  M              1, 2
: 7. Reactor Coolant Flow - Low          S                R                  M              1, 2
: 8. Local Power Density - High          S                                                  1, 2 See Core Protection Calculation System
: 9. DNBR - Low                          S                                                  1, 2 yg          B. Excore Neutron Flux er e j E.              1. Variable Overpower                  S      D (2, 4), M (3, 4)          M              1, 2 Q!e                                                                Q (4) yg:                2. Logarithmic Power Level - High      S              R (4)        M and S/U (1)  1,2,3,4,5,
                  ,g.                                                                                                    and
* TABLE 4.3-1  (Cont'd)
REACTOR PROTECTIVE INSTkUMENTATION'SURVEILLANC2 REQUIREMENTS CHANNEL    MODES IN WHICH CHANNEL          CHANNEL        FUNCTIONAL    SURVEILLANCE FUNCTIONAL UNIT                                    CHECK        CALIBRATION        TEST          REQUIRED C. Core Protection Calculator System
: 1. CEA Calculators                    S                R            M,R(6)            1, 2
: 2. Core Protection Calculators        S      D (2, 4), R (4, 5)  M (9), R (6)        1, 2 M (8), S (7)
D. Supplementary Protection System 5                      1. Pressurizer Pressure - High        S                R                M            1, 2 w
1          II. RPS LOGIC A. Matrix Logic                            N.A.            N.A.              M      1, 2, 3*, 4*, E*
B. Initiation Logic                      N.A.            N.A.              M      1, 2, 3*, 4*, 5*
III. RPS ACTUATION DEVICES A. Reactor Trip breakers                  N.A.            N.A.          M, R (10), 1, 2, 3*, 4*, 5*
B. Manual Trip                            N.A.            N.A.          M, S/U (1)) 1, 2, 3*, 4*, 5*
a" .E E
    *'T E.?r 3i 2 .'
 
TABLE 4.3-1    (Cont'd)
TABLE NOTATION
    *    -    With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawaly ad. p a tko.Mac.ter vtual (1) Each startup or when required with the reactor trip breakers closed and the CEA drive system capable or rod withdrawal, i' not performed in the previous 7 days.                                            g (2) Heatbalanceonly(CHANNEL.FUNCTIONALTESTnotincluded))above15%of RATED THERMAL POWER; adjust the Linear Power Level, CPC & Power and CPC
@        nuclear power signals to agree with calorimetric calculation if absolute difference is > 2%. During PHYSICS TESTS, these daily calibrations may be suspended provided these calibrations are performed upon reaching each major test power plateau and prior to proceeding to the next major test power plateau.
(3) Above 15% of RATED THERMAL POWER, verify that the linear power subchannel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core Protection Calculators.
(4) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5) After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine the shape anneal-ing matrix elements and the Core Protection Calculators shall use these elements.
(6) This CHANNEL FUNCTIONAL TEST shr.ll include the injection of simulated.
process signals into the channel as close to the sensors as practible to verify OPERABILITY including alarm and/or trip functions.
(7) Above 70% of RATED THERMAL POWER, verify that the steady state total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation (conservatively compensated for measurement uncertainties) or by calorimetric calculations (conservatively compensated for measurement uncertainties) and if necessary, adjust the CPC addrestable constant flow coefficients such that each CPC indicated flow is less than or equal to the actual flow rate. The flow measurement uncertainty may be included in the BERR1 term in the CPC and is equal to or greater than 4%.
(8) Above 70% of RATED THERMAL POWER, verify that the steady state total RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation and the ultrasonic flow meter adjusted  pump) uncertainties      curves (conservatively compensated for measurement orbycalorimetriccalculations(conservatively compensated for measurement uncertainties).
Amendment Number 9 February 27, 1984 3/4 3-14
 
(9) The monthly CHANNEL FUNCTIONAL TEST shall inClJde verification that the correct values of addressable constants are installed in each OPERABLE CPC per Specification 2.2.2.
(10) At least once per 18 months and following maintenance or adjustment of the reactor-trip breakers, the CHANNEL FUNCTIONAL TEST shall include independent verification of the undervoltage and shunt trips.
Amendment Niaber 9 February 27, 1984 3/4 3-15 t
i
 
3/4.3.2          ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERABILITY 3.3.2 The Engineered Safety Feature Actuation Sy~ stem (ESFAS) instrumentation channels and bypasses shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.
APPLICABILITY: As shown in Table 3.3-3.
ACTION:
: a. With the ESFAS instrumentation channel-trip setpoint less conserva-tive than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value, b.-  With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.
SURVEILLANCE REQUIREMENTS 4.3.2.1    Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTION-AL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.
4.3.2.2 The logic for the operating and trip channel bypasses shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation. The total operating bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the
" Total No. of Channels" column of Table 3.3-3.
Amendment Number 9 February 27, 1984 3/4 3-16
 
TABLE 3.3-3 ENCIDEEIIED SAFETY FEATURES ACTUATION SYSTEN IIISTRINGITATION NININIM TOTAL NO.          CHAISIELS          CHAISIELS  APPLICABLE ESFA SYSTDI FUNCil0 MAL LSilf                0F OlANIELS        TO TRIP            OPERA 8LE    NDDES        ACTION
: 1. SAFETY IIUECTION (SIAS)
A. Sensor / Trip Units
: 1. Containment Pressure - High                  4                2                  3      1,2,3,4      13*, 14*
: 2. Pressurizer Pressure - Low                  4                2                  3    1, 2, 3(a) 4  13*, 14*
: 8. ESFA System Logic
: 1. Matrix Logic                                6                1                  3      1,2,3,4          17
: 2. Initiation Logic                        4 ICI              2(d)                  4      1,2,3,4          12
: 3. Manual SIAS (Trip Buttons)              4 ICI              2 IdI                4      1, 2 3, 4        12 w
~
m W                C. Automatic Actuation Logic                        2                1                  2    1,2,3,4          16 U
ll. CONTAlteBIT 150LAT1000 (CIAS)
A. Sensor / Trip Units
: 1. Containment Pressure - High                  4                2                  3        1,2,3      13*, 14*
: 2. Pressurizer Pressure - Low                  4                2                  3    1,2,3(a)      13*, 14*
: 8. ESFA System Logic yg                1. Matria Logic                                6                1                  3      1,2,3          17 Q@                2    Initiation Logic                        4I *I              2(d)                  4    1,2,3,4          12 5$                3. Manuel CIAS (Trip Buttons)                4 ICI              2(d)                  4    1,2,3,4          12 E$
y            C. Automatic Actuation Logic                        2                1                  2    1,2,3,4          16
    -5 5.Ito'
 
TAALE 3.3-3  (Cont'd)
ENCIIEEMD SAFETY FEATURES ACTUATION SYSTEM letSTm2ENTAT1000 MINilaat TOTAL MO.          CHAISIELS        CHAteIELS  APPLICA8LE ESFA SYSTEM Fl30CT10NAL lasiT                OF OtAteELS          TO TRIP          OPERASLE    IIllDES    ACTIGId lit. CIBfTAlleEhT $ FRAY (CSAS)
A. Sensor /Trfp Unfts
: 1. Containment P. essure - High-High          4                  2                3      1,2,3      13*,14*
: 8. ESFA System Logic              e
: 1. Metria Logic                                6                  1                3      1,2,3        17
: 2. Initiation Logic                        4I *I              2(d)                4    1,2,3,4        12
: 3. Manual CSAS (Trip Buttons)              4I *I                IdI g                                                                                    2                    4    1,2,3,4        12 8;8          C. Automatic Actuation Logic                        2                  1                2    1,2,3,4        16 IV. MAIN STEAM LINE ISOLAT1004 (MSIS)                                                  .
A. Sensor / Trip Units
: 1. Steam Cenerator Pressure - Low          4/ steam            2/ steam          3/ steam 1, 2, 3(b), 4 13*,14*
generator          generator        generator
: 2. Steam Cenerator Level - High            4/steme            2/ steam          3/ steam  1, 2, 3, 4  13e,14*
generator          generator        generator
: 3. Containment Pressure - High                4                  2                3    1,2,3,4      13*,14*
: 8. ESFA System Logic m                  1. Metrix Logic                                6                  1                3    1,2,3,4        17 cr                  2. Inftfation Logic                        4I *I              2(d)                4    1,2,3,4        12 hh                  3. Manual MSIS (Trip Buttons)              4I *I              2(d)                4    1,2,3,4        12 O$
ro "          C. AutcastIc Actuation Logic                        2                  1                2    1,2,3,4        16
,N la:
u*
 
TABLE 3.3-3  (Cont'd)
ENCINEERED SAFETY FEATUES ACTUATiels SYSTEM IIISTiltmENTATION MINiltm TOTAL NO.          CHAfteELS            CHAISIELS APPLICABLE ESFA SYSTD4 FueCTICIiAL tmlT              OF 04A80ELS          TO Tit lP            OPEltA8tE  NODES    ACTION V.      RECIRCULATION (RAS)
A.          Sensor /Trfp Units
: 1. Refueling Water Storage Tank - Low        4                  2                  3      1,2,3    13*,14*
: 8.          ESFA System Logic
: 1. Matrix Logic                              6                  1                  3    1,2,3,4      17
: 2. Initiation Logic                      4(c)                2(d)                  4    1,2,3,4      12 g                          3. Manual RAS                            4 ICI              2(d)                  4    1,2,3,4      12 6
y              C.          Automatic Actuation Logic                    2                  1                  2    1,2,3,4      16 G
VI. EMEltGENCY FEEDWATER (SC-1)(EFAS-1)
A.          Sensor / Trip Unfts
: 1. Steam Cenerator #1 Level - Low            4                  2                    3      1,2,3    13*,14*
: 2. Steam Generator k Pressure D              4                  2                    3    1, 2,      13*,14*
                                                                      -- 3. in_          1  -"
4( d .~C; t$p. :Diff r ireafid (4-                  2                    s    T m lb)    UQD A
B.          ESFA System Logic
: 1. Matrix Logic                              6                  1                  3      1,2,3      17
: 2. Initiation Logic                      4 ICI              2(d)                  4    1,2,3,4      12
: 3. Manual EFAS                            4(c)                2(d)                  4    1,2,3,4      15 2
C.          Automatic Actuation Logic                    2                  1                  2    1,2,3,4      16
                                            *$ a L
                                            -5 GW E .'a A
 
P I
TABLE 3.3-3  (Cont'd)
ENGINEEltED SAFETY FEATullES ACTUATION SYSTEM INSTRUMENTATION 4
MINIMUN TOTAL MO.          CHANNELS          CHANNELS APPLICA8LE ESFA SYSTEM FUNCTIONAL UNIT                OF CHAIGFLS          TO TRIP            OPERi 1E  MODES    ACTION i
VII. EMERCENCY FEEDWATER (SC-2)(EFAS-2) 1 A. Sensor / Trip Units
: 1. Steam Generator #2 Level - Low              4                    2                3      1,2,3    13*,14*
: 2. Steam Generato M Pressure b                4                    2                3    1,2,3b)    13*,14*
: 3.                    "?"-------'-              4                    4                s    1. 2. 3(b) isw. nu)
::n  ^_._ .;..kif OI((tetebInL
                        $1M4%=                                    o\
: 8. ESFA System Logic
: 1. Matrix Logic                                6                    1                3      1,2,3      17 N                2. Initiation Logic                        4 ICI                2 IdI              4    1,2,3,4      12
: 3. Manual EFAS                              4(c)                2(d)                4    1,2,3,4      15 s
O C. Automatic Actuation Logic                        2                    1                2    1,2,3,4      16 l
Ek V3 4EI C3 2 5
  - er
 
TABLE 3.3-3    (Cont'd)
TABLE' NOTATION (a) In MODES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall _be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed
            .whenever pressurizer pressure is greater than or equal to 500 psia.
      -(b) In MODES 3-6, the value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal.to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(c) Four channels provided, arranged in a selective two-out-of-four configur-ation (i.e., one-out-of-two taken twice).
g                    he-oi+t-of-feut                                                            ,
The propera 2 Of 4 combination.
    @D(# (d)  TheprovisionsofSpecification3.0.3arenotapplicab3
        *-    The provisions of Specification 3.0.4 are not applicable.
ACTION STATEMENTS ACTION 12 -      With the number of OPERABLE channels one less than the Total              '
Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
ACTION 13 -      With the number of channels OPERABLE one less than the Total Number of Channels STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour. If the inoperable channel is                I bypassed, the desirability of maintaining this channel in the                !
bypasse            on shall be reviewed in accordance with Speci-            l ficatio    .5.1.6    The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN.
With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below.
(s-ap e )
Amendment Number 9 February 27, 1984 3/4 3-21 t                                                                                                  l l
 
TABLE 3.3-3  (Cont'd)
M                  YA9tfHGTAMON ACOM STAf6MEUS b                                            j l
Process Measurement Circuit            Functional Unit Bypassed / Tripped
: 1. Steam Generator Pressure -      ' Steam Generator Pressure - Low                      l Low                              Steam Generator Level #1              Low Steam Generator Level #2 - Low
: 2. Steam Generator Level - Low      Steam Generator Level - Low (RPS)
(Wide Range)                    Steam Generator Level #1 - Low (ESF-)
Steam Generator Level #2 - Low (ESF)
ACTION 14 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:
: a. Verify that one of the inoperable channels has been                                  i bypassed and place the other inoperable channel in the tripped condition within 1_ hour.                                                    '
: b. All functional units affected by the bypassed / tripped channel shall- also be placed in the bypassed / tripped condition as listed below:
Process Measurement Circuit            Functional Unit Bypassed / Tripped
: 1. Steam Generator Pressure -      Steam Generator Pressure - Low Low                              Steam Generator Level #1 - Low Steam Generator Level #2 - Low (ESF
: 2. Steam Generator Level - Low      Steam Generator Level - Low (RP (Wide Range)                    Steam Generator Level #1 - Low              SF)
Steam Generator Level #2 - Low              SF)
STARTUP and/or POWER OPERATION may continue until the perfor-mance of the next required CHANNEL FUNCTIONAL TEST. Subsequent                              i STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 14 are satisfied.
ACTION 15 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
ACTION 16 - With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 1 hour for surveillance testing provided the other channel is OPERABLE.
Amendment Number 9 February 27, 1984                        l 3/4 3-22
 
T'                                                                                  \
l i
                                                                                    )
i ACTION 17'- With the number of OPERABLE channels one less than the minimum    l number of channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
Amendment Number 9 3/4 3-23
 
TABLE 3.3-4
                                  -ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ESFA SYSTEM FUNCTIONAL UNIT                                              TRIP SETPOINT            ALLOWABLE VALUES I. SAFETYINJECTION(SIAS)
A. Sensor / Trip Units
: 1. Containment Pressure - High
: 2. Pressurizer Pressure - Low B. ESFA System Logic                                          Not Applicable              Not Applicable C. Actuation Systems                                          Not Applicable              Not Applicable
  $        II. CONTAlleqENT ISOLATION (CIAS)          -
y                A. Sensor / Trip Units
  %                    1. Containment Pressure - High                                    *                          *
: 2. Pressurizer Pressure - Low
: 8. ESFA System Logic                                          Not Applicable              Not Applicable C. Actuation Systems                                          Not Applicable          . Not Applicable III. CONTAll88ENT SPRAY (CSAS)
A. Sensor / Trip Units N                1. Containment Pressure High - High                                *
* Not Applicable 3h          d. ESFA System logic                                          Not Applicable Ek          C. Actuation Systems                                          Not Applicable              Not Applicable b
      -M
      ;;;g
* See Applicant's SAR.
E'.
 
TABLE 3.3-4  (Cont'd)
ENGINEERED SAFETY FEATURES AC111ATION SYSTEM INSTRUENTATION TRIP VALUES ESFA SYSTEM FUNCTIONAL UNIT                                                                  TRIP SETPOINT          ALLOWABLE VALUES IV. MAIN STEAM LINE ISOLATION (MSIS)
A.        Sensor / Trip Units
: 1. Steam Generator Pressure - Low                                      *                        *
: 2. Steam Generator Level - High                                        *                        *
: 3. Containment Pressure - High                                        *                        *
: 8.        ESFA System Logic                                                Not Applicable            Not Applicable y                            C.        Actuation Systems                                                Not Applicable            Not Applicable Y
01          V.              RECIRCULATION (RAS)
A.        Sensor / Trip Units
: 1. Refueling Water Storage Tank - Low                                  *                        *
: 8.        ESFA System Logic                                                Not Applicable            Not Applicable C.        Actuation System                                                Not Applicable            Not Applicable l
VI. EMERGENCY FEEDWATER (SG-1)(EFAS-1)
A.        Sensor / Trip Units N                                    1. Steam Generator il Level - Low                                      *
: 2. Steam Generator APressure (SG #2 > SG #1) - High                                              *
                          $ h"
: 8.        ESFA System Logic                                                Not Applicable            Not Applicable
                          ]%                        C.        Actuation Systems                                                Not Applicable            Not Applicable
                          .]                                                                  .
GK T'=
1 1
 
1 TABLE 3.3-4  (Cont'd)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ESFA SYSTEM FUNCTIONAL UNIT                                          TRIP SETPOINT          ALLOWA8LE VALUES VII. EMERGENCY FEEDWATER (SG-2)(EFAS-2)
A. Sensor / Trip Units
: 1. Steam Generaotr #2 Level - Low                                *                        *
: 2. Steam Generator APressure (SG #1 > SG #2) - High              *
* B. ESFA System Logic                                          Not Applicable            Not Applicable C. Actuation Systems                                          Not Applicable            Not Applicable R.
La N
w.
E&
Q3"
    ~
    .F
    -?
t
 
TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION                                                                                                    RESPONSE TIME IN SECONDS
: 1. Manual
: a. SIAS Safety Injection (ECCS)                                                                                                    Not Applicable Containment Isolation                                                                                                      Not Applicable Containment Purge Valve Isolation                                                                                          Not Applicable
: b. CSAS Containment Spray.                                                                                                          Not Applicable
: c. CIAS Containment Isolation                                                                                                      Not Applicable
: d. MSIS                  -
Main Steam Isolation                                                                                                        Not Applicable
.        e. RAS Containment Sump Recirculation                                                                                              Not Applicable
: f. CCAS 4
C'atainmect Cooling                                                                                                          Not Applicable
: g. EFAS
* Emergency Feedwater Pumps                                                *_.                                                Not Applicable Amendment Number 9          )
February 27, 1984 27
_ _ _ . - - _ - . _ _ . _ . . _ _ - , . ~ , _ . _ _ . _ . - _ _                          . - _ _ _
 
TABLE 3.3-5 (Cont'd)
ENGINEERED SAFETY FEATURES RESPONSE TIMES l
INITIATING SIGNAL AND FUNCTION                          RESPONSE TIME IN SECONDS
: 2. Pressurizer Pressure-Low
: a. Safety Injection (ECCS)                                <    */    **
: b. Containment Isolation                                  <    */    **
: 3. Containment Pressure-High
: a. Safety Injection (ECCS)                              <    */    **
: b. Containment Isolation                                <    */    **
: c. Main Steam Isolation                                <    */    **
: 4. Containment Pressure--High-High
: a. Containment Spray                                    1    */    **
: 5. Steam Generator Pressure-Low
: a. Main Steam Isolation                                1    */    **
: 6. Refueling Water Tank-Low
: a. Containment Sump Recirculation                      1    */    **
: 7. Steam Generator Level-Low
: a. Emergency Feedwater                                  <    */    **
: 8. Steam Generator Water Level-High
: a. Main Steam Isolation                                1    */    **
: 9. Steam Generator AP-High-Coincident With Steam Generator Level Low
: a. Emergency Feedwater Isolation from Ruptured          1    */    **
Steam Generator Response time for Motor-Driven Auxiliary
  @  OTE:
dwater Pumps on all S.I. signal starts              1(60.0)
Amendment Number 9 February 27, 1984 3/4 3-28 l'
 
TABLE 3.3-5  (Cont'd)
ENGINEERED SAFETY FEATURES RESPONSE TIMES TABLE NOTATION                                      .
* Diesel generator starting and sequence loading delays included. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.
    **  Diesel generator starting and sequence loading delays not included.
Offsite power available. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.
    ***  See Applicant's SAR for Response Times.
I i
Amendment Number 9 February 27, 1984 3/4 3-29
 
                                                                                                                              .1 TABLE 4.3-2 ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL        MODES FOR WHICH CHANNEL      CHANNEL    FUNCTIONAL        SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT                          CHECK      CALIBRATION      TEST            IS REQUIRED I. SAFETY INJECTION (SIAS)
A. Sensor / Trip Units
: 1. Containment Pressure - High                S            R              M            1,2,3,4
: 2. Pressurizer Pressrue - Low                S            R              M            1,2,3,4 m            B. ESFA System Logic
: 1. Matrix Logic                            NA            NA              M            1, 2, 4,              2. Initiation Logic                        NA            NA              M            1,2,3,4
: 3. Manual SIAS                              NA            NA              M            1,2,3,4 C. Automatic Actuation Logic                    NA            NA        M(1)    2)        1, 2, 3, 4 SA II. CONTAINMENTISOLATION(CIAS)
A. Sensor / Trip Units
    ,,            1. Containment Pressure - High              S            R              M                1,2,3
: 2. Pressurizer Pressure - Low                S            R              M                1, 2, 3 43 ti$
    -s ani 2 .'
a                                                                                                                    __ _  _s
 
  ~
TABLE 4.3-2    (Cont'd)
ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL        MODES FOR WHICH CHANNEL      CHANNEL    FUNCTIONAL        SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT                                        CHECK    CALIBRATION      TEST            IS REQUIRED II. CONTAINMENTISOLATION(CIAS) (Cont'd)
: 8. ESFA System Logic
: 1. Matrix Logic                              NA            NA            M                1, 2, 3
: 2. Initiation Logic                          NA            NA            N            1, 2, 3, 4
: 3. Manual CIAS                              NA            NA            M            1, 2, 3, 4 R,
t w                    C.        Automatic Actuation Logic                      NA            NA      M(1), $J2)          1, 2, 3, 4 u                                                                                                            A
~
SA III. CONTAINMENT SPRAY (CSAS)
A.        Sensor / Trip Units
: 1. Containment Pressure - High-High            S              R            M              1, 2, 3 B.        ESFA System Logic
: 1. Matrix Logic                              NA            NA            M              1, 2, 3
: 2. Initiation Logic                          NA            NA            M            1, 2, 3, 4 gg                        3. Manual CSAS                              NA            NA            M            1, 2, 3, 4 E .&
[5              C.        Automatic Actuation Logic                      NA            NA      M(1),t W2)          1, 2, 3, 4 uz
    -5                                                                                                    SA GK 2 .'
 
TABLE 4.3-2  (Cont'd)
ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL            M00ES FOR WHICH CHANNEL      CHANNEL      FUNCTIONAL                ' SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT                                                    CHECK      CALIBRATION            TEST                  IS REQUIRED IV. MAIN STEAMLINE ISOLATION (MSIS)
A. Sensor / Trip Units
: 1. Steam Generator Pressure - Low                                        S            R                  M                    1, 2, 3, 4
: 2. Steam Generator Level - High                                          S            R                  M                    1, 2, 3, 4
: 3. Containment Pressure - High                                          S            R                  M                    1, 2, 3, 4 R
[          B. ESFA System Logic h                1. Matrix Logic                                                        NA            NA                  M                    1,2,3,4
: 2. Initiation Logic                                                    NA            NA                  M                    1, 2, 3, 4
: 3. Manual MSIS                                                        NA            NA                  M                    1, 2, 3, 4 C. Automatic Actuation Logic                                                NA            NA        M(1),      2)                1, 2, 3, 4 sA V. RECIRCULATION (RAS)
  ,        A. Sensor / Trip Units Er            1. Refueling Water Storage Tank - Low                                    S            R                  M                      1,2,3 0&
a=                                                                                                                                                            !
m" N 27
  -5 Gi 2 .'
f a
f
                                                                                                                                ~
 
                                                                                    ,,                                r<                    , -;;(. , . '
i; LL j                                                                                                  #
                                                                                                                      ,1<
N,
                                                                                                                            ,t            .. t r.7 TABLE 4.3-2    (Cont'd)          - ,-                  ,              f ENGINEERING SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREE NTS
[,_.
CHANNEL            MODES FOR WHICH CHANNEL          CHANNEL FUNCTIONAL            SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT                                  CHECK      CALIBRATION    TEST                  IS REQUIRED V. RECIRCULATION (RAS) (Cont'd)
B. ESFA System Logic
: 1. Matrix Logic                                    NA              NA          M                    1, 2, 3
: 2. Initiation Logic                                NA              NA          M                  1, 2, 3, 4 NA          M                  1, 2, 3, 4
: 3. Manual RAS                                      NA C. Automatic Actuaticn Logic                            NA              NA    M(1),      2)          1, 2, 3, 4 W                                                                                                    A En                                                                                                  S4 VI. EMERGENCYFEEDWATER(SG-1)(EFAS-1)
A. ~ Sensor / Trip Units M                    1, 2, 3
: 1. Steam Generator #1 Level - Low                      S              R R          M                    1,2,3
: 2. Steam Generator k Pressure h                      S c5                "          "    ~-
3 4Ythrgiitssd=                                                                                        2~7      ~              l B. ESFA System Logic Matrix Logic                                    NA              NA          M                    1, 2, 3 77k            1.
            ,=
NA          M                  1, 2, 3, 4 Eg            2. Initiation Logic                                NA ER            3. Manual AFAS                                      NA              NA          M                  1,2,3,4
            -5                                -
1, 2, 3, 4 "K        C. Automatic Actuation Logic                            MA              NA M(1), X2)
E-                                                                                              sa
?
L  ________.____                                    _.    . . . . . . .
 
qw,'                                                          v
                                                                                                                /                                                                              -
TABLE 4.3-2  (Cont'd)
ENGINEERINGSAFETYFEkTURESACTUATIONSYSTEMINSTRUMENTATIONSURVEILLANCEREQUIREME CHANNEL  MODES FOR WHICH CHANNEL      CHANNEL    FUNCTIONAL    SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT                            CHECK    CALIBRATION        TEST        IS REQUIRED VII. EMERGENCY FEEDWATER (SG-2)(EFAS-2)
A. Sensor / Trip Units
: 1.      Steam Generator #2 Level - Low S            R            M            1, 2, 3
: 2.      Steam GeneratorkPressureh              S            R            M            1, 2, 3 M team-Genera M 1-Pressure - L w              (S -~~ ~ ~ R ' ~~ ~ ~ ~~      M            ~1,' 2, 3 R  u
                                                                                                  # f a*4 6 - fF h3 0;fferenh'at
[                              B. ESFA System Logic
: 1.      Matrix Logic                          NA            NA              M            1, 2, 3
: 2.      Initiation Logic                      NA            NA              M        1, 2, 3, 4
: 3. Manual EFAS                            NA            NA              M        1, 2, 3, 4 t
C. Automatic Actuation Logic                      NA          NA        M(1),E(2)        1, 2, 3, 4 SA N,.
E&
48"
                                                                                ~
NZ S
Gni 2 .'
 
TABLE 4.3-2    (Cont'd)
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE NOTATION l
(1) Testing of automatic actuation logic shall include energization/de-ener-                I gization of each initiation relay and verification of proper operation of each initiation relay.
(2) Testing of the actuation relays shall include the energization/de-energi-zation of each actuation relay and verification of proper operation of each actuation relay.
l s
i
,                                                                              Amendment Number 9 l
February 27, 1984 3/4 3-35
                                                                                                  .l
 
l 3/4.3.3        MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3.3.1        Radiation Monitoring Instrumentation See Applicant's SAR.
l l
Amendment Number 9 February 27, 1984 3/4 3-36
 
INCORE DETECTORS LIMITING CONDITION FOR OPERATION 3.3.3.2 The incore detection system shall be OPERABLE with:
: a. At least 75% of all incore detector locations, and
: b. A minimum of two quadrant symetric incore detector locations per core quadrant.
An OPERABLE incore detector location shall consist of a fuel assembly contain-ing a fixed detector string with a minimum of four OPERABLE rhodium detectors or an OPERABLE movable incore detector capable of mapping the location.
APPLICABILITY: When the incore detection system is used for monitoring:
: a. AZIMUTHAL POWER TILT,
: b. Radial Peaking Factors,
: c. Local Power Density,
: d. DNB Margin.
ACTION:
With the incore detection system inoperable, do not use the system for the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.2 The incore detection system shall be demonstrated OPERABLE:
: a. By performance of a CHANNEL CHECK within 24 hours prior to its use and at least once per 7 days thereafter when required for monitoring the AZIMUTHAL POWER TILT, radial peaking factors, local power
* density or DNB margin:              g        u  ,
gj
: b. At least once per 18 months by perf5TmiadcHf TCHANNEL CALIBRATION      ,
operation which exempts the neutron detectors but includes all electronic components. The neutron detectors shall be calibrated prior to installation in th      eactor core.
h ul $1UAL Amendment Number 9 February 27, 1984 3/4 3-37
 
4 SEISMIC INSTRUMENTATION LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3.3.3 Seismic Instrumentation See Applicant's SAR.
i 4
i Amendment Number 9 February 27, 1984 3/4 3-38
 
METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3.3.4 Meteorological Instrumentation See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 3-39 i
i
 
4 Sys(ertf REMOTE SHUTOOWN MONMOE NG INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.5 The remote shutdown controls and monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with readouts displayed external to the        l control room.
APPLICABILITY: MODES 1, 2, and 3.                                                      ;
t      ACTION:
: a. With the number of OPERABLE remote shutdown monitoring onannels less
,                  than required by Table 3.3-6, either restore the inoperable channel to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.
: b. With one or more remote shutdown system instrumentation control circuits required by Table 3.3-6 inoperable, restore the inoperable circuit to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 12 hours,
: c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.5.1    Each remote shutdown monitoring instrumentation channel shall be demonstrated OPERABLE by perfomance of the CHANNEL CHECK and CHANNEL CALIBRA-TION operations at the frequencies shown in Table 4.3-3.
4.3.3.5.2 Each remote shutdown system instrumentation control circuit shall be demonstrated OPERABLE by verifying its capability to perfom its intended function (s) at least once per 18 months.
l l
Amendment Number 9 3/4 3-40
 
TABLE 3.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION Readout            Heasurement /                                                                        Minimum Channels Instrument                      Location              \ Range      /                                                                          OPERABLE
: 1. Logarithmic Neutron Channel                    RSP*            2x 0-8 -2        %                                                                              1
: 2. Reactor Coolant Hot Leg Temperature              RSP              50 - 750 F                                                                                1/ loop
: 3. Pressurizer Pressure                            RSP            15 - 3 0/ psia                                                                                  1
: 4. Pressurizer Level                                RSP                0-        0%                                                                                  1
: 5. Steam Generator Pressure                        RSP              0 - 1385 sig                                                                          1/ steam generator 0 - 100                                                                            1/ steam generator g    6. Steam Generator Water Level                      RSP y    7. Refueling Water Tank Level                      RSP                  - 100%                                                                                    1 2                                                                          0 - 150 gpm
: 8. Charging Flow                                    RSP                                                                                                              1
: 9. Charging Pressure                                RSP                - 3000 psig                                                                                  1 Control Circuits Piessurizer Heater 2P
    ![ @*  RSP - Remote Shutdown Panel Q="
    ~
N2
    -5 GXi s'.
L                                                                                              _ _ _ - _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _
 
4 TABLE 4.3-3 REMOTE SHUTDCWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL        CHANNEL INSTRUMENT                CHECK      CALIBRATION h 1.          Logarithmic Neutron Channel                                                        M              gR
: 2.          Reactor Coolant Hot Leg Temperature                                                M            R
: 3.          Pressurizer Pressure                                                -
M            R
: 4.          Pressurizer Level                                                                  M            R
: 5.          Steam Generator Pressure                                                            M            R
: 6.          Steam Generator Water Level                                                        M            R      ,
: 7.          Refueling Water Tank Level                                                          M            R
: 8.          Charging Flow                                                                      M            R
!  9.          Charging Pressure                                                                  M            R I
l Amendment Number 9 February 27, 1984 3/4 3-42 i
1
 
POST-ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-7 shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
: a. With the number of OPERABLE accident monitoring channels less than the Required Nucher of Channels shown in Table 3.3-7, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours,
: b. With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-7; either restore the inoperable channel (s).to OPCRABLE status within 43 hours or be in at least HOT SHUTDOWN within the next 12 hours.
l
: c. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS' 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrat-ed OPERABLE by performance of the CHANNEL. CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-4.
i Amendment Number 9 February 27, 1984 3/4 3-43
 
i TABLE 3.3-7
~
POST-ACCIDENT MONITORING INSTRUMENTATION Required              Minimum O'
Number of            Channels
;                        Instrument (Illustrational only)                                Channels            OPERABLE    ACflM
: 1. Containment Pressure                                                        2                    1 a  ca.4
: 2. Reactor Coolant Outlet Temperature - THot (Wide Range)                      2                1/ loop
: 3. Reactor Coolant Inlet Temperature - TCold (Wide Range)                      2                1/ loop
: 4. Pressurizer Pressure - Wide Range                                          2                    1 5    Orac*r cambt Sp fr.nssuu,                                                  1                    1 g    6L    Pressurizer Water Level                                                    2-                    1
    +
7E. Steam Generator Pressure                                          2/ steam generator  1/ steam generator
* Steam Generator Water Level - Wide Range                          2/ steam generator  1/ steam generator
            %1.
: 98. Refueling Water Storage Tank Water Level                                    2                    1 I
lo9. Auxiliary Feedwater Flow. Rate                                              2                    1 If M. Reactor Cooling System Subcooling Margin Monitor                            2                    1
  @      IzM. Safety Valve Position Indicator                                              valve              1/ valve 2                    1 y ht.13 Containment Water Level (Harrow Range)
T :s 2                    1 5 g iaN Containment Water Level (Wide Range) g  {~ g is. cou Edt 6eemp QLW                  h
      . g nr. 6 a hr Lun a.u> aux un a                                                    y                        it
      ;;;g n- %uk Ftax tim %e (feweA Raep)                                                2                        1 P.
Y  t            ~A l    SedGOf$  (L              [.              b            ~W      '  W l$
w-e                    upp y as. w m ukusa % auc+aus,
 
i TABLE 4.3-4' i                                                    POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i
CHANNEL                                                                                    CHANNEL 1                                          Instrument (Illustrational Only)                                    CHECK                                                                                  CALIBRATION
!              1.      Containment Pressure                                                                    M                                                                                          R
: 2.      Reactor Coolant Outlet Temperature - THot (Wide Range)                                  M                                                                                          R j              3.      Reactor Coolant Inlet Temperature - TCold (Wide Range)                                  M                                                                                          R.
: 4.      Pressurizer Pressure - Wide Range                                                        M                                                                                          R S-      flackr contM systm Prossuu,                                                              H                                                                                          R 1.6 Pressurizer Water Level                                                                      M                                                                                          R w      4.7 Steam Generator Pressure                                                                    M                                                                                          R
.      1 i
w      X.? Steam Generator Water Level - Wide Range                                                    M                                                                                          R 8s.9 Refueling Water Storage Tank Water Level                                                    M                                                                                          R R.10 Auxilian i-eedwate.* Ficw Rate                                                              M                                                                                          R 14.11 Reactor Coolant System Subcooling Margin Monitor                                          M                                                                                          R ti.it Safety Valve Position Indicator                                                            M                                                                                          R lh:3 Containment Water Level (Narrow Range)                                                      M                                                                                          R g 1114 Containment Water Level (Wide Range)                                                          M                                                                                          R.
is.                                                                                            H                                                                                          R Cou_ Gst (m4 S [3ay    i6, (2 tatter M ukh twd.                                                                  M                                                                                          A N
          ~
R. huh *, Fba. Mbe youe Q                                                                                                                                                                  R
 
CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3.3.7                            Chlorine Detection Systems See Applicant's SAR.
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  !-                                                                                                                                                                                Amendment Number 9 February 27, 1984 3/4 3-46 i
 
FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION AND SURUILLANCE REQUIREMENTS 3/4.3.3.8            Fire Detection Instrumentation See Applicant's SAR.
d
' ;                                                                                            l l
Amendment Number 9 I February 27, 1984  I 3/4 3-47
 
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1 LOOSE - PART DETECTION INSTRUMENTATION                                        !
LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3.3. M      Loose Part Detection Instrumentation See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 3-48
 
TURBINE OVERSPEED PROTECTION LIMITING' CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.3.4          Turbine Overspeed Protection See Applicant's SAR.
Amendment Number 9
.                                                                                                                                              February 27, 1984 3/4 3-49 f
 
THIS PAGE INTENTIONALLY BLANK
 
                            .3/4.4          REACTOR COOLANT SYSTEM 3/4.4.1        REACTOR COOLANT LOOPS AND COOLANT CIRCULATION
                            .STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION
          @                    3.4.1.1  Both loop shall bein$operation.
actor l$polant loops and both bactor (polant pumps in each APPLICABILITY: MODES 1 and 2*.
ACTION:
With less than the above required (eactor holant pumps in operation, be in at h                    least HOT STANDBY within 1 hour.
SURVEILLANCE REQUIREMENTS                                                                  .
4.4.1.1 The above required Reactor Coolant loops shall be verified to be in operation and circulating Reactor Coolant at least on:e per 12 hours.
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* See Special Test Exception 3.10.3.
Amendment Number 9 February 27, 1984
 
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HOT STANDBY LIMITING CONDITION FOR OPERATION
  @ 3.4.1.2 The hactor bolant loops listed below shall be OPERABLE and at least one of these (eactor joolant Loops shall be in operation.A 1                                                      l
: a. Reactor Coolant Loop {A)-and its associated steam generator and at
  @                    least one associated Reactor Coolant pump.
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Q-                                                      i
    @              b. Reactor Coolant Loop {&} and its associated steam generator and at                    l least one associated Reactor Coolant pump.
l APPLICABILITY: MODE 3 ACTION:
: a. With less than the above required $ actor Qolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
: b. With no kactor (oolant loop in operation, suspend all operations
  @                      involving a reduction in boron concentration of the Reactor Coolant System and imediately initiate corrective action to return the required (eactor 0polant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required %eactor bolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
: 4. 4.1. 2. 2 At least one hactor Qolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
4.4.1.2.3 The required steam generator (s) shall be detemined OPERABLE                          '
verifying the secondary side water level to be > 25% on the wide range level                      i i;;dinter at least once per 12 hours.
1
* All Reactor Coolant pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is                    ;
maintained at least 10*F below saturation temperature.                                    '
Amendment Number 9 February 27, 1984                ,
3/4 4-2
 
i HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1. 3 At least two of the loop (s)/ train (s) listed below shall be OPERABLE and at least one Reactor Coolant and/or shutdown cooling loop shall be in operation.*
1                                                            :
h      a. Reactor Coolant Loop fA-and its associated steam generator and at leastoneassociated{eactor$olantpump,**
: b. Reactor Coolant Loop      and its associated steam generator and at least one associated eactorqpolantpump,**
: c. Shutdown Cooling Train #1,
: d. Shutdown Cooling Train #2.
h  APPLICABILITY: MODE 4        1d
* 350*F ACTION:
: a. With less than the above required %eactor holant and/or shutdown cooling loops OPERABLE, immediatelf initiate corrective action to return the required loops to OP.ERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 24 hours.
: b. With no hactor holant or shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
SURVEILLANCE REQUIREMENTS
: 4. 4.1. 3.1 The required hactor holant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker align-ments and indicated power availability.
All Igeactor plant pumps and shutdown cooling pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F below saturation
    @        temperature.
      **    Ahactorholantpumpshallnotbestartedwithoneormoreofthe Reactor    Coolant
(***)*F during      System cooldown        cold **)'F or (b)    leg (*  temperatures during            lessthe heatup unless    than    or equal to (a) secondary water temperature of each steam generator is less than 140*F above each
    @      of the Reactor Coolant System cold leg temperatures.          Ipo*F l
      ***  See Applicant's SAR.                                          mnendment Number 9 February 27, 1984 3/4 4-3 p
 
4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying the secondary side water level to be > 255 er.mdthe
                                                              -                              apdide range level i-dicate at least once per 12 hours.
: 4. 4.1. 3. 3 At least one keactor Qolant or shutdown cooling loop shall be verified to be in operatfon and circulating Reactor Coolant at a flowrate i
greater than or equal to 4000 gpm at least once per 12 hours.
j Amendment Number 9 February 27, 1984
 
                            ~                                              .                .          . - -                        ..      . .          .. -    _
j                    COLD SHUTDOWN - LOOPS FILLED l                    LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one shutdown cooling loop shall be OPERABLE and in opera-
:                    tion *, and either:
: a.          One additional shutdown cooling loop shall be OPERABLE #, or
                  @              b.          The secondary side water level of at least two steam generators shall be greater than 25% p g de range level 'ad ;;ter.
1 APPLICABILITY: MODE 5 with Reactor Coolant loops filledif.
I ACTION:
,                                  a.          With less than the above required loops OPERABLE or with less than
:                                              the required steam generator level, imediately initiate corrective action to return the required loops to OPERABLE status or to restore
:                                              the required level as soon as possible,
: b.            With no shutdown cooling loop in operation, suspend all operations 4
involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be deterinined to be within limits at least once per 12 hours.
4.4.1.4.1.2: At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flowrate greater than or equal to 4000 gpm at least once per 12 hours.
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                    #            One shutdown cooling loop may be inoperable for up to 2 hours for surveil-                                                            l
)                                lance testing provided the other shutdown cooling loop is OPERABLE and in                                                            j operation.                                                                                                                            j i
                    ##            A reactor p lant pump shall not be started with one or more of the                                                                  H Reactor Coolant System cold leg temperatures less than or equal to (**)*F g                during cooldown or (**)*F during heatup unless the secondary water temperature (saturation temperature corresponding to the steam generator pressure) of each steam gener6 tor is less than tWF above each of the Reactor Coolant System cold leg temperatures. 100
* The shutdown cooling pump may be de-energized for up to I hour provided
: 1) no operation:: are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.
                    **            See Applicant's SAR.
9 February 27, 1984 3/4 4-5
 
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COLD SHUTDOWN - LOOPS NOT FILLED t
l LIMITING CONDITION FOR OPERATION                                                            !
3.4.1.4.2 lhe shutdown cooling loops shall be OPERABLE # and at least one shutdown cooling loop shall be in operation.*
APPLICABILITY: MODE 5 with reactor coolant loops not filled.                        ,
ACTION:
: a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.                                                    '
: b. With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation.
SURVEILLANCE REQUIREMENTS f
4.4.1.4 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flowrate greater than or equal to 4000 gpm at least once per 12 hours.
1
                    #    One shutdown cooling loop may be inoperable for up to 2 hours for surveil-lance testing provided the other shutdown cooling loop is OPERABLE and in operation.
* The shutdown cooling pump may be de-energized for up to 1 hour provided
: 1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.
1 Amendment Number 9 February 27, 1984 3/4 4-6
 
l I
l 3/4.4.2                SAFETY VALVES
            -SHUTDOWN.
LIMITING CONDITION FOR OPERATION 3.4.2.1    A minimum of one pressurizer code safety valve shall be OPERABLE with a lift setting of 2500 psia t 1%.*                                                                                      ;
i APPLICABILITY:                MODE 4[when the temperature of all of the RCS cold legs is
        .b cgreater Inan                  **          "F during cooldown or ** 'F during heatup.                                      l ACTION:
: a. With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.
: b. The provisions of Specification 3.0.4 may be suspended for up to 12 hours for entry into and during operation in M00E 4 for purposes of setting the pressurizer code safety valves under ambient (hot) conditions provided a preliminary cold setting was made prior to heatup.
SURVEILLANCE REQUIREMENTS J
4.4.2.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
I
* The lift setting pressure shall correspond to ambient conditions of the                                          )
valve at nominal operating temperature and pressure.
I See Applicant k                                                Amendment Number 9 February 27, 1984          j 3/4 4-7
 
OPERATING LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 psia + 1%.*
APPLICABILITY: MODES'1, 2, 3 ACTION:
: a. With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERA %LE status within 15 minutes or be in
          'at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours with the shutdown cooling system suction relief valves aligned to provide overpressure protection for the Reactor Coolant System.
SURVEILLANCE REQUIREMENTS 4.4.2.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.
The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
Amendment Number 9 3/4 4-8
 
1 i
l 3/4.4.3          PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.3*I, The pressurizer shall be OPERABLE with:
: a. A steady state water volume less than or equal to 58% indicated level ft.),an ( qdcu. ft) but greater than 27% indicated level (4450cu.
ci%
: b. At least two groups of pressurizer heaters capable of being powered from 1E buses each having a nominal capacity of at least 150 kw.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
: a. With one group of the above required pressurizer heaters operable, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.                                                      g h    b.
Wh ""QwW        ?
With the pressurizer otherwise inoperab e, be in at least Hui              -
sr STANDBY with the reactor trip breakers open within 6 hours and in                      '
HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.3.1    The pressurizer water volume shall be determined to be within its limit at least once per 12 hours.
4.4.3.2 The capacity of each of the above required groups of pressurizer                            i heaters shall be verified to be at least 150 kw at least once per 92 days.
4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power:
: a. the pressurizer heaters are automatically shed from the emergency power sources, and
: b. the pressurizer heaters can be reconnected to their respective buses manually from the control room.
I 3/4 4-9                      Amendment Number 9            ,
February 27, 1984
 
AUXILIARY SPRAY LIMITING CONDITION FOR OPERATION 3.4.3.2 Both auxiliary spray valves shall be OPERA 8LE.
APPLICA8ILITY: MODES 1, 2, 3, and 4.
ACTION:
: a. With only one of the above required auxiliary spray valves OPERABLE, I
restore both valves to OPERA 8LE status within 72 hours or be in at least HOT STAN08Y within the next 6 hours and in HOT SHUTDOWN within I              the following 6 hours,
: b. With none of the above required auxiliary spray valves OPERA 8LE, restore at least one valve to OPERA 8LE status within the next 6 hours or be in at least HOT STAN08Y within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The auxiliary spray valves shall be verified to have power available to each valve every 24 hours.                      .
4.4.3.2.2 The auxiliary spray valves shall be cycled at least once per 18 months.
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3/4.4.4        STEAM GENERATORS LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS i
3/4.4.4        Steam Generators                                                i See Applicant's SAR.
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k Amendment Number 9 February 27, 1984 3/4 4-10 l
                                                                                        \
 
k 3/4.4.5        REACTOR COOLANT SYSTEM LEAXAGE                                  I LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.4.5.1        Leakage Detection Systems See Applicant's SAR I
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i Amendment Number 9 February 27, 1984 3/4 4 J
 
s.,
OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION
#                3.4.5.2 Reactor Coolant System leakage shall be limited to:
l l                      a. No PRESSURE BOUNDARY LEAKAGE, b        b. 1 % UNIDENTIFIED LEAKAGE,
          , @        c. 1 g total primary-to-secondary leakage through steam generators,
            @        d. 10 Q4 IDENTIFIED LEAKAGE from the Reactor Coolant System.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
: a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY i
within 6 hours and in COLD SHUTDOWN within the following 30 hours.
: b. With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE, reduce the leakage rate to within its limit within 4 hours or be in at least t
HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.5.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
: a. Monitoring the containment atmosphere gaseous or particulate radio-activity at least once per 12 hours.
: b. Monitoring the containment sump inventory and discharge at least once per 12 hours.
: c. Performance of a Reactor Coolant System water inventory balance at leas                                          nete g e e m i, eacc 6 t.er, h
o,~.{..n3cncepe{,72}ogsjppge
                                          .u sue anu6uu-o    wv....,  -___.
Amendment Number 9 February 27, 1984 3/4 4-12
  , , , ,,    ,    ,    ,                          ,, ,        ,                    ,        ,        , , , , ,    i
 
I 3/4.4.6          CHEMISTRY                                                                            ,
l LIMITING CONDITION FOR OPERATION 3.4.6 The Reactor Coolant System chemistry shall be maintained within the
          . limits specified in Table 3.4-1.
APPLICABILITY: At all times.
ACTION:
          . MODES 1, 2, 3, and 4:
: a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Steady State Limit within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
1 L~              b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUT-DOWN within the following 30 hours.
At All Other Times:
With 'the concentration of either chloride or fluoride in the Reactor Coolant System in excess of its Steady State Limit for more than 24 hours or in excess of its Transient Limit, reduce the pressurizer pressure to
      ,          less than or equal to 500 psia, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the                        1
,                Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pres- sure above 500 psia or prior to proceeding to MODE 4.
SURVEILLANCE REQUIREMENTS 4.4.6 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-1.
Amendment Number 9 February 27, 1984 3/4 4-13 1
 
TABLE 3.4-1 REACTOR COOLANT SYSTEM CHEMISTRY Steady State                                    Transient Parameter                              Limit                                                  Limit Dissolved Oxygen
* 1 0.10 ppm                                < 1.00 ppm Chloride                              < 0.15 ppm                                i 1.50 ppm Fluoride                              1 0.10 ppm                                i 1.00 ppm
* Limit not applicable with T cold less than or equal to 250*F.
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Amendment Number 9 February 27, 1984 3/4 4-14
 
TABLE 4.4-1 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS l
Sample and Analysis Frequency
                                                                                                  )j Parameter l
Dissolved Oxygen
* At least once per 72 hours Chloride                                              At least once per.72 hours Fluoride                                              At least once per 72 hours Not required with Tcold ess than or aqual to 250*F.
Amendment Number 9 February 27, 1984 3/4 4-15
 
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l 3/4.4.7'        SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION
        -3.4.7 The specific activity of the primary coolant shall be limited to:
: a. Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and
: b. Less than or equal to 100/E microcuries/ gram.
APPLICABILITY: MODES 1, 2, 3, 4 and 5.
ACTION:
l        MODES 1, 2 and 3*:
1
: a. With the specific activity of the primary coolant greater than 1.0 l                  microcurie / gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation
!.                  may continue for up to 48 hours provided that the cumulative operat-ing time under these circumstances does not exceed 800 hours in any i
consecutive 12 month period. With the total cumulative operating time at a primary coolant specific activity greater than 1.0 micro-curie / gram DOSE EQUIVALENT I-131 exceeding 500 hours in any consec-utive 6 month period, prepare and submit a Special Report to the Comission pursuant to Specification 6.9.2 within 30 days indicating the number of hours above this limit. The provision of Specifica-tion 3.0.4 are not applicable.
: b. With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT SHIEGOWN with T O          within 6 hours.                        sfMD69      cold less than 500*F
: c. With the specific activity of the primary coolant greater than 100/E microcuries/ gram, be in at least HOT Stitif00WN with T O          500*F within 6 hours.                      STM8'l      cold less than          <
4 MODES 1, 2, 3, 4 and 5:
: d. With the specific activity of the primary coolant greater than 1.0
;                  microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/E micro-curies / gram, perfom the sampling and analysis requirements of item i                  - 4 ~a) of Table 4.4-2 until the specific activity of the primary            Sp M coolant is restored to within its limits. A REPORTA8tttCCt)RRENGE g
',        @        shall be prepared and submitted to the Comission pursuant to Specification 6. .      This report shall contain the results of the specific activity analyses together with the following information:
: 1. Reactor power history starting 48 hours prior to the first                I sample in which the limit was exceeded,                                    ;
With Tcold greater than or equal to 500*F.
3/4 4-16                  Amendment Number 9 February 27, 1934          ,
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: 2.      Fuel burnup by core region,
: 3.      Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded,
: 4.      History of degassing operation, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and
: 5.      The time duration when the specific activity of the primary coolant exceeded 1.0 microcurie / gram DOSE EQUIVALENT I-131.
SURVEILLANCE REQUIREMENTS
          - 4.4.7 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and aaalysis program of Table
          - 4.4-2.
t Amendment Number 9 February 27, 1984 3/4 4-17
 
h.
TABLE 4.4-2 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 4
Type of Measuremer.t                            Sample and Analysis                          Modes in Which Sample And Analysis                                        Frequency                              And Analysis Required
,                                                      1. Gross Activity Determination                    At least once per 72 hours                              1, 2, 3, 4
: 2. Isotopic Analysis for D0sE                      1 per 14 days                                                  1 EQUIVALENT I-131 Concentration i
I                                                      3. Radiochemical for E Detemination                1 per 6 months
* 1 4
j                                                      4. Isotopic Analysis for Iodine                    a)    Once per 4 hours, whenever the                li,2#,3#,4#,5#
:                                                            Including I-131,1-133, and I-135                      specific activity exceeds 1.0 1                                                                                                                  pCi/ gram, DOSE EQUIVALENT I-131 or 100/E pCi/ gram, and R.
s
                                            ?                                                              b)    One sample between 2 and 6 hours                      1, 2, 3 l
2;;                                                                  following a THERMAL POWER change l
1 exceeding 15 percent of the
:                                                                                                                  RATED h ur period THERMAL  &esePOWER. within a one , 4 i                                                                                                                                              Jo_ '2 sufh cu D
!                                                        #  Until the specific activity of the primary coolant system is restored within its limits.
3*          Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor l                                                gg          was last subcritical for 48 hours or longer.
Ea 1                                                a8 i                                                m"
                                                ??s                                                                                                                                  t 4
!                                                2'*
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E 52        i  i      i                i    i                i                i                      i G
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                                              - 250  -                                                                                                              -
E a
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                                              $                                          UNACCEPTABLE S 200 O
E G
w
                                              $ 150  -                                                                                                              -
5 O
o ACCEPTABLE m
                                              @ 100  -                                                                                                              -
EE cL m
                                              .T'
                                              '.-i E    50 -                                                                                                              ' -
w
                                              ?
8 ui W    0 8        20  30      40            50      60                70              80                      90              100 PERCENT OF RATED THERMAL POWER Araendment Number 9 February 27, 1984 DOSE EQUlVALENT l 131 PRIMARY COOLAfJT SPECIFIC C-E                    ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL                                                                Rgure POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY                                                              3.4 1 8                                                                >1.0gCi/ GRAM DOSE EQUIVALENT l 131 3/4 419
 
3/4.4.8            PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8.1      The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:
: a.      A maximum heatup rate of 100*F/hr.
: b.      A maximum cooldown rate of 100*F/hr.
: c.      A maximum temperature change of 10 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
h    APPLICABILITY: At all times.*
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to stmefuAal h determine the effects of the out-of-limit condition on the fr;;tur; t:;;'.c.;;s7Pf c4j      i pr;p;cti e of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next 6 hours and reduce the RCS T            and pressure to less than 210*F and 500 psia, respectively, within the fofiding 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-3. The results of these examinations shall be used to update Figure 3.4-2.
$# % Syelot, b5t btL(NW                        b'IO Amendment Number 9        ,
February 27, 1984 3/4 4-20
 
7                                                                        _
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l See Applicant's SAR 9
FIGURE 3.4-2 REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITATIONS 1
1 Amendment Number 9 February 27, 1984 3/4 4-21
                                                                          )
 
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At ;,,a;..;                                                      o - ::1
  -@                                  Location                              Lead                            Time                  1 Capsule No.                      (de;=:)                              Factor                        (EFPY) 1                                      38*                        1.5                      Standby 2                                      43*                          1.5                      Standby 3                                      137*                        1.5                            4-5 4                                      142                        1.5                      Standby 5                                      230                        1.5                          12-15 6                                      310
* 1.5                          18-24 1
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Amendment Number 9 February 27, 1984
 
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PRESSURIZER H6MuP\ f.coLOcWN LlHlf5 LIMITING CONDITION FOR OPERATION 3.4.8.2 The pressurizer temperature shall be limited to:
: a. A maximum heatup rate of 200*F/hr.,                                    l
: b. A maximum cooldown rate of 200*F/hr.
APPLICABILITY: At all times.
ACTION:
With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition
      @  on the fr +m tap.,es; pap;rties of the pressurizer; detemine that the pressurizer emains acceptable for continued operation or be in at least HOT STAND 8Y with    the next 6 hours and reduce the pressurizer pressure to less than 500 psig ithin the following 30 hours.
SfruetuA0l bQNY SURVEILLANCE REQUIREMENTS 4.4.8.2.1 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.
4.4.8.2.2 The spray water temperature differential shall be detemined for use in Table 5.7-2 for each cycle of main spray with less than 4 reactor coolant pumps operating and for each cycle of auxiliary spray operation.
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Amendment Number 9 February 27, 1984 i
 
OVERPRESSURE PROTECTION SYSTEMS                          b bM A                                    .
TING CONDITION FOR OPERATION                                                          ,
(      3.4.8. Both Shutdown Cooling System suction line relief valves with lift settin of less than or equal to ** psig shall be OPERABLE and aligned t provide verpressure protection for the Reactor Coolant System.
1 APPLICABILI Y: MODES 4, 5, and 6*.        When the temperature of one or          re of            I he RCS cold legs is less than or equal to a    (**)*Fduringcooldown                                                          !
b)    **) *F during heatup ACTION:
i a)  With one or more Shutdown Cooling System suct n line relief valves inoperable restore the valves to OPERABLE status within 1 hour.
Otherwise cooldow4 and depressurize. Do not start a Reactor Coolant Pump if the steam generator secondary watfr temperature (saturation temperature correspogding to the S/G pressure) is greater than the RCS cold leg temperature. Do not start any other pump which could overpressurize the RCS        A bubble she'll be maintained in the pressur-
,                  izer.
        }    b)  In the event the SCS suctken line relief valves actuate to mitigate a RCS pressure transient, a\Special Report shall be prepared and f          submitted to the Comissionjursuant to Specification 6.9.2 within i
30 days. The report shallfdescribe the circumstances initiating the transient, the effect of,the Sci suction line relief valves on the transient and any corrective action necessary to prevent recurrence.
c)  The provisions of Spec fication 3.          4, are not applicable.
                                                                  'N SURVEILLANCE REQUIREMENTS                                  \
4.4.8.3.1 Each SCS s        ion line relief valve shall be verified to be aligned f to provide overpress e protection for the RCS once every 8 hours during:
a)  heatup ith the RCS temperature less than or equal t (**)*F.
b)  coo own with the RCS temperature less than or equal to (**) *F.
,        4.4.8.3.2 he SCS suction line relief valves shall be verified OPE                  LE with the requ ed setpoint at least once per 18 months.
1 This Specification is not applicable with the Reactor Vessel head remo ed.
          **  See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 4-24
 
1 LIMITING CONDITION FOR OPERATION 3.4.8.3 Both"shutdownhoolingSystem                                                    k
    /
lift settings of less than or equal tD(JCS) MR psigsuction  line relief and shall be OPERABLE    valves  with aligned i          to provide overpressure protection for the Reactor Coolant System.
i          APPLICA8ILITY': When the reactor vessel head is installed and the temperature i          of one or more of the RCS cold legs is less than or equal to:
(fr)                        .
I                  a. 266*F during cooldown i
    !                  b.  (*
295*')F during heatup ACTION:
: a. With one SCS relief valve inoperable, restore the inoperable valve to OPERA 8LE status within seven days or reduce Tcold to less than 200*F and, depressurize and vent the RCS through a greater than or equal to
{                  (F4 14 square inch vent (s) within the next eight hours. Do not start a reactor coolant pump if the steam generator secondary water temperature is greater than 100*F above any RCS cold leg temperature.
: b. With both SCS relief valves inoperable, reduce Tcold to less than 200*F i                    and, depressurize and vent the RCS through a greater than or equal to (kk)16 square inch vent (s) within eight hours. Do not start a rea'ctor coolant pump if the steam generator secondary water temperature is greater than 100'F above any RCS cold leg temperature,
: c. In the event either the SCS suction line relief valves or an RCS vent (s) are used to mitigate an RCS pressure transient, a Special s                  Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the SCS suction line relief valves or RCS vent (s) on the transient and any corrective action necessary to prevent recurrence.
: d. The provisions of Specification 3.0.4 are not applicable.
t i          l SURVEILLANCE REQUIREMENTS 4.4.8.3.1 Each SCS suction line relief valve shall be verified to be aligned to provide overpressure protection for the RCS once every 8 hours during l                                                                              (W) j              a. Cooldown with the RCS temperature less than or equal to 2 E*F.
(t #)
: b. Heatup with the RCS temperature less than or equal to 295'F.
i
        }        4.4.8.3.2 The SCS suction line relief valves shall be verified OPERABLE with the required setpoint at least once per 18 months.                                          l s
            \
x'~ . _ _ _ _
 
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i 3/4.4.9                            STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.9 The structural integrity of ASME Code Class 1, 2 and 3 componeats shall                                                              I be maintained in accordance with Specification 4.4.9.
APPLICABILITY: ALL MODES ACTION:
d a)            With the structural integrity of any ASME Code Class 1 component (s) not confonning to the above requirements, restore the structural i
integrity of the affected component (s) to within its limit or j'                                              isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50'F above the minimum tempera-
  ~
j                                                ture required by NDT considerations.
b)            With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor
,                                                Coolant System temperature above 210'F.
c)            With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate the affected component from service.
d)            The provisions of Specification 3.0.4 are not applicable.
3 SURVEILLANCE REQUIREMENTS
;                          4.4.9 In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14 of October 27,1971 (originally issued as Safety Guide 14 on October 27,1971).
i r
h l
Amendment Number 9 February 27, 1984 3/4 4-25 r .w.%,-  ,<r...m%,e,  .r.- y,. ,,..p,...w...,.,,mm._,-.~,,__,w,,            , , . _ - -,%.%-, , , ,_,,w-_, ,_.%,,,._,,, . , _ , , _ , _ _ , , - - _ - _ , - -
 
REACTOR COOLANT SYSTEM 3/4.4.10 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3 4.10 'Both reactor coolant system vent paths from the reactor vessel head shall be OPERABLE and closed.
APPLICABILITY : -Modes 1. 2. 3. and 4 ACTION :
: a. With none of the above required reactor vessel head vent paths OPERABLE. restore at least one path to OPERABLE status within 6 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
: b. With only one of the above, required reactor vessel. head vent paths OPERABLE restore both paths to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.10 Each Reactor. Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:
: a. Verifying all manual isolation valves in each vent path are locked in the open position.
: b. Cycling each. vent through at least one complete cycle from the control room.
: c. Verifying flow through the reactor coolant system vent paths during venting.
 
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e THIS PAGE INTENTIONALLY BLANK.
 
3/4.5            EMERGENCY CORE COOLING SYSTEMS (ECCS)
SAFETY INJECTION TANKS LIMITING CONDITION FOR OPERATION 3.5.1    Each Safety Injection System safety injection tank shall be OPERABLE with:
to Ike nivt (C) a.        The isolation valve keylocked open and power3 removed.G4 ge. Et A co . ned borated water level of between 28%(1802                ) and 72%
        @ b.          1914 t ) on narrow range indication.
: c.      A boron concentration of between 4000 and 4400 ppm of boron.
: d.      A nitrogen cover pressure of between 600 and 625 psig.
: e.      The nitrogen vent valves closed and power removed.**                            *
: f.      The nitrogen vent valves capable of being operated upon restoration of power.
APPLICABILITY:        MODESh,2*,3*+,and4*+.
ACTION:
: a. With one safety injection tank inoperable, except as a result of a closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.                    ikut
        +      With pressurizer pressure > 1750 psia. When pressurizer pressure i's mF below 1750 psia, at least three safety injection tanks must be operable                    '
with a minimum pressure3 of 254 psig and a contained barated water volume m            of between 60%(1415 ft ) level on wide range indication and 72%(1914 Vi)          f t3) level on narrow range indication. With pressurizer pressure below                      ,
1750 psia and four safety injection tanks operable, a minimum pr$ssure of; 254 levelpsig    and range on wide  a contained indication  borated    water volumelevel and 72%(1914ft            of)between on narrow(962 rangeft ) 39%
Qndicationisallowable.                    __            _____ _ ___              _
* SeeSpecialTestException3.10.(.I l
N, vent valves may be cycled as necessary to maintain the required N2 cover pressure per 3.5.1.d.
@    c'***                                      ressurizer pressure less than 430 psia the In M00f
      \M afety,J          4 operet M ection_      h tfomu'ay beTlosed valves                      %
                                          . ~ _ .
e Amendment Number 9
.                                                                                      February 27, 1984 3/4 5-1
 
                                                                          . --n          _m,
.; ; u ., w                  - .,          .
            '                                                                                T 1758                    pro I With pressurizer pressure          ter than or equal to @ psia. When pressur-I    izer pressure is less than          psia, at least three safety injection tanks    i fmustbeOPERA8LE,eachwithaminimumpressureof254psigandamaximum i  pressure of 625 psig, and a contained borated water volume of between 05 narrow range (corresponding to 605 wide range indication or 1415 cubic feet) and 725 narrow range indication (corresponding to 81% wide range indication or 1914 cubic feet). With all four safety injection tanks OPERA 8LE, each tank shall have a minimum pressure of 254 psig and a maximum pressure of 625 psig, and a contained borated water volume of between 05 narrow range (corresponding
        , to 395 wide range indication or 962 cubic feet) and 725 narrow range indica-tion (corresponding to 815 wide range indication or 1914 cubic feet). In M00E 4 with pressurizer pressure less than 430 psia, the safety injection tanks may be isolated.
          \
4
: b. With one safety-injection tank inoperable due to the isolation valve being closed, either imediately open the isolation valve or be in at least HOT STAND 8Y within one hour and be in HOT SHUTDOWN.within the next 12 hours.
SURVEILLANCE REQUIREMENTS 4.5.1    Each safety injection tank shall be demonstrated OPERABLE:
: a. At least once per 12 hours by:
: 1. Verifying the contained borated water level and nitrogen cover-pressure in the tanks is within the above limits, and
: 2. Verifying that each safety injection tank isolation valve is open and the nitrogen vent valves are closed,
: b. At least once per 31 days and within 6 hours after each solution level increase of >
boron concentrationof_7%        of tankinjection the safety    narrow tank rangesolution level, by  verifying the is between 4000 and 4400 ppm.
: c. At least once per 31 days when the RCS pressure is above 715 psia by verifying that power to the isolation valve operator is removed.
: d. At least once per 18 months by verifying that each safety injection tank isolation valve opens automatically under each of the following conditions:                                                .
: 1. When an actual or simulated the RCS pressure signals exceeds 515 psia, and
: 2. Upon receipt of a safety injection actuation test signal.
: e. At least once per 18 months by verifying OPERABILITY of the RCS-SIT differential pressure alarm by simulating RCS pressure >715 psia with SIT pressure <600 psig,
: f. At least once per 18 months, when the SIT's are isolated, by verify-ing the SIT nitrogen vent valves can be opened.
: g. At least once per 31 days, by verifying that power is removed from the nitrogen vent valves.
I i
i i
Amendment Number 9 February 27, 1984 3/4 5-2
 
                            -3/4.5.2                                ECCS SUBSYSTEMS                                                              l HOT STANDBY, STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall                                        !
be OPERABLE with each subsystem comprised of:
: a.            One OPERABLE high-pressure safety injection pump,
: b.            One OPERABLE low-pressure safety injection pump, and
: c.            An independent OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.
.                              APPLICABILITY: MODES 1, 2, and 3' ACTION:
: a.          With one ECCS subsystem inoperable, restore the inoperable subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
: b.            In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days i                                                          describing the circumstances of the actuation and the total accumu-l                                                          lated actuation cycles to date. The current value of the usage t                                                          factor for each affected injection nozzle shall be provided in this
!                                                          Special Report whenever its value exceeds 0.70.
b                    V    hXie preswdyt passu.u cys fra.Tt**. M. @ fa ( W o psia -
SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OP'ERABLE:
: a.          At least once per 12 hours by verifying that the following valves are in the indicated positions with power to the valve operators removed:
Valve Number                                Valve Function            Valve Position SI 604**                              High Pressure Hot Leg      Closed SI 609**                              Injection Line Isolation
                              **            Requirement applicable only if SI 604, SI 609, SI 321, SI 331 not each supplied by an independent and redundant emergency power source (four sourcestotal).                                                                    Amendment Number 9 February 27, 1984 3/4 5-3
: b. At least once per 31 days by:
: 1.        Verifying that each valve (manual, powered operated or automa-tic) in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position, and
: 2.    - Verifying that the ECCS piping is full of water by venting the
                                                                                              -' accessible discharge piping high points.
                                    .                                                                                                                                  4 4
: c. By a visual inspection which verifies that no loose debris (rags, trash clothing, etc) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. The visual inspection shall be performed:
: 1.        For all accessible areas of the containment prior. to establish-ing CONTAINMENT INTEGRITY, and
: 2.        For the affected areas within containment at the completion of containment entry when CONTAINMENT INTEGRITY is established.
: d. At least once per 18 months by a visual inspection of the contain-ment' sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump components (trash. racks, 4                                          screens, etc) show no evidence of structural distress or corrosion.
: e. At least once per 18 months, during shutdown; by:*
: 1.        Verifying that each automatic valve in the flow path actuates j                                                    to its correct position on a SIAS and RAS test signal, and
: 2.  . Verifying that each of the following pumps start automatically -
upon receipt of a Safety Injection Actuation Test Signal:
4
: a. High-Pressure Safety Injection pump.
,                                                    b. Low-Pressure Safety Injection pump.
: 3.      Verifying that on a Recirculation Actuation Test Signal, th l
h                        containme minimum ypas sump isolation valves open, the HPSI D LPS acirculation) flow line isolation valves close, mp a
3 and the LPSI pumps stop,                                                  w aw st reu
: f. By verifying that each of the following pumps develop the indicated differential pressure at or greater than their respective minimum allowable bypass recirculation flow rates whe tested pursuant to Specification 4.0.5:                                    g ah * '^ d              M
!                                  b      1.      High-PressureSafe$                              Injections e
(fre O g ic a J S e R-)
cp wmt
                                  @        2.      Low Pressure Safety Injectionkg                                          . ( % pO M CAld i
* The testing sequence shall not allow actual initiation of flow to the RCS.                                                                                    Amendment Number 9 February 27, 1984                      ,
I,
      - - . , ,-- ,w, ,n.n.-,--.v,.              _n,w      ,. - - -  rn,-r,,_-,n,--w,,-.-,-                          .,-my        ,,,,-n e-n  _ _ --e_,,--,.,,,v---,
 
                                                                                              \
: g. By verifying that correct position of each electrical and/or mechan-ical position stop for the following ECCS throttle valves:
: 1. Within 4 hours following completion of each valve's stroke testing or maintenance on the valve when the ECCS subsystems are required to be OPERABLE.
: 2. At least once per 18 months.
HPSI System          LPSI System          Hot Leg Injection Valve Number        Valve Number            Valve Number
: a. SI-617, SI-616      a. SI-615, SI-306    a. SI-604
: b. SI-627, SI-626      b. SI-625, SI-307    b. SI-609
: c. SI-637, SI-636      c. SI-635            c. SI-321
: d. SI-647, SI-646      d. SI-645            d. 51-331
: h. By perfoming a flow balance test during shutdown following comple-tion of modifications to the ECCS subsystem that alter the subsystem flow characteristics and verify the following flow rates:
x HPSI System - Single Pump                    LPSI System - Single rump a      jection Leg 1, 277 + 5 gpm            a. Injecti                        pm
: b. Injection Leg 2, 277 T 5 gpm            bp  Inject      eg 2, 2450 7 50 gpm
: c. Injection Leg 3,~277 7 5 gpmy-c . Injection Leg 3, 2450 7 50 gpm
: d. Injection' Leg 4, 277 T M pm            d. Injection Leg 4, 2450 T 50 gpm Simultan                    old Injection - Single Pump I              ot Leg, 545 + 20 gpm Cold Legs Total 545 + 20 g                                  _ _ _ _
  @    sm@
4 i
i Amendment Number 9 February 27, 1984 3/4 5-5 i
 
  /
HPSI System - Single Pump The sum of the injection line flow rates, excluding the highest                      '
flow rate, is greater than or equal to S W gps.
LPSI System - Single Pump V
: 1. Injection Loop 1, total flow equal to 4000 1 100 gpa
: 2. Injection Legs 1A and IB when tested individually, with                          ,
the other leg isolated, shall be within 166 gpm of each other.
r'
: 3. Injection Loop 2, total flow equal to 4000 ; 100 gpa  .
: 4. Injection Legs 2A and 28 when tested individually, with the otherlegisolated,shallbewithinlpgpmofeachother.                                                    !
Simultaneous Hot Leo and Cold Leo Injection - Single Pump                                                ;
w
: 1. Hot Leg, flow equal to--04; 1 324 gpa              *
                                                                                                                )
: 2. Cold Leg, flow equal to 545-t-20 gpa x                                                                    z
  ~ ~ - - - ~ _ . _ .                _ _ _ , . ~ ~ ~ - .  .
 
f ECCS SUBSYSTEMS HOT SHUTDOWN _AND u,4 S(AND6V LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem, comprised of the following shall be OPERABLE:
: a. An OPERABLE high-pressure safety injection pump, and
: b. An OPERABLE flow path capable of taking suction from the refueling water tank on a Safety Injection Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal.
APPLICABILITY: MODE 5k f a d 4 ACTION:
: a. With no ECCS subsystem OPERABLE, restore at least one ECCS subsystem to OPERABLE status within 1 hour or be in COLD SHUTDOWN within the next 20 hours,
: b. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumu-lated actuation cycles to date. The current value of the usage factor for each affected injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.
SURVEILLANCE REQUIREMENTS 4.5.3 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirementsg 4.5.2.
ef Sf aRwHeu.
    $cto Tolu p<rsstud tA.  $ pcsssau.        Atu M W (isla-4 Amendment Number 9 February 27, 1984 3/4 5-6
 
3/4.5.4        REFUELING WATER TANK LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water tank (RWT) shall be OPERABLE with:
                      --e-
@      a. A minimum Gntaineibborated water volume as specified in Figure 3.1-2, of Specification 3.1.2.5,
: b. A boron concentration between 4000 and 4400 ppm of boron, and
: c. A solution temperature between 60*F and 120*F.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within 6 hours and in COLD SHUT-DOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.5.4 The RWT shall be demonstrated OPERABLE:
: a. At least once per 7 days by:
: 1. Verifying the contained borated water volume in the tank, and
: 2. Verifying the boron concentration of the water.
: b. At least once per 24 hours by verifying the RWT temperature when the (outside) air temperature is outside the 60*F to 120*F range.
Amendment Number 9 February 27, 1984 3/4 5-7
 
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b            . - -
 
3/4.6        CONTAINMENT SYSTEMS 3/4.6.1        PRIMARY CONTAINMENT LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6.1.1        CONTAINMENT INTEGRITY See Applicant's SAR.
3/4.6.1.2        CONTAINMENT LEAKAGE See Applicant's SAR.
3/4.6.1.3 CONTAINMENT AIR LOCKS See Applicant's SAR.
3/4.6.1.4        CONTAINMENT ISOLATION VALVE AND CHANNEL WELD PRESSURIZER SYSTEM See Applicant's SAR.
3/4.6.1.5        INTERNAL PRESSURE See Applicant's SAR.                                                                I 3/4.6.1.6        AIR TEMPERATURE See Applicant's SAR.
3/4.6.1.7        CONTAINMENT STRUCTURE INTEGRITY See Applicant's SAR.
3/4.6.1.8        CONTAINMENT VENTILATION SYSTEM See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 6-1
 
                                                                              .~., . - .-              .
3/4.6.2          DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION F
3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWT on a Containment, Spray
  ~  Actuation Signal and automatically transferring suction to the containment sump on a Recirculation Actuation Signal. Each spray system flow path from the containment sump shall be via an OPERABLE shutdown cooling heat exchanger, j  APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
k  With one containment spray system inoperable, restore the inoperable spray f  system to OPERABLE status within 72 hours or be in at least HOT STANDBY within L  the next 6 hours; restore the inoperable spray system to OPERABLE status
[    within the next 48 hours or be in COLD SHUTDOWN within the following 30 hours.
I SURVEILLANCE REQUIREMENTS 4.6.2.1  Each containment spray system shall be demonstrated OPERABLE:
: a. At least once per 31 days be verifying that each valve (manual, power operated or automatic) in the flow path is positioned to take suction from the RWT on a Containment Spray Actuation test signal,
: b. By verifying, that each pump develops the indicated differential
  =
pressure of _>          sid at or greater than the minimum allowable bypass recircu ation flow rate when tested pursuant to Specification
]              4.0.5.
: c. At least once per 31 days by verifying that the system piping is full of water from the RWT to the header isolation valves.
: d. At least once per 18 months, during shutdown, by:
t i
        @        1. Verifying that each automatic valve in the flow path actuates to its correct cahodposition  on a(CSAS) and/
spray ar.fudeA                    RAS) test rhado-fice        ca.fwtsignal.'V8h.
3  '
: 2. Verifying that upon a Recirculation Actuation Test Signal, the containment sump isolation valves open and that a recirculation mode flow path via on OPERABLE shutdown cooling heat exchanger is established.
3.
  '                  Verifying that each spray pump starts automatically on a CSAS test signal.
  ;        e. At least once per 5 years by performing an air or smoke flow test 3
through each spray header and verifying each spray nozzle is unob-structed.
Amendment Number 9 3/4 6-2                          February 27, 1984
 
IODINE REMOVAL SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.2 The Iodine Removal System shall be OPERABLE with:
: a. A spray chemical addition tank containing a level of between 90% and g            100% (816 to 896 gallons) of between 33% and 35% by weight N H solution, d                                                          24
: b. Two spray chemical addition pumps each capable of adding N H
      @          solutionfromthespraychemicaladditiontanktoacontai$m$nt spray, pump flow.
sptw APPLICABILITY: MODES 1, 2, 3 and 4.
l ACTION:
,      With the Iodine Removal System inoperable restore the system to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours; restore the Iodine Removal System to OPERABLE status within the next 48 hours I-    or be in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS f
t'  4.6.2.2 The Iodine Removal System shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, l                sealed, or otherwise secured in position,- is-in its correct position.
i          b. At least once per 6 months by:
: 1. Verifying the contained solution volume in the tank, and
: 2. Verifying the concentration of the N24H solution by chemical analysis.                                                                  <
l
: c. By verifying, that on recirculation flow, each Spray Chemical addition pump develops a discharge pressure of 100 psig when tested pursuant to Specification 4.0.5.
: p.          d. At least once per 18 months, during shutdown, by:
g          1. Verifying that each automatic valve in the flow path actuates to its correct position on a CSAS test signal) acl.
: 2. Verifying that each spray chemical addition pump starts automa-tically on a CSAS test signal.
Amendment Number 9 February 27, 1984 3/4 6-3
: e. At least once per 5 years by verifying each solution flow rate (h
: u. - . . ,    ....  ...,,,3 pre-uyw .u =:1 t::t:} from the following drain connections in the spray additive system:
: 1.                                0.63 + 0.02 gpm.
: 2.      (Drain (Orain line linelocation) location) M I M gpm.
)
Amendment Number 9 February 27, 1984 3/4 6-4
 
                                                                                              ~l LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS
            '3/4.6.2.3          CONTAINMENT COOLING SYSTEM See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 6-5 l
 
  ..                        .                  .    .        .=
3/4.6.3          CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION                                                            l l
3.6.3 The containment isolation valve (s) specified in Table 3.6-1 shall be i
OPERABLE with isolation times as shown in Table 3.6-1. Unless specifically.
permitted by the Applicant's SAR the containment purge system shall be closed when operating at RATED THERMAL POWER.
APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With one or more of the isolation valve (s) specified in Table 3.6-1 and the containment purge valve specified in the Applicant's SAR inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and either:
: a. Restore the inoperable valve (s) to OPERABLE status within 4 hours, or                                        .
: b. Isolate each affected penetration within 4 hours by use of at least one deactivated automatic valve secured in the isolation position, or
: c. Isolate the affected penetration within 4 hours by use of at least one closed manual valve or blind flange, or
: d. Be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.6.3.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERA 3LE prior to returning the valve to service after maintenance, repair or replacement work is perfomed on the valve or its associated actuator, control or power circuit performance of a cycling test and verification of isolation                )
1-      time.                                                                                      ;
4.6.3.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE.at least once per 18                    )
months by verifying that on its applicable actuation test signal, each isola-              4 tion valve actuates to its required position. Valves which may be required to open or close following an accident will be actuated to demonstrate their capability to achieve both positions.
4.6.3.3 The isolation time of each power operated or automatic valve of Table 3.6-1 shall be detemined to be within its limit when tested pursuant to Specification 4.0.5.
Amendment Number 9 February 27, 1984 3/4 6-6 l
l
 
1 TABLE 3.6-1 CONTAINMENT ISOLATION VALVE ACTUATION TIMES i'
Valve Location    ESF      Required    Maximum Valve Penetratica                                                Relative to  Actuation  Pcst-Accident  Actuation Number      Yalve Number          Description          Containment    Signal  Valve Position  Time (sec)
A. Remotely Actuated Valves 11          SI-331      Hot Leg Injection Valve        Outside      None        Open          10 12          SI-321      Hot Leg Injection Valve        Outside      None        Open          10 l                                    13        SI-616, 617 14        SI-626, 627  High Pressure Cold 15        SI-636, 637  Leg Injection Valves          Outside      SIAS        Open          10 16        SI-646, 647 w              17          SI-615 3              18          SI-625      Low Pressure Cold
!                    o,            19          SI-635      Leg Injection Valves          Outside      SIAS        Open          10 4              20          SI-645 23          SI-673      Containment sump isolation    Inside        RAS        Open          20 valve
!                                                SI-674      Containment sump isolation    Outside        RAS        Open          20 l
valve 24          SI-675      Containment sump isolation    Inside        RAS        Open          20 valve
'                            ng                  SI-676      Containment sump isolation    Outside        RAS        Open          20 gg                                valve 5g        27          SI-690      Shutdown Cooling Warmup        Outside      None    Open or Closed      30 Qs                                bypass valve O$
SI-656      Shutdown Cooling isolation    Outside      None    Open or Closed      80 8                              valve hk 51-654      Shutdown Cooling isolation    Inside        None    Open or Closed    80
                              .                              valve
 
o
                                                    ' TABLE 3.6-1  (Cont'd)
CONTAINMENT ISOLATION VALVE ACTUATION TIMES d
Valve Location    ESF      Required      Maximum Valve Penetration                                              Relative to  Actuation _ Post-Accident-    Actuation Number      Valve Humber        Description            Containment    Signal  Valve Position    Time (sec) 28            SI-691  Shutdown Cooling Warmup          Outside      None    Open or Closed        30 bypass valve SI-655  Shutdown Cooling isolation      Outside      None    Open or Closed        80 valve SI-653  Shutdown Cooling isolation      Inside      None    Open or Closed        80 valve 29            SI-682  Safety Injection Tank fill      Inside      SIAS        Closed              5
,                                  and drain isolation valve w          40            CH-523  CVCS Letdown Line                Outside      CIAS        Closed              5 2                        CH-516  Isolation Valves                Inside    CIAS/SIAS    Closed              5
  ?          41            CH-524  CVCS Charging Line Iso-          Outside      None    Open or Closed          5 m                                lation Valves 43            CH-505  Reactor Coolant Pump Con-        Outside      CIAS        Closed              5 CH-506  trolled Bleedoff Contain-        Inside      CIAS        Closed              5 ment Isolation Valves 44            CH-560  Reactor Drain Tank              Inside      CIAS        Closed              5 CH-561  Suction Isolation Valves        Outside      CIAS        Closed              5 45            CH-580  Reactor Makeup Water Supply      Outside      CIAS        Closed              5 yy                              Isolation Valve to the RDT y
    =a 57            CH-255  Seal Injection Containment Isolation Valve Outside      None    Open or Closed          5 Q@
      " 8. Manual Valves J        29            51-463  Safety Injection Tank Fill      Outside      None        Closed      Not Applicable
    $%                              and Drain Isolation Valve N
 
TABLE 3.6-1    (Cont'd)
CONTAINMENT ISOLATION VALVE ACTUATION TIMES 1
Valve Location    ESF          Required    Maximum Valve 1                        Penetration                                                      Relative to  Actuation    Post-Accident      Actuation Number        Valve Number          Description              Containment    Signal    Valve Position      Time (sec) 41          CH-393      CVCS Charging Line                Inside      None          Closed    Not Applicable CH-854      Isolation Valves                  Outside      None          Closed    Not Applicable C. Check Valves i
11          SI-533      Hot Leg Injection Line            Inside      None-          Open      Not Applicable 12          SI-523      Isolation Valve 13          SI-113 R                    14          SI-123      High Pressure Cold Leg            Inside      None            Open      Not Applicable
* 15          SI-133      Injection Line Isolation
            ?                    16          SI-144      Valve 17          SI-114 18          SI-124      Low Pressure Injection            Inside      None            Open      Not Applicable
;                                19          SI-134      Line Isolation Valves 20          SI-144 41          CH-431      CVCS Charging Line                Inside      None      Open or Closed Not Applicabic CH-433      Isolation Valves 1                              45          CH-494      Reactor Makeup Water Supply        Inside      None          Closed    Not Applicable Isolation Valve to the ROT l                    n2        57          CH-835      Seal Injection Containment        Inside      None      Open or Closed Not Applicable
!                  fr 2                                  Isolation Valve 4
g=
                  *W
,                .N.,
I i
 
3/4.6.4        COMBUSTIBLE GAS CONTROL i  LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.6.4.1        HYDROGEN MONITORS
_See Applicant's SAR.
.i 3/4.6.4.2 ELECTRIC HYDROGEN RECOMBINERS See Applicant's SAR.                                                            ,
3/4.6.4.3        HYDROGEN PURGE CLEANUP SYSTEM i  See Applicant's SAR.
3/4.6.4.4        HYDROGEN MIXING SYSTEMS See Applicant's SAR.
3/4.6.4.5        PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM See Applicant's-SAR.
;  3/4.6.4.6        VACUUM RELIEF VALVES See Applicant's SAR.
1 1
I Amendment Number 9 February 27, 1984 3/4 6-10
 
3/4.6.5            IODINE CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 6-11
 
THIS PAGE INTENTI0f1 ALLY BLAf1K.
t 4
 
3/4.7                                PLANT SYSTEMS 3/4.7.1                                    TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION
_ (Q    3.7.1'.1                All main steam                      safety valves shall be OPERABLE with lift setting as specified in Table 3.7-1.
APPLICABILITY: MODES 1, 2, 3 and 4*.
ACTION:
: a.                With both reactor coolant loops and associa ed steam generators in q{} -                              operation and with one or more** main steam. ne co _ safety valves inoperable per steam generator, operation in MODES 1, 2, and 3 may proceed provided, that within 4 hours, either all the inoperable valves are restored to OPERABLE status or the maximum allowable steady state power level is reduced per Table 3.7-2; otherwise, be
,                                            in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
,                        b.                  Operation in MODES 3 and 4* may proceed with one reactor coolant loop and associated steam generator in operation, pro ded that.
there are no more than four inoperable main steam ne en safety (5)                            valves associated with the operating steam generator,)otherwise, be in COLD SHUTDOWK within the following 30 hours.
: c.                  The provision of Specification 3.0.4 are not applicable.
SURVEILLANCE RE0VIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
4 1
* Until the steam gererators are no longer required for heat removal.
I'
                **      The maximum number of inoperable safety valves on any operating steem                                                                          I generator is four (4).
I                                                                                                                                                                      l l
l 4
l                                                                                        3/4 7-1                            Amendment Number 9 February 27, 1984 T
                    ------,,--.s,-,----o-.wm-,-,--amm..m.,,--,,m          ,-me,-------wrew-      . . ~ -~,- -n- - -----    ,-------.--------.m ,+ -m , - . - --- --~-
 
TABLE 3.7-1 STEAM LINE SAFETY VALVES PER LOOP MINIMUM VALVE NUMBER                                                      LIFT SETTING (11%)*  RATED CAPACITY **
Line No. I                              Line No. 2
: a. See Applicants' SAR                    See Applicants' SAR                  1255 psig        904,000 lb/hr
: b.                      _
1255 psig        904,000 lb/hr
: c.                                                "
1290 psig        931,000 lb/hr
: d.                                                                              1290 psig        931,000 lb/hr-
: e.                                                                              1315 psig        950,000 lb/hr
):
[  f.
1315 psig        950,000 lb/hr
: g.                                                "
1315 psig        950,000 lb/hr
: h.                                                                              1315 psig        950,000 lb/hr
: 1.                                                                              1315 psig        950,000 lb/hr J.                                                                              1315 psig        950,000 lb/hr ng
  @g
* The lift setting pressure shall correspond to ambient conditions at the valve at nominal operating jg      temperature and pressure.
E+ **  Capacity is rated at lift setting +3% accumulation.f5ese capacities are based on a minimum total y@g    Qapacity of 19,530,900 lb/hr at 13bb psig (1315 psig + 3% accumulatio
  ;W co' 1
i                                                                                                                        .
 
6 TABLE 3.7-2    AM VAf1A8te N6RPo6 NtW W W%M MAXIMUM ALLOWABLE STEADY STATE POWER LEVEL WITH INOPERABLE STEAM LINE SAFETY VALVES DURING TWO LOOP OPERATION WITH FOUR PUMPS OPERATING g
Mximum Steady State Maximum Number of Inoperable Safety                                ,
_/        Power Level Valves on Any Operating Steam Generator                            '
( (Percent of RATED THERMAL POWER)
I                                  '
94.1 103 9 2
13.f                                          83.7 3                                    99,o                                          73,7 4                                                                                    62.8
;    m
      ?
w f4AXIHun AlLDLOABLE SfEADtl
:                                                                                          SW6 WG8. LGV6L CPERL6N(
cf RM6D @ESHAL. fbW6R)
O d
w.
Ei &
  ; a=-
    ~
f.
 
l LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7.1.2          Emergency Feedwater System See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 7-4
 
CONDENSATE STORAGE TANK LIMITING CONDITION FOR OPERATION 3.7.1.3 The condensate storage tank (CST) shall be OPERABLE with a level of
      ~a t least **                feet (300,000 gallons).
g    APPLICABILITY: MODES 1, 2, 37 od 4 *-*t.
ACTION:
With the condensate storage tank inoperable, within 4 hours either:
: a. Restore the CST to OPERABLE status or be in                                    east ^^                    ANDBY
  @                    within the next 6 hours and in HOT SHUTDOWN (T Cold 1 350'F) within the following 6 hours, or
: b. Demonstrate the OPERABILITY of the applicable alternative service water system with a water volume of at least 300,000 gallons as a backup supply to the emergency feedwater pumps and restore the condensate storage tank to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN with an OPERABLE shutdown cooling loop in operation within the following 6 hours.
i SURVEILLANCE REQUIREMENTS 4.7.1.3.1 The condensate storage tank shall be demonstrated OPERABLE at least once per 12 hours by verifying theJontained water volume) is within its limits
  @    when the tank is the supply source for the emergency feedwater pumps.
i.M 4.7.1.3.2 The applicable alternative water source shall be demonstrated OPERABLE at least once per 12 hours by verifying:
: a. That the alternate water source to Emergency Feedwater System isolation valves are either open or OPERABLE whenever the alternate water source is the supply source for the emergency feedwater pumps, and                                                      -
: b. That the alternate water source contains a water level of at least
                              ** feet (300,000 gallons) whenever the alternate water source is the supply source for the emergency feedwater pumps.
    #*              See Applicant's SAR.
R(.wid 11a. sttawt ydwr W m Js9t. .arguiud, for _lich.ameuAl..
4%f @co64,-tdux. cosd/&o*.13 $4 yJLu..
3/4 7-5                              Amendment Number 9 February 27, 1984 a-
 
ACTIVITY LIMITING CONDITION FOR OPERATION I
3.7.1.4 The specific activity of the secondary coolant system shall be less than or equal to 0.10 microcuries/ gram DOSE EQUIVALENT I-131.
                -APPLICABILITY: MODES 1, 2, 3 and 4.
ACTION:
With the specific activity of the secondary coolant system greater than 0.10 micro- curies / gram DOSE EQUIVALENT I-131, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.4 The specific activity of the secondary coolant system shall be determined to be within the limit by performance of the sampling and analysis program of Table 4.7-1.
4 k
Amendment Number 9 4
February 27, 1984 3/4 7-6
  , ., ,, .--.      , . , , _ _ , _ - . - _ . . . , , _ ,..,,_,,.._._m      ,, . _ _ , . . . , . . ~ . . . _ , , , - . _ . - . . .-....,..,.m__ -
                                                                                                                                                        . - - , , . . _ . , .,___.,,-.__,,__m-,
 
TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT                              SAMPLE AND ANALYSIS AND ANALYSIS                                      FREQUENCY
: 1. Gross Activity Determination                  At least once per 72 hours
: 2. Isotopic Analysis for. DOSE        a)        1 per 31 days, whenever the gross EQUIVALENT I-131 Concentration                activity determination indicates iodine concentrations greater than 10% of the allowable limit.
b)        1 per 6 months, whenever the gross activity determination indicates iodine concentrations below 10% of the allowable limit.
J l
Amendment Number 9 February 27, 1984 3/4 7-7 I
 
MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
: h. APPLICABILITY: MODES 1, 2h 3, oxd 1.
ACTION:
MODE 1          -
With one main steam line isolation valve inoperable but open, POWER OPERATION may continue provided the inoperable valve is re, stored to OPERABLE status within 4 hours; otherwise, be in at 1            " " " " " " + " - ' " - - - + " '      ^ ~ ~ ~ ' ' ' ' " ~ ~
: w. t!-l$
                                "^'' lc555E k 35''E N s M N 5 2 d ' N' N i MODES 2, 3, and 4          -
With one main steam line isolation valve inoperable, subsequent operation in MODES 2, 3, and 4 may proceed provided:
!                      a.      The isolation valve is maintained closed.
: b.      The provisions of Specification 3.0.4 are not applicable.
Otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.7.1.5NEachmainsteamlineisolationvalveshallbedemonstratedOPERABLE by verifying full closure within M seconds when tested pursuant to Specifi-cation 4.0.5.                                        %6
: 9. T. I . G.1    @wt A 4
Amendment Number 9 February 27, 1984 3/4 7-8
 
O 4.7.1.5.2 The provisions of Specification    0.4arenotapplicableforenths into MODE 3 or MODE 4 to perform the surveillance testing of Specification
  /
i 4.7.1.5.1 provided the testing is performed within 12 hours after achieving normal operating steam pre:sure and normal operating temperature for the        ,
secondary side to perform the test.
j t        _ _ _                                                                .
1 I
J
 
f                                              l h        Pf, ANT SYST9tg y - [y 5 a55 ocia ed
                                                                                                                          .. l l
ATNDSPHEluc Otfip VAlvts                                        blo<.E V&Ive -
(
LIMITD66 CONDITTON FCit OPERATION Gasi.                        9            -
3.7.L 8                atmospheric dump valve 0 shall be OPERA 8LI.
APPUCASILITY: MODES 1, 2, 3, and 4.8 h:                              dwof
: 2. With less than                atmospheric dump valv                steam generator OPERA 4LE, restore the remaired atmosoneric duma valve to OPERA 8LI status within 72 hours; or be in at least HOT STANDEY within the next 6 hours.
{        SURVIILLANCE REQUIREMENTS
[.gg[ b loc l( 4 fv a L_
: 4. 7.1. 8    Each atmospheric dump valvejsnall be demonstrated CPERA8 [
: a. At least once per 24 himes by verifying that the nitrogen accumu tant is at a pressure > 400 PSIG.
to startup follow            any refueling shutdown or cold shutdown Nof                  ,          d f,3;y, 'Y8 'r log                            enat all alves will open and close
                                                                                  & ,.,f                    va l u s  au    ssoe uale>
byvey&ggn3                                            csyh e r;e blos k                                                                      ]
    \
b          Vu W  any blocIc vale ef) Naf eraUs , rest 6/e. 111e blac N VelVe V)
                                                                                                                . se h e in nor s w oe
                'ro      OP@ABL E                  stafa.s          wi %                                                                              ;
dh;n 4 hou's an d                              hot uuToow    7 day $A',, theNUo"'ik                                              \
b h.avs. Lim;Y u te of ft,e 2 Ss o cis ted atm osply erie damy Valvefs) darin3 tio:s period .
L                                                                                                                                        o h steam generators are being used for decay heat removal.
3/4 7- S a
 
                                                                                      =_
l 3/4.7.2        STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2 The temperature of the secondary coolant in the steam generators shall be greater than **F when the pressure of the secondary coolant in the steam generator is greater than
* psig.
APPLICABILITY: At all times.
ACTION:
With the requirements of the above specification not satisfied:
: a. Reduce the steam generator pressure to less than or equal to
* psig within 30 minutes, and
: b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam genera-tor. Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above **F.
SURVEILLANCE REQUIREMENTS 4.7.2      a. The pressure in the secondary side of the steam generators shall be determined to be less than
* psig at least once per hour when the temperature of the secondary coolant is less than
                  * 'F.
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* See Applicant's SAR.
Amendment Number 9 February 27, 1984      1 3/4 7-9
 
LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7.3        COMPONENT COOLING WATER SYSTEM See Applicant's SAR.
3/4.7.4        SERVICE WATER SYSTEM See Applicant's SAR.
LtelsHAfE H6AY SIAK 3/4.7.5        EMER6ENCtt00L4NG-PONG-See Applicant's SAR.
3/4.7.6        FLOOD PROTECTION See Applicant's SAR.
3/4.7.7        CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM See Applicant's SAR.
3/4.7.8        ECCS PUMP ROOM EXHAUST AIR CLEANUP SYSTEM See Applicant's SAR.
Amendment Number 9 February 27, 1984 3/4 7-10
 
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3/4.7.9        SNUBBERS LIMITING CONDITION FOR OPERATION
            @                              3.7.9    8''
redber: 'i;te i . Tebles 3.7-Oe er.d 0.7-Ob shell be 0^ "J,0;.C. Iosset A APPLICABILITY: MODES 1, 2, 3, and 4.                                              (MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES).                                                                                                                                                          i ACTION:
See Applicant's SAR.
SURVEILLANCE REQUIREMENTS See Applicant's SAR.
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                                                                                                                                                                                                                                                \
Amendment Number 9 February 27, 1984 3/4 7-11 1
      . _ . . . . _ _ , . _ _ _ . . . _ _ . _            , . _ _ . _ , _ . _ _ . _ _ _ , _ _ _ , _ _ _ _ .    ..c.__ , , _ . . . , , . . _ , , _ _ , , , , , , , , _ . _ _ , _ _ _ _ _ _ _ , _ , . . _ _ _ _ _ _ _ _ _ . _ _ _ _ , _ _ .
 
      <        All hydt aulic and mechanical snubbers shall be OPERA 8LE. The only snubbers excluded from this requirement are those installed on nonsafety-N      i related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-
[      related system.
                              ._.                    ~ _ . . -- ....  . - - _ -
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3/4.7.10        SEALED SOURCE CONTAMINATION LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS See Applicant's SAR.
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Amendment Number 9 February 27, 1984 3/4 7-14
 
w 3/4.7.11                  FIRE SUPPRESSION SYSTEMS LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.7.11.1                    FIRE SUPPRESSION WATER SYSTEM See_ Applicant's SAR.
3/4.7.11.2                    SPRAY AND/0R SPP.INKLER SYSTEM                                    -
l      See Applicant's SAR.
3/4.7.11.3                    CO SYSTEMS 2
j      See Applicant's SAR.
3/4.7.11.4                    HALON SYSTEM See Applicant's SAR.
      -3/4.7.11.5                    FIRE HOSE STATIONS See Applicant's SAR.
3/4.7.11.6                    YARD FIRE H'lDRANTS AND HYDRANT HOSE HOUSES See Applicant's SAR.
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Amendment Number 9
                                        >                                            February 27, 1984 3/4 7-15 4
 
6 3/4.7.12        FIRE BARRIER PENETRATIONS LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS See Applicant's SAR.
1 i                                                              Amendment Number 9 February 27, 1984 3/4 7-16
 
3/4.7.13        AREA TEMPERATURE MONITORING LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS See Applicant's SAR.
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Amendment Number 9 February 27, 1984 3/4 7-17
 
3/4.7.14        SHUTDOWN COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.14 Two independent shutdown cooling subsystems shall be OPERABLE, with each subsystem comprised of:
: a. One OPERABLE low-pressure safety injection pump, and
: b. An independent OPERABLE flow path capable of taking suction from the RCS hot leg and discharging coolant through the shutdown cooling heat exchanger and back to the RCS through the cold leg injection lines.                                                                                  -
ud h APPLICABILITY: MODES 1, 2,33, ACTION:
: a. With one shutdown cooling subsystem inoperable, restore the inoper-able subsystem to OPERABLE status within 72 hours or be in at least HOT STANDBY within one hour, be in at least HOT SHUTDOWN within the next 6 hours and be in COLD SHUTDOWN within the next 30 hours and continue action to restore the required subsystem to OPERABLE status,
: b. With both shutdown cooling subsystems inoperable, restore one subsystem to OPERABLE status within one hour or be in at least HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 6 hours and continue action to restore the required subsystems to OPERABLE status.
: c. With both shutdown cooling subsystems inoperable and both reactor coolant loops inoperable, initiate action to restore the required subsystems to OPERABLE status.                                          -
p SURVEILLANCE REQUIREMENTS 4.7.14 Each shutdown cooling subsystem shall be demonstrated OPERABLE:
: a. At least once per 18 months, during shutdown, by establishing shutdown cooling flow from the RCS hot legs, through the shutdown cooling heat exchangers, and returning to the RCS cold legs,
: b. At least once per .18 months, during shutdown, by testing the automa-                      l tic and in.erlock action of the shutdown cooling system connections                        i from the RCS. The shutdown cooling system suction valves shall not open when RCS pressure is > (*) psia. The shutdown cooling system                          i suction valves located outside containment shall close automatically when RCS pressure > (*) psia. The shutdown cooling system suction valve located inside containment shall close automatically when RCS pressure > 700 psia.
Amendment Number 9
* See Applicant's SAR.                                                  February 27, 1984 3/4 7-18
 
3/4.8        ELECTRICAL POWER SYSTEMS 3/4.8.1        A.C. SOURCES LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS 3/4.8.1.1        OPERATING See Applicant's SAR.
3/4.8.1.2        SHUTDOWN See Applicant's SAR.
P Amendment Number 9 February 27, 1984 i                                                3/4 8-1
                                    . _ . _ _ - - _ - _ _ _ , . _ _ . _ _ . ~ . _ _ _ _ _ _ _ _ _ . _ _ _                                _ _ -
 
3/4.8.2        D.C. SOURCES LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8.2.1        OPERATING See Applicant's SAR.
3/4.8.2.2        SHUTDOWN See Applicant's SAR.
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Amendment Number 9 February 27, 1984 3/4 8-2 1
 
3/4.8.3        ONSITE POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8.3.1        OPERATING See Applicant's SAR.
3/4.8.3.2        SHUTDOWN See Applicant's SAR.
1 Amendment Number 9 February 27, 1984 3/4 8-3
    = S .- m          +  -
                                -e-.                              ,-w- -e-    y y        w~m-w-        - y          -m--    -1*-r-
 
3/4.8.4        ELECTRICAL EQUIPMENT PROTECTIVE DEVICES LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.8.4.1        CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEV-ICE 5 See Applicant's SAR.
3/4.8.4.2        -MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES See Applicant's SAR.
S
                                                  /
Amendment Number 9 February 27, 1984 3/4 8-4
 
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    '3/4.9          REFUELING OPERATIONS 3/4.9.1          BORON CONCENTRATION l
LIMITING CONDITION FOR OPERATION l
3.9.1 With the reactor vessel head closure bolts less than fully tensioned or with the head removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling pool shall be maintairied uniform and sufficient to ensure that the more restrictive of following reactivity condi-tions is met:
a ._  Either a K,ff of 0.95 or less, or
: b. A boron concentration of greater than or equal to 2150 ppm.
APPLICABILITY: MODE 6*.
ACTION:
With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 40 gpm is of a solution reduced        containing to less          > 4000 than or equal    ppm or to 0.95 boron  or its equivalent the boron              until concentration is K'Nstored to greater than or equal to 2150 ppm, whichever is the more restrictive.
SURVEILLANCE REQUIREMENTS
!    4.9.1.1 The more restrictive of the above two reactivity conditions shall be I
determined prior to:
l
: a. Removing or unbolting the reactor vessel head, and l-
!        .b. Withdrawal of any full length CEA in excess of 3 feet from its fully        l l                inserted position within the reactor pressure vessel.                      !
4.9.1.2 The boron concentration of the reactor coolant system and the refuel-            i ing pool.shall be detemined by chemical analysis at least once per 72 hours.
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* The reactor shall be maintained in MODE 6 whenever fuel is in the reactor vessel with the reactor vessel head closure bolts less than fully tension-ed or with the head removed.
Amendment Number 9 February 27, 1984 3/4 9-1
 
3/4.9.2        INSTRUMENTATION
' LIMITING CONDITION FOR OPERATION 3.9.2 As a minimum, two sources range neutron flux monitors shall be OPERABLE and operating, each with continuous visual indication in the control room and one with audible indication in the containment and control room.
APPLICABILITY: MODE 6.
ACTION:
: a. With one of the above required monitors inoperable or not operating, insnediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes,
: b. With both of the above required monitors inoperable or not operating, determine the boron concentration of the reactor coolant system at least once per 12 hours.
SURVEILLANCE REQUIREMENTS 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of:
: a. A CHANNEL CHECK at least once per 12 hours,
: b. A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
: c. A CHANNEL FUNCTIONAL TEST at least once per 7 days.
Amendment Number 9 February 27, 1984 3/4 9-2
 
3/4.9.3        DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for at least 72 hours.
APPLICABILITY: During movement of irradiated fuel in the reactor pressure vessel.
ACTION:
With the reactor subcritical for less than 72 hours, suspend all operations involving movement of irradiated fuel in the reactor pressure vessel.                                      ,
SURVEILLANCE REQUIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least 72 hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.
                                                                                                                              't l
l Amendment Number 9 February 27, 1984 3/4 9-3
_ . . _ . _ . . ~ _ _ - - - . _ _ . . _ - . _ _ _ . _ . . _ _ _                  _ . . _ _ _ .
 
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.9.4        CONTAINMENT BUILDING PENETRATIONS See Applicant's SAR.
3/4.9.5        COMUNICATIONS See Applicant's SAR.
3/4.9.6        REFUELING MACHINE OPERABILITY See Applicant's SAR.
3/4.9.7        CRANE TRAVEL - SPENT FUEL P0OL BUILDING See Applicant's SAR.
Amendment Number 9 February 27, 1984  ;
3/4 9-4
 
3/4.9.8        SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL                                                                                l LIMITING CONDITION FOR OPERATION 3.9.8.1  At least one shutdown cooling loop shall be OPERABLE and in opera-tion.*
APPLICABILITY: MODE 6 when the water level above the top of the reactor pressure vessel flange is greater than or equal to 23 feet.
ACTION:
With no shutdown cooling loop OPERABLE and in operation, suspend all opera-tions involving an increase ir. the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmos-phere within 4 hours.
SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one shutdown cooling loop shall be verified to be in opera-tion and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm at least once per 12 hours.
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* The shutdown cooling loop may be removed from operation for up to I hour per 8 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.
    .                                                                        Amendment Number 9 i                                                                        February 27, 1984 f                                          3/4 9-5
 
LOW WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.2 Two independent shutdown cooling loops shall be OPERABLE and at least              ,
one shutdown cooling loop shall be in operation *.                                          '
@ APPLICABILITY: MODE        when the water level above the top of the reactor.
pressure vessel flange is less than 23 feet.
ACTION:
: a. With less than the required shutdown cooling loops OPERA 8LE, imme-diately initiate corrective action to return the required loop to OPERA 8LE status, or to establish greater than or equal to 23 feet of water above the reactor pressure vessel flange, as soon as possible.
: b. With no shutdown cooling loop in operation, suspend all operations involving an increase in reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation. Close all containment penetrations providing direct access from tne containment atmosphere to the outside atmos-phere within 4 hours.
SURVEILLANCE REQUIREMENTS.
4.9.8.2 At least one shutdown cooling loop shall be verified to be in opera-tion and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm at least once per 12 hours.
* The shutdown cooling loops may be removed from operation for up to 1 hour per 8 hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.
l Amendment Number 9 February 27, 1984  ,
3/4 9-6                                          :
 
3/4.9.9        CONTAINMENT PURGE VALVE ISOLATION SYSTEM l.IMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS See Applicant's SAR.
I Amendment Number 9 February 27. 1984 3/4 9-7
 
l 3/4.9.10        WATER LEVEL - REACTOR VESSEL N MITING CONDITION FOR OPERAllun                                                  j
  ~
'3.9.        least 23 feet of water shall be maintained over the top of reactor      ure vessel flange.
APPLICA8ILITY    u    g movement of fuel assemblies or CEAP w thin the reactor pressure vessel who either the fuel assemblies being        ed or the fuel assemblies sedted within N the reactor pressure vess    are irradiated.
ACTION:
With the requirements of the above        (fication not satisfied, suspend all operations involving rnvement o uel assemblies or CEAs within the pressure vessel.
SURVEILLANCE      IREMENTS 4  . The water level shall be determined to be at least it        mum requir-4d depth within 2 hours prior to the start of and at least once per          hours thereafter during movement of fuel assemblies or CEAs.
l Amendment Number 9 February 27, 1984 3/4 9-8
 
t FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION i
3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor pressure vessel flange.
{
      !  APPLICA8ILITY: During movement of fuel assemblies within the reactor pressure
      !  vessel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are irradiated.
:  ACTION:
I  With the requirements of the above specification not satisfied, suspend all l  operations involving movement of fuel assemblies within the pressure vessel.
SURVEILLANCE REQUIREMENTS 4.9.10.1 The water level shall be determined to be at least its minimum
!          required depth within 2 hours prior to the start of and at least once per 24 hours thereafter during movement of fuel assemblies.
f
 
4  __
(
s                                                                            (
                                                                        ' ~~                        !
LIMITING CONDITION FOR OPERATION i
k 3 .9.10.2 At least 23 feet of water shall be maintained over the top of the I fuel seated in the reactor pressure vessel.
t lAPPLICASILITY: During movement of CEAs within the reactor pressure vessel,
                ; when the fuel assemblies seated within the reactor pressure vessel are
                ' irradiated.
t l ACTION:
1 With the requirements of the above specification not satisfied, suspend all
          ,.        operations involving movement of CEAs within the pressure vessel.
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SURVEILLANCE REQUIREMENTS
      !              4.9.10.2 The water level shall be determined to be at least its minimum required depth within 2 hours prior to the start of and at least once per i            24 hours thereafter during movement of CEAs.
e
 
3/4.9.11        WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet,8 inches of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY: Whenever irrediated fuel assemblier are in the storage pool.
ACTION:
With the requirement of the specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours.
SURVEILLANCE REQUIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
Amendment Number 9
    .                                                                February 27, 1984 3/4 9-9 m
 
.3/4.9.12        STORAGE POOL AIR CLEANUP SYSTEM LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS See Applicant's SAR.
Amendment Number 9 February 27. 1984 3/4 9-10
 
3/4.10          SPECIAL TEST EXCEPTIONS 3/4.10.1          SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s) or the reactor is subcritical by at least the reectivity equivalent of the highest CEA worth.
APPLICABILITY: MODE 2 and 3 ACTION:
: a. With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediate-ly initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUT 00WN MARGIN required by Specification 3.1.1.1 is restored.
: b. With all full length CEAs fully inserted and the reactor subcritical by less than the above reactivity equivalent, imediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 4000 ppm boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length and part length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours.
4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 24 hours prior to reducing the SHUTDOWN MARGIN to less than the limits of Spect-fication 3.1.1.1.
4.10.1.3 When in MODE 3, the reactor shall be determined to be subcritical by at least the reactivity equivalent of the highest estimated CEA worth or the reactivity equivalent of the highest estimated CEA worth is available for trip insertion from OPERABLE CEA'S at least once per 2 hours by consideration of at least the following factors:
* Operation in MODE 3 shall be limited to 6 consecutive hours.
Amendment Number 9 February 27, 1984 3/4 10-1
 
o.1. Reactor coolant system boron concentration, CEA position,
            @        b t.
c Q. Reactor coolant system average temperature, 4 4. Fuel burnup based on gross thermal energy generation, e.1. Xenon concentration, and f1    Samarium concentration.
l Amendment Number 9 February 27, 1984 l                                                                  3/4 10-2 l
l
 
3/4.10.2          MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMIT 5 LIMITING CON 0! TION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion and power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7 and the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided:
: a. The THERMAL POWER is restricted to the test power plateau which shall not exceed 85% of RATED THERMAL POWER, and
: b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below.
APPLICABILITY: MODES I and 2.
ACTION:
With any of the limits of Specification 3.2.1 being exceeded while the require-ments of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6. 3.2.2, 3.2.3, 3.2.7 and the Minimum Channels OPERABLE requirement of I.C.1 (CEA Calculators) of Table 3.3-1 are suspended, either:
: a. Reduce THERMAL POWER sufficiently to satisfy the requirements of Specification 3.2.1, or
: b. Be in HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be detennined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7 or the Minimum Channels OPERABLE require-mentofI.C.1(CEACalculators)ofTable3.3-1aresuspendedandshallbe verified to be within the test power plateau.                                      l 4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector          ,
Monitoring System pursuant to the requirements of Specifications 4.2.1.2 and        1 3.3.3.2 during PHYSICS TESTS above 20% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.2.2, 3.2.3, 3.2.7 ortheMinimumChannelsOPERABLErequirementofI.C.1(CEA Calculators)ofTable3.3-1aresuspended.
Amendment Number 9 February 27, 1984 3/4 10-3
 
3/4.10.3          REACTOR COOLANT LOOPS LIMITING CONDITION FOR OPERATION 3.10.3 ThelimitationsofSpecificationk.4.1.bandnotedrequirementsof Tables 2.2-1 and 3.3-1 may be suspended during the performance of startup and PHYSICS TESTS, provided:
: a. The THERMAL POWER does not exceed Si of RATED THERMAL POWER, and
: b. The reactor trip setpoints of the OPERABLE power level channels are set at less than or equal to 20% of    TED THERMAL POWER.
OC      C. . Beth twtcrcoM Jcops & M                    m aadec & p q w to.gx APPLICABILITY: During startup and PHYSICS TESTS. N "^2 W We*t-ACTION:
With the THERMAL POWER greater than 5% of RATED THERMAL POWEL, immediately trip the reactor.
                                                                      }
@    SURVEILLANCE REQUIREMENTS wmm"g~g M
* 4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%
of RATED THERMAL POWER at least once per hour during startup and PHYSICS TESTS.
4.10.3.2 Each logarithmic and variable overpower level neutron flux monitor-ing channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating startup and PHYSICS TESTS.
LO.
                  %o4wf, M d6AAM7 4.10 3 3 k .br ca y tali n a d % f a d*r & 'N W*
haak Ha ?^- l2 4wtum-Amendment Number 9 February 27, 1984 3/4 10-4
 
3/4.10.4                                        CEA POSITIONJNO REGULATING CEA INSERTION LIMIT!L AND                                                                                                                                  '
RUCTa4 C" M T COLD LEG TfMPEMTV4r" LIMITING CONDITION FOR OPERATION AND 2.t,6 3.10.4 The requirements of Specifications 3.1.3.1 antr 3.1.3.6(may be suspend-1 ed during the perfonnance of PHYSICS TESTS to determine the isothermal temper-ature coefficient, moderator temperature coefficient and power coefficient provideditM '' 'te a' tascificatian 1 S 1 == == 4ataiad -f it. c.;;.e4 e;
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I                                                                                                                                                                                                                                                                      -en        l 3/4 10-5                                                                                        _ - .
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APPLICABILITY: MODES 1 and 2.
ACTION:
: 4. With any of the limits of Specification 3.2.1 re being exceeded while the require-suspended, either:
ments of Specifications 3.1.3.1)And 3.1.3. as 3 t 4 Reduce THERMAL POWER sufficiently o satisfy the requirements of l f.      Specification 3.2.1LorWs7 din i How t, $ . Se in H0T STAN08Y within 6 hours.
vaein,N i ets TNsu aA E00AL
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: i. Insaisrm      aan,xi    THERHAl        PowfR ro sariuv THE    R10 uinCNfN. rs of    Sirciricari a s.a, on
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SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be detemined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.3.
3.1.3.6Jre suspended and shall be verified to be within the tes[1 power aa#er plateau \.yang/e4 3,2,(,
4.10.4.2 The linear heat rate shall be detemined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specification 3.3.3.2 during PHYSICS TESTS above 20% of RATED THERMAL POWER in which the requirements of Specifications 3.1.3.1;aa# er 3.1.3.6 are suspended.
NuoAnxr.L 4.n g.s [ e-p w n              vA CUDM .3 mas t M AffrdM1Mfb Tu R1 enam rusa w incur <                os usarany av soanonws tr nr wsr sver.m so. Huivrts pysins PHYsrS TES rs on wana ras orn >>umnr< as Sircisicarion ., 1, a w x,n nnnin a.io.u Tu, Pracroa C-r L 1<< rsusanarsu SMALL Rf hiYfAMlMfD To Af WJfMIM TM/ L)Mlf$
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            'InDjCA TJoM    A'T LCAsT aner )TA Mot)R            DtJAAAl&
PHYSICS TESTS su wnua rar arouirrazurs os    Sn,n,,-a d a~,x                  su,---o W **    El    ffL)tAAIT s  t  Y                                                                        .
4        io .x 1                                                    .
I
 
SPECIAL TEST EXCEPTIONS 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION I
3.10.5 The minimum temperature and pressure for criticality limits of Speci-fications 3.1.1.4 and 3.2.8 may be suspended during low temperature PHYSICS TESTS to a minimum temperature of 300'F and a minimum pressure of 500 psia provided-                                                                                -
                                                                                                )
l
: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER.
: b. The reactor trip setpoints on the OPERABLE Variable Overpower trip channels are set at 5,20% of RATED THERMAL POWER, and                        {
: c. The Reactor Coolant System temperature and pressure relationship is maintained within the acceptable region of operation required by              l Specification 3.4.8 except that the core critical line shown on Figure 3.4-2 does not apply.
APPLICABILITY: MODE 2".
ACTION:
: a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.
f
: b. With the Reactor Coolant System temperature and pressure relationship within the region of unacceptable operation on Figure 3.4-2, immediately open the reactor trip breakers and restore the temperature pressure relationship to within its limit within 30 minutes; perform the engineering evaluation required by                  I
!                  Specification 3.4.8.1 prior to the next reactor criticality.
SURVEILLANCE REQUIREMENTS 4.10.5.1 The Reactor Coolant System temperature and pressure relationship shall be verified to be within the acceptable region for operation of Figure 3.4-2 at least once per hour.
4.10.5.2 The THERMAL POWER shall be determined to be < 5% ~  of RATED THERMAL POWER at least once per hour.
4.10.5.3 The Reactor Coolant System temperature shall be verified to be greater than or equal to 300'F at least once per hour.
4.10.5.4 Each Logarithmic Power Level and Variable Overpower channel shall be            l subjected to a CHANNEL FUNCTIONAL TEST within 12 hours prior to initiating low          !
temperature PHYSICS TESTS.                                                              l
      *First core only, prior to first exceeding 5% RATED THERMAL POWER.
PALO VERDE - UN                              3/410-h6
 
,                                                                                                                                                                s
                                                                            +o g                                      4N
                                  '      ~
'                                                                                                      s 3/4.10.U          SAFETY INJECTION TANKS                        ,
LIMITING CONDITION FOR OPERATION 3.10.[ The Safety Injection Tank isolatfori valve requirement of Specification partial stroke testing of the low pressure l                        3.5.la    may be safety injection      suspended check    valves. during(SI-114, SI 124, SI-134, SI-144)                                                        provided:
: a.      That power to the isolation valve is restored and the SIAS signal is not overridden.
!                .            b.      Only one isolation valve at a time is closed during the testing for no longer than one hour.
l          &                  c.
Vsq- CDdttd-That the valve is opened x
with power removed before the next isolation valve is closed.
APPLICABILITY: While partial stroke testing of the low pressure injection check valves during nonnal plant operation.
ACTION;      If requirement J.S.la was suspended to perform the Specification 3.10,5" partial stroke test an if any of the Specification 3.10.Prequirements are not met during the'3.10. partial stroke testing, the Limiting Condition for Operation shall revert to Specification 3.5.1 and the 3.5.1 ACTION shall
  .,                    be applicable.
s-                                                                                      l SURVEILLANCE REQUIREMENTS                              _ _ _
      "                4.10. k.1 A valve alignment shall be performed within 4 hours following completion of testing to verify that all valves operated during this testing are restored to their normal positions and that power is removed to the SIT isolation valves.                                                          ~
h k                        g        '%
            ~                                                          '
s            .
                                                                      \'l                                                  ,
Amendment Number 9 February 27, 1984 3/410-0              \
y'~        e!'
                                                                                                            ~ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
 
                                                                                                                            =
3/4.10.d        SAFETY INJECTION TANK PRESSURE LIMITING CONDITION FOR OPERATION 3.10.6I The safety injection tank (SIT) pressure of Specification 3.5.1d may be suspended for low temperature PHYSICS TESTS provided:
: a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER;
: b. The SITS have been filled per Specificat n 3.5.lb and pressurized
                                                                                                'ssur N dhh M N c.
hk          h ').Nc N M .El.
All valves in the injection lines from the SITS to the RCS are open and the SITS are capable of injecting into the RCS if there is a                                                I decrease in RCS pressure.
APPLICABILITY: MODES 2,        3, CL4) 4 ACTION:
IfallthegITsdonotmeetthelevelandpressurerequirementsofSpecifica-                                                            i tion 3.10.b restore all the SITS to meet these requirements or be in HOT STANDBY within 6 hours and be in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.10.d.1 The THERMAL POWER shall be determined to be less than 5% of RATED THERMAL POWER at least once per hour during low pressure FriYSICS TESTS.
4.10.f.2 Every 8 hours verify:
: a. All the SITS levels meet the requirements of Specificaton 3.5.lb.
: b. All the SITS pressures meet the requirements of Specificaton 3.10. .
: c. The valve alignment from the SITS to the RCS has not changed.
=
Amendment Number 9 February 27, 1984 3/4 10-7
 
k THIS PAGE INTENTI0f1 ALLY BLAllK, f-l l
l
 
BASES FOR
  ,              SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION
  ,                      AND SURVEILLANCE REQUIREMENTS i
l Amendment Number 9 >
      ,                                                                  February 27. 1984
 
THIS PAGE INTENTIONALLY BLANK, l
1
 
NOTE The BASES contained in the succeeding pages sumarize the reasons for the specifications of Sections 3.0 and 4.0 but, in accordance with 10CFR50.36, are not a part of these Technical Specifications.
Amendment Number 9 February 27, 1984 B 3/4-N
 
O THIS PAGE INTENTIONALLY BLAllK.
 
i e
3/4.0-      APPLICABILITY BASES i
The specifications of this section provide the general requirements applicable to each of the Limiting Conditions for Operation and Surveillance Requirements within Section 3/4.
3.0.1 This specification defines the applicability of each specification in terms of defined OPERATIONAL MODES or other specified conditions and is provided to delineate specifically when each specification is applicable.                                                      ,
3.0.2 This specification defines those conditions necessary t' constitute compliance with the terms of an individual Limiting Condition 1or Operation and associated ACTION requirement.
3.0.3 This specification delineates the measures to be taken for circumstances not directly provided for in the ACTION statements and whose occurrence would violate the intent of a specification. For example, Specification 3.6.2.1 requires two Containment Spray Systems to be OPERABLE and provides~ explicit ACTION requirements if one spray system is inoperable. Under the terms of Specification 3.0.3, if both of the required Containment Spray Systems are inoperable, within one hour measures must be initiated to place the unit in at least HOT STANDBY within the next 6 hours, in at least HOT SHUTDOWN within the following 6 hours, and in COLD SHUTDOWN in the subsequent 24 hours.
3.0.4 This specification provides that entry into an OPERATIONAL MODE or other specified applicability condition must be made with (a) the full comple-ment of required systems, equipment or. components OPERABLE and (b) all other 4
parameters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out of service provisions contain-ed in the ACTION statements.
4
,              The intent of this provision is to insure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded.
Exceptions to this specification have been provided for a limited number as i-            specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropri-ate specifications.                                                                                                              i 4.0.1 This specification provides that surveillance activities necessary to insure the Limiting Conditions for Operation are met and will be performed during the OPERATIONAL MODES or other conditions for which the Limiting Conditions for Operation are applicable. Provisions for additional surveil-
            -lance activities to be performed without regard to the applicable OPERATIONAL MODES or other conditions are provided in the individual Surveillance Require-ments. Surveillance Requirements for Special Test Exceptions need only be performed when the Special Test Exception is being utilized as an exception to an individual specification.
Amendment Number 9 February 27, 1984 i
B 3/4 0-1
 
i 4.0.2 The provisions of this specification provide allowable tolerances for performing surveillance activities beyond those specified in the nominal surveillance interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a surveillance frequency does not negate this allowable tolerance value and permits the performance of more frequent surveil-lance activities.
The tolerance values, taken either individually or consecutively over 3 test inter- vals, are sufficiently restrictive to ensure that the reliability associated with the surveillance activity is not significantly degraded beyond that obtained from the nominal specified interval.
4.0.3 The provisions of this specification set forth the criteria for deter-mination of compliance with the OPERABILITY requirements of the Limiting Conditions for Operation. Under this criteria, equipment, systems or compon-ents are assumed to be OPERABLE if the associated surveillance activities have been satisfactorily performed within the specified time interval. Nothing in this provision is to be construed as defining equipment, systems or components OPERABLE, when such items are found or known to be inoperable although still          '
meeting the Surveillance Requirements.
4.0.4 This specification ensures that the surveillance activities associated with a Limiting Condition for Operation have been performed within the speci-fied time interval prior to entry into an OPERATIONAL MODE or other applicable condition. The intent of this provision is to ensure that surveillance activities have been satisfactorily demonstrated on a current basis as requir-ed to meet the OPERABILITY requirements of the Limiting Condition for Opera-tion.
Under the terms of this specification, for example, during initial plant startup or following extended plant outages, the applicable surveillance activites must be performed within the stated surveillance interval prior to placing or returning the system or equipment into OPERABLE status.
4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of TSME Code Class 1, 2 and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI'of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Connission and is not part of these Technical Specifications.
This specification includes a clarification of the frequencies for performing the inservice inspection and testing activites required by Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. This clarifica-
          . tion is provided to ensure consistency in surveillance intervals throughout these Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
Amendment Number 9 February 27, 1984 8 3/4 0-2
 
I l
Under the terms of this specification, the more restrictive requirements of          !
the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. For example, the requirements of Specifi-cation 4.0.4 to perform surveillance activities prior to entry into an OPERA-TIONAL MODE or other specified applicability condition taked precedence over the ASME Boiler and Pressure Vessel Code provision which allows pumps to be tested up to one week after return to normal operation. And for example, the Technical Specification definition of OPERABLE does not grant a grace period before a device that is not capable of performing its specified function is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of perfoming its '
specified function for up to 24 hours before being declared inoperable.
Amendment Number 9 February 27, 1984 B 3/4 0-3
 
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I
 
3/4.1          REACTIVITY CONTROL SYSTEMS BASES                                                                                          ,
l i
3/4.1.1          B0 RATION CONTROL l
3/4.1.1.1 and 3/4.1.1.2              SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcriti-cal from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, assuming the insertion of the regulating CEA's are within the limits set b, of %ec$miien
      @  3.1.3.6, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tc. The most restrictive condi-tion occurs at E0L, with Tc at no load operating temperature, and is asso-ciated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN of 6.0% delta k/k is required to control the reactivity transient. Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with the criteria used to establish the power dependent CEA4nser-tion limits and with the assumptions used in the FSAR safety analyses.-WWith
      . Tc less than or equal to 210 F, the reactivity transients resulting from uncontrolled    RCS cooldown are minimd1 and :Sut h ergia re';ui            r:=t: ar; ;;t
                - " ' -      --**''4 N t v = 4 =a + e  -- +'-- ' ^ -  ' '
                                                                                "        '-N h  J+ Z
        ,t- --- l.!,'C mmm    N% 'miniga.1 o W Ak/w"'EiaM8ti>im . y' -*
IM
                                                                                                  ~
3/4.1.1.3          MODERATOR TEMPERATURE COEFFICIENT (MTC)
The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the accident and transient analysis remain valid through each fuel cycle. The surveillance requirements for measurement of the MTC during each fuel cycle are adequate to confirm the MTC value since
    ,    this coefficient changes slowly due principally to the reduction in RCS boron I
concentration associated with fuel burnup. The confirmation that the measured MTC value is within its limit provides assurances that the coefficient will be maintained within acceptable values throughout each fuel cycle.
3/4.1.1.4          MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System cold leg temperature less than 552*F. This limitation
  .I    is required to ensure (1) the moderator temperature coefficient is within its analyzed temperature range, (2) the protective instrumentation is within its normal operating range, (3) the pre::;rizcr is ;;pbh of beins in on GF;"#LE status with a steam bubble, (4) the reau.ur p.ciagr; vessel is eb;;: it
      @  mh' r "T Analysis. NDT te weraturc cad (1) t ensure consistency with the FSAR Safety 3
Amendment Number 9 February 27, 1984 B 3/4 1-1
 
l 1
3/4.1.2          B0 RATION SYSTEMS The boron injection system ensures that negative reactivity control is avail-able during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3)
  . m separate flow paths, and 4) an emergency power supply from OPERABLE diesel W  generators. Dsert A                                                                                                        l i
With the RCS avecago temperature above 210*F, a minimum of two separate and i
redundant boron injection systems are provided to ensure single functional                                                  I
;    capability in the event an assumed failure renders one of the systems inoper-
;    able. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.
The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN.from expected operating conditions of 6.0% ak/k after xenon decay and cooldown to 210*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires (*)
gallons of 4000 ppm borated water from either the refueling water tank or the spent fuel pool.
With the RCS temperature below 210*F one injection system is acceptable without single failure considerations on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable. The restrictions of one and only one OPERABLE
,    charging pump, whenever the reactor coolant level is below the bottom of the pressurizer, is based on the assumptions used in the analysis of the boron dilution event.
The boron capability required below 210*F is based upon protiding a 4% ak/k SHUTDOWN MARGIN after xenon decay and cooldown from 210 F aM 120*F. This
  @  condition requires (*) gallons of 4000 ppm borated water from either the refueling water tank or the spent fuel pool.
The values of water volume, temperature and boron concentration in the refuel-ing water tank are provided to ensure that the assumptions used in the initial conditions of the LOCA Safety Analyses remain valid.
The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.
With the RCS temperature below 210*F while in Modes 5 and 6, a source of borated water is required to be available for reactivity control and makeup for losses due to contraction and evaporation. The requirement of 33,500 gallons of 4000 ppm borated water in either the refueling water tank or spent fuel pool ensures that this source is available.
BORON DILUTION ALARMS The startup channel high neutron flux alams alert the operator to an inadver-tent boron dilution. Both channels must be operating to assure detection of a Amendment Number 9 February 27, 1984
* See Applicant's SAR.
B 3/4 1-2 1
              ~              -
                                                          ,,..,,,,_._,,,u,        - , . , , , . . .      mc, . , . , . _ . ~ - -
 
7 l
l 4
7 Ib
                                                                                                                        ~~~
                      . . ~ _ _ _ _ . -- - . - . - _ _ _ _ . . _ _ _ _ _ . . _ _ . . . _ _ _  . _ -
                                                                                                                                ~ ' N x '
The nominal capacity of each charging pump is 44 gpa at its dis-                          g charge. Up to 16 gpa of this may be diverted to the volume control tank via                                        '
the RCP control bleedoff. Instrument inaccuracies and pump performance uncertainties are limited to 2 gpu yielding the 23 gpa value.
i I
l
 
boron dilution event by the high neutron flux alarms.            If one or both of the alarms are inoperable at anytime, the bases for ACTION statements are as follows:
: a. One startup channel high neutron flux alarm not operating:
With only one startup channel high neutron flux alarm OPERABLE while in MODE 3, 4, 5, or 6, a single failure to the alarm could prevent detection of boron dilution. By periodic monitoring of the RCS boron concentration by either boronometer or RCS sampling, a decrease in the boron concentra-This tion during provides      an inagygg;
                      ; diver:e sa ra boron      dilution event ndet methodlof            will beofobserved.
detection      boron dilution, with sufficient time for termination of the event beforegreturn to g        b d;;rt:nt criticality.                              ce,qtztr.au q smoowM HAErN M
: b. Both startup channel high neutton flux alarms not operating:
When both startup channel high neutron flux alarms are inoperable, there is no means of alarming on high neutron flux when subcritical. Therefore, Jtinw simultaneous use of boronometggCS sampling to monitor the RCS boron concentration provides dhen =. red =drt ind cations of an inadver-tent boron dilution. This will allow detection w'th sufficient time for g terminati              b g i gtion beforef eturn to S            ::rt:nt criticality.
3/4.1.3          MOVABLE CONTROL AS            ES b W Rc.5Mca&d*ds@                          Y The specifications of this section ensure that (1) acceptable power distribu-tion limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of CEA misalignments are limited to acceptable levels.
The ACTION statements which permit limited variations from the basic require-ments are accompanied by additional restrictions which ensure that the origi-nal design criteria are met.
The ACTION statements applicable to a stuck or untrippable CEA, to two or more inoperable CEAs and to a large misalignment (greater than or equal to 19 inches) of two or more CEAs, require a prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOWN MARGIN.
For small misalignments (less than 19 inches) of the CEAs, there is 1) a small effect on the time dependent long tem power distributions relative to those used in generating LCOs and LSSS setpoints, 2) a small effect on the available SHUTDOWN MARGIN, and 3)' a small effect on the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with small misalignments of CEAs pemits a one hour time interval during which attempts may be made to restore the CEA to within its alignment requirements. The one hour time limit is sufficient to (1) identify causes of a misaligned CEA, (2) take appropriate corrective action to realign the CEAs and (3) minimize the effects of xenon redistribution.
The CPCs provide protection to the core in the event of a large misalignment (greater than or equal to 19 inches) of a CEA by applying appropriate penalty factors to the calculation to account for the misaligned CEA. However, this misalignment would cause distortion of the core power distribution. This                          j B 3/4 1-3                  Amendment Number 9 February 27, 1984 l
 
distribution may, in turn, have a significant effect on 1) the available
                              . SHUTDOWN MARGIN, 2) the time. dependent long term power distributions relative to those used .in. generating LCOs and LSSS setpoints, and 3) the ejected CEA e                            worth used in the safety analysis. Therefore, the ACTION statement associated with the large misalignment of a CEA requires a prompt realignment of the misaligned CEA.
The ACTION state'ments applicable to misaligned or inoperable CEAs irclude                                      i requirements to align the OPERABLE CEAs in a given group with the inoperable 3 ..                            CEA. Conformance with these alignment requirements bring the core, within a
:short period of time, to a configuration consistent with that assumed in                                        i i
generating LCO and LSSS setpoints. .However, extended operation with CEAs significantly inserted in the core may lead to perturbations in 1) local                                        ~
burnup, 2) peaking factors and 3) available shutdown margin which are more                                    -!
adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination. Therefore, time limits have been imposed on
'                              operation with inoperable CEAs to preclude such adverse conditions from developing.
Operability of at least two CEA position indicator channels is required to
  ,~                          determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits. The CEA " Full In" and " Full Out" limits provide an additional independent means for determining the CEA positions when the CEAs I
i are at either their fully inserted of fully withdrawn positions. Therefore, the ACTION statements-applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the " Full In" or " Full Out" limits.
CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours with more frequent veriff-cations required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
The maximum CEA drop time restriction is consistent with the assumed CEA. drop time used i the safety analyses. Measurement with T                        greater than or b          equal to 55kF and with all reactor coolant pumps opeMUng ensures that the                                          I e                            measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.
Several design steps were employed to accommodate the possible CEA guide tube wear which could arise from CEA vibrations when fully withdrawn. Specifical--
i                            ly, a programmed insertion schedule will be used to cycle the CEAs between the
;                            full out position-(" FULL OUT" LIMIT) and 3.0 inches inserted over the fuel i                            cycle. This cycling will distribute the possible guide tube wear over a larger area, thus minimizing any effects. To accomodate this programmed
                            -insertion' schedule, the fully withdrawn position was redefined, in some cases, 1
                            . to be 144.75 inches (193 steps) or greater.
The establishment of LSSS and LCOs require that the expected long and short term behavior of the radial peaking factors be determined. The long term
'                            behavior relates to the variation of the steady state. radial peaking factors with core burnup and is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed and the expect-ed power level variation throughout the cycle. The short term behavior Amendment Number 9          l
,                                                                  B 3/4 1-4                                        February 27, 1984          !
L                                                                                                                                                ;
I
 
relates to transient perturbations to the steady-state radial peaks due to radial xenon redistribution. The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions and load maneuvering. Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analyses CEA inser-tions are determined and a consistent set of radial peaking factors defined.
The Long Term Steady State and Short Term Insertion Limits are determined based upon the assumed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used. The limits specified serve to limit the behavior of the radial peaking factors wi(Min the bounds determined from analysis. The actions specified serve to limit the extent of radial xenon redistribution effects to those acconinodated in the analyses. The Long and Short Term Insertion Limits of Specification 3.1.3.6 are specified for the plant which has been designed for primarily base loaded operation but which has the ability to accommodate a limited amount of load maneuvering.
The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion Limits of Specification 3.1.3.5 ensure that 1) the minimum SHUTDOWN MARGIN is maintained, and 2) the potential effects of a CEA ejection accident are limited to acceptable levels. Long tenn operation at the Transient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to deter-mine the behavior of the radial peaking factors.
l 1
Amendment Number 9
                                                                                                            *        "U '
8 3/4 1-5
 
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                            .3/4.2                POWER DISTRIBUTION LIMITS BASES 3-                            3/4.2.1                  LINEAR HEAT RATE The -_ limitation en linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.
Either of the two core power distribution monitoring systems, the Core Operat-
'                            ing Limit Supervisory System (COLSS) and the Local Power Density channels in the Core Protection Calculators (CPCs), provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. - The COLSS performs this function by continuously monitoring the core power distribution and calculating a core power operating limit corresponding to the allowable peak linear heat rate. Reactor operation at or below this calculated power level assures that the limits of 14.0 kw/ft are not exceeded.
The COLSS calculated core power and the COLSS calculated core power operating limits based on linear heat rate are continuously monitored and displayed to 4
                            -the operator. A COLSS alam is annunciated in the event that the core power exceeds the core power operating limit. This provides adequate margin to the linear heat rate operating limit for normal steady state operation.. Nomal                                              '
reactor power transients or equipment failures which do not require a reactor
;                            trip may result in this core power operating limit being exceeded. In the event this occurs, COLSS alarms will be annunciated. If the event which causes the COLSS limit to be exceeded results in conditions which approach the core safety limits, a reactor trip will be initiated by the Reactor Protective Instrumentation. The COLSS calculation of the linear heat rate includes appropriate penalty factors which provide, with a 95/95 probability / confidence level, that the maximum linear heat rate calculated by COLSS is conservative with respect to the actual maximum linear heat rate existing-in the core.
These penalty factors are determined from the uncertainties associated with planar radial peaking measurement, engineering heat flux uncertainty, axial densification, software algorithm modelling, computer processing, rod bow and
      ,                    core power measurement.
  .i 1-l                    Parameters required to maintain the operating limit power level based on 4      i I
linear heat rate, margin to DNB and total core power are also monitored by the                                            ,
CPC's (assuming minimum core power of 20% of RATED THERMAL POWER). The 20%                                                l RATED THERMAL POWER threshold is due to the neutron flux detector system being inaccurate below 20% core power. Core noise level at low power is too large
      !..                  to obtain usable detector readings. Therefore, in the event that the COLSS is                                            i
      ;                    not being used, operation within the limits of Specification 3.2.1 can be                                                {
      +
maintained by utilizing a predetermined local power density margin and a total                                            '
* '~j                      core power limit in the CPC trip channels. The above listed uncertainty
              @            penalty factors plus those associated withAIIIIt*UPAaCCeptance vcr iteria are also included in the CPC's.                                Du ctc.    . tut t
l Amendment Number 9 February 27, 1984 8 3/4 2-1 1
 
l l
l ptANAR' 3/4.2.2' a                            RADIAL PEAKING FACTORS                                                                                                          l Limiting the values of the PLANAR-RADIAL PEAKING FACTORS (Fc ) used in the                                                                                            l E                            COLSS and CPCs to galues equal or to greater than the measuNd PLANAR RADIAL PEAKING and the CPCs                      FACTORS remai      (F"Ev)alid.provides Data from theassurance    incore detectors that the      are      limits usedcalculated for                    by COLSS detemining the measured PLANAR RADIAL PEAKING FACTORS. A minimum core power at 20% of RATED THERMAL' POWER is assumed in determining the PLANAR RADIAL PEAKING FACTORS. The 20% RATED THERMAL POWER threshold is due to the neutron flu'x detector system being inaccurate below 20% core power.                                                              Core noise level
:                            at low power .is too large to obtain usable detector readings. The periodic.
.                            surveillance requirements for determining the measured PLANAR RADIAL PEAKING i
FACTORS provides assurance that the PLANAR RADIAL PEAKING FACTORS used in COLSS and the CPCs remain valid throughout the fuel cycle. Determining the measured planar radial peaking factors after each fuel loading prior to exceeding 70% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.
3/4.2.3                              AZIMUTHAL POWER TILT-T q The limitations on the AZIMUTHAL POWER TILT are provided to ensure that design safety margins are maintained. An AZIMUTHAL POWER TILT greater than 0.10 is not expected and if it should occur, operation is restricted to only those conditions required to identify the cause of the tilt. The tilt is normally L                            calculated by COLSS. A minimum core power of 20% of RATED THERMAL POWER is assumed by the CPCs in its input to COLSS for calculation of AZIMUTHAL POWER TILT.' The 20% Rated Themal Power threshold is due to the neutron flux 1                        . detector system being inaccurate below 20% core power. Core noise level at
!                            low power is too large to obtain usable detector readings. The surveillance
;                            requirements specified when COLSS is out of service provide an acceptable means of detecting the presence of a steady state tilt. It is necessary to explicitly account for power asymetries because the radial peaking factors
:                            used in the core power distribution calculations are based on an untilted
!                            power distribution.
l i
        'Q                  4ka.Aunusat.PeWEAfitf is eg to (Petit /P tat)- t.o tulud AZIMUTHAL POWER TILT is measured by assuming that the ratio of the power at any core location in the presence of a tilt to the untilted power at the location is gi.er, by; of rw form :
P tilt/Puntilt        =1+T3cesq          Mh                                                                      -
where: _ZIMUTHALPOWERTILT(T)iscalculatedfromthefollowingfunctio%q-rm: To = T;
* g
* cos(e - S ) f
  .                                              T                        isth'e$fl ap tude N "                                        b                                .
          @                                        9 9                    is the ne=li::d rediel tilt f=: tion MMM/quy fa' der' e
is the azimuthal eniSewce(tocatiem is the azimuthal crier,teti....of maximum tilt
[                                                  e{1lt/Puntilt P                        is the ratio of the power at a core location in the presence of a tilt to the power at that location with no tilt, i
1 Amendment Number 9 February 27, 1984 8 3/4 2-2
      ..  . _ _ . ~ . . _ . . _ . _ _ - _ . . . _ _ , - _ . .                                        _ . , . _          _ . _ . _ _      - , . . , _ .          _ . . - _ _ _ . _ _ _            -
 
3/4.2.4          DNBR MARGIN I
The limitation on DNBR as a function of AXIAL SHAPE INDEX represents a conser-            I vative envelope of operating conditions consistent with the safety analysis                I assumptions and which have been analytically demonstrated adequate to maintain            !
an acceptable minimum DNBR throughout all anticipated operational occurrences,            l of which the loss of flow transient is the most limiting. Operation of the core with a DNBR at or above this limit provides assurance that an acceptable minimum DNBR will be maintained in the event of a loss of flow transient.
Either of the two core power distribution monitoring systems, the Core Operat-            ;
ing Limit Supervisory System (COLSS) and the DNBR channels in the Core Protec-tion Calculators (CPC's), provide adequate monitoring of the core power distribution and are capable of verifying that the DNBR does not violate its limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating a core operating limit corresponding to the allowable minimum DNBR. Reactor operation at or below this calculated power level assures that the limits of Figure 3.2-1 are not violated.4The COLSS alculation of core power operating limit based on the minimum DNBR limit                i ncludes appropriate penalty factors which provide, with a 95/95 probability /
h_,onfidence level, that the core power limit calculated by COL 55 (based on the
  '@  Qinimum DNBR Limit) is conservative with respect to the actual core power limit. These penalty factors are detemined from the uncertainties associated with planar radial peaking measurement, engineering heat flux, state parameter heasurement, software algorithm modelling, computer processing, rod bow and
      - p ore power measurement.
Parameters required to maintain the margin to DNB and total core power are also monitored by the CPC's. Therefore, in the event that the COLSS is not being used, operation within the limits of Figure 3.2-2 can be maintained by utilizing a predetemined DNBR as a function of AXIAL SHAPE INDEX and by monitoring the CPC trip channels. The above listed uncertainty and penalty factors are also included in the CPCs which assume a minimum core power of 20%
      @    of RATED THERMAL POWER. The 20% Rh%8 Tik%hel PbWh threshold is due to the neutron flux detector system being inaccurate below 20% core power. Core noise level at low power is too large to obtain usable detector readings.-
The DNBR penalty factors listed in'Section 4.2.4.4 are penalties used to accomodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly. Fuel assemb-lies that incur higher average burnup will experience a greater magnitude of -
rod bow. Conversely, lower burnup assemblies will experience less rod bow.
The penalty for each batch required to compensate for rod bow is determined from a batch's maximum average assembly burnup applied to the batch's maximum integrated planar-radial power peak. A single net penalty for COLSS and CPC is then detemined from the penalties associated with each batch, accounting for the offsetting margins due to the lower radial power peaks in the higher i-        burnup batches.
3/4.2.5        RCS FLOW RATE This specification is provided to ensure that the actual RCS total flow rate              l is maintained at or above the minimum value used in the safety analyses.
Amendment Number 9 February 27, 1984 8 3/4 2-3 i.
 
3/4.2.6        REACTOR COOLANT COLD LEG TEMPERATURE This specification is provided to ensure that the actual value of reactor coolant cold leg temperature is maintained within the range of values used in the safety analyses.
3/4.2.7        AXIAL SHAPE INDEX This specification is provided to ensure that the actual value of core average AXIAL SHAPE INDEX is maintained within the range of values used in the safety analyses.
3/4.2.8        PRESSURIZER PRESSURE This specification is provided to ensure that the actual value of pressurizer pressure is maintained within the range of values used in the safety analyses.
Amendment Number 9 February 27, 1984 8 3/4 2-4
 
l .
4 3/4.3                INSTRUMENTATION 4-                          BASES 3/4.3.1 and 3/4.3.2                                REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPTRABILITY of the reactor protective and ESFAS instrumentation systems and by$pages ensure that 1) the associated ESFAS action and/or reactor trip will be iniE)tiated when the parameter monitored by each channel or combina-tion thereof reaches its setpoint, 2) the specified coincidence logic is 4
maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional                                                                                  .
capability is available from diverse parameters.
The OPERABILITY of these systems if required to provide the overall reliabil-l                          ity, redundancy and diversity assumed available in the facility design for the
            @              protection integrated operationand mitigation          of each of of IWth;3 f    't and transient conditions. The these= systems is consistent with the assump-tions used in the safety analyses,                                gg e design' of the Control Element Assembly Calculators (CEAC) provides reactor i      ksgt protection in the event one or both CEACs become inoperable. If one CEAC is ir@) in test or inoperable, verification of CEA position is perfonned at least every four hours. If the second CEAC fails, the CPC's in conjunction with plant Technical Specifications will use DNBR and LPD penalty factors and increased DNBR and LPD margin to restrict reactor opeJr tion to a power level
;.                          that will ensure safe operation of the plant.                                            If the g margins are not main-tained, a reactor trip will occur.
.                          The surveillance' requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original
                          ' design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
The measurement of response time at the specified frequencies provides assur-ance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit                                                                                    l was taken in the analys for those channels with response times indicated as                                                                                      ;
not applicable. Ins
!                          Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total
                          ' channel response time as defined. Sensor response time verification may be                                                                                      !
demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.
3/4.3.3                MONITORING INSTRUMENTATION                                                                                                                1 3/4.3.3.1                RADIATION MONITORING INSTRUMENTATION
                          'See Applicant's SAR.
Amendment Number 9 February 27, 1984 8 3/4 3-1 l
i
    ,,.  - . _ - - , , ,            , _ - . ~ _      . , - _ . _ _ _ .      .          _      __      . . . . _ , _ _ _ . . . - . . - . _ _ _ _ - , . . - . -                            -
 
                                                                                                                                                ~
                                                              ~- _        _
Response time testing of resistance temperature devices, which are a part-
  '  of the reactor protective system, shall be performed by using in-situ loop current test techniques or another NRC approved method.
Any modifications which are made to the core protection calculator soft-ware (including changes of algorithms and fuel cycle specific data) shall be performed in accordance with "!"! ". L. M... Aivo,....._ rr''-                                    ^ ::;; " :::
      ....,' ; G OL'M ", "..".:....                  : f !;;!: :-* 1 ", ".:;':':- 01 ; ;;;tt:- NRC
    , approved procedure on CPC software modifications.
k.
N The response times in Table 3.3-2 are made up of T the time to generate the trip signal at the detector (sensor response time) and i the time for the signal to interrupt power to the CEA drive mechanism (signal or trip delay time). The response times are taken from the sequence-of-events Tabps in Section 15 of CESSAR.
i 1
i
 
3/4.3.3.2          IN-CORE DETECTORS The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.
3/4.3.3.3          SEISMIC INSTRUMENTATION See Applicant's SAR.
3/4.3.3.4          METEOROLOGICAL INSTRUMENTATION
:            See Applicant's SAR.
SY5fEH 3/4.3.3.5          REMOTE SHUTDOWN, INSTRUMENTATION systw gwn4 instrumentation                        ensures that sufficient
      @ The OPERABILITY of the remote                  shcapability Tutdown  and maintenance    is available      of HOT STANDBY to permit.
of the facility from locations outside the control room. This capability is required in the' event control room habitability is lost and is consistent with
;            General Design Criteria 19 of 10CFR50.
The parameters selected to be monitored ensure that (1) the condition of the reactor is known, (2) conditions in the RCS are known, (3) the steam genera-tors are available for residual heat removal, (4) a source of water is avail-able for makeup to the RCS, and (5) the charging system is available to makeup water to the RCS.
3/4.3.3.6          POST-ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consis-h          tent with the recomendations of Regulatory Guide 1.97, " Instrumentation for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG 0578, "TMI-2 Lessons Learned h6        Task Force Status Report and Short-Tenn Recommendations."
3/4.3.3.7            CHLORINE DETECTION SYSTEMS See Applicant's SAR.
3/4.3.3.8            FIRE DETECTION INSTRUMENTATION See Applicant's SAR.
3/4.3.3.9            LOOSE-PART DETECTION INSTRUMENTATION See Applicant's SAR.
3/4.3.4          TURBINE OVERSPEED PROTECTION See Applicant's SAR.
Amendment Number 9 February 27, 1984 B 3/4 3-2                                                              1 1
h
  -g  r  .-.,-.:-,.,..- -~,      - . . , , .                                  ,.n.,_--,,,_,n-.      -nm,.e.,.    .-
 
m                      .__                                  _. . . . _              __                                                                              ,
4 4
4 l
4 f
The Subcooled Margin Monitor (SM), the Heat Junction Themocouple (HJTC), ,
  !                [' and    the Core Exit instrumentation                        required        Thermocouples                  (CET) comprise by Itse II.F.2 NUREG-0737,                                  the Post the  TMI-2Inadequate Action Plan. tCore Cool
                  '  The function of the ICC instrumentation is to enhance the ability of the plant                                                                                    1' i                operator to diagnose the approach to existance of, and recovery from ICC.
Additionally, they aid in tracking reactor coolant inventory. These instruments are included in the Technical Specifications at the request of NRC Generic Letter 83-37. These are not required by the accident analysis, nor to bring the plant to Cold Shutdown.                                                                                                                                            i i            i                                                                                                                                                                          !
In the event more than four sensors in a Reactor Vessel Level channel                                                                                        j are inoperable, repairs may only be possible during the next refueling outage.                                                                                        f This is because the sensors are accessible only after the missile shield and                                                                                                      ,
l              t reactor vessel head are removed. It is not feasible to repair a channel                                                                                                            !
except during a refueling outage when the missile shield and reactor vessel                                                                                                        l head are removed to refuel the core. If both channels are inoperable, the                                                                                              j channels shall be restored to OPERA 8LE status in the nearest refueling out-                                                                                          /
[
age.      If only one channel is inoperable, it is intented that this channel be                                                                                  /
restored to OPERA 8LE status in a refueling outage as soon as reasonably possible.                                                                                                                                        ~
4 6
l
        . - .-        , - .      _ _ _ _ . _ . _ _ . _ . _ _ . - . _ - . ~                            _ . _ - _ . . . _ . _ _ _ _ _ _ _ . . . _ _ . -_ _ _ _ _ _                            _ - _ _  l
 
i
                                                                                                                                                                                                                  -1
              - 3/4.4                            REACTOR COOLANT SYSTEM
                                                                                                                                                                                                                  .l BA_SES 3/4.4.1                                  REACTOR COOLANT LOOPS AND COOLANT CIRCULATION l.23I The plant is designed to operate with both reactor coolant loops and associat-
        .@      ed reactor coolant pumps in operation, and maintain DNBR above -hfe-during all                                                                                                                    l nomal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour.                                                                                                                                                  t In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capabil-ity for removing decay heat; but single failure considerations required that at least two loops (either shutdown cooling or RCS) be OPERABLE. Thus, if the reactor coolant loops are not OPERABLE, this specification requires that two shutdown cooling loops be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.
            - The operation of one Reactor Coolant Pump or one shutdown cooling pump pro-vides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. A flowrate of at least 4000 gpm will circulate one equivalent reactor coolant system volume of 12,097 cubic feet in approximately 23 minutes. The reactivity change rate associated with boron reductions will, 7
therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump during MODES 4 and 5 with                                                                                                                        l one or more RCS cold leg less than or equal to * 'F during cooldown or * *F                                                                                                                          !
during heatup are provided to prevent RCS pressure transients, caused by                                                                                                                            l energy additions from the secondary system, which could exceed the limits of                                                                                                                        l Appendix G to 10 CFR Part 50. The RCS'will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting start-ing of the RCPs to when the secondary water temperature of each steam genera-                                                                                                                        ,
      - OE      tor is less than M9'F above each of the RCS cold leg temperatures.                                                                                                                                  '
l00
* See Applicant's SAR.
4 Amendment Number 9 February 27, 1984 i
B 3/4 4-1 p.w+  v- w.-._      ,~y_. + . ,_, _ . , _ _ , _ _ - , _ _ . , _ _ _ . . _ , . - , - _        _ _ . . . . _ . . , . - - _ , - - - . - - _ - _ . -        . _ _ . . _ , . , . . , _ _ - - , _ _ _ , _ - , , -
 
5--
N T_
3/4.4.2        SAFETY VALVES                                                              q The pressurizer code safety valves operate to prevent the RCS from being                  [
pressurized above its Safety Limit of 2750 psia. Each safety valve is design-ed to relieve a minimum of 460,000 lbs per hour of saturated steam at the                  -d valve setpoint plus accumulation. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur                      1
  @  during shutdown. In the event that no safety valves are OPERABLE ; M i W ;                f cpr:tions at leo th w .2t rat, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS                    a overpressurization.                                                                        j During operation, all pressurizer code safety valves must be OPERABLE to                  I prevent the RCS from being pr'essurized above its safety limit of 2750 psia.        '
T The combined relief capacity of these valves is sufficient to limit the System            :
pressure to within its Safety Limit of 2750 psia following a complete loss of              '
turbine generator load while operating at RATED THERMAL POWER and assuming no              't=
reactor trip until the first Reactor Protective System trip setpoint (Pressur-izer Pressure-High) is reached (i.e., no credit is taken for a direct reactor              4 trip on the loss of turbine) and also assuming no operation of the steam dump              4 valves.                                                                                      1 Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI              =
of the ASME Boiler and Pressure Vessel Code.                                                )
3/4.4.3        PRESSURIZER h,
An dPERABLE pressurizer provides pressure control for the reactor coolant                    "
system during operations with both forced reactor coolant flow and with                      j natural circulation flow. The minimum water level in the pressurizer assures                g the pressurizer heaters, which are required to achieve and maintain pressure                    -
control, remain covered with water to prevent failure, which could occur if                g the heaters were energized uncovered. The maximum water level in the pressur-                  --
izer ensures that ~this parameter is maintained within the envelope of opera-              E tion assumed in the safety analysis. The maximum water level also ensures                  5 that the RCS is not a hydraulically solid system and that a steam bubble will              2 '"
be provided to accommodate pressure surges during operation. The steam bubble also protects the pressurizer code safety valves against water relief. The                      "2 requirement to verify that on an Engineered Safety Features Actuation test                Y signal concurrent with a loss of offsite power thc pressurizer heaters are                    -
automatically shed from the emergency power sources is to ensure that the                    =
non-Class IE heaters do not reduce the reliability of or overload the emer-                E gency power source. The requirement that a minimum number of pressurizer                    i heaters be OPERABLE enhance the capability to control Reactor Coolant System                j pressure and establish and maintain natural circulation.                                    ..e 4.4.4        STEAM GENERATORS Insat See Applicant's SAR.
E g                                                                                              j
                                                                                                  'm 3/4.4.5        REACTOR COOLANT SYSTEM LEAKAGE g
3/4.4.5.1          LEAXAGE DETECTION SYSTEMS See Applicant's SAR.                                                                          g B 3/4 4-2            Amendment Number 9            -;
February 27, 1984            _5
 
                                                                                    ''s
                                                                                            ~
wi j
          .\
N                                                                                                                            ..
s e                                                                                    .\    .x f#.-
F        %.,
            -+
s s
                                                        ~
s~
                  ~
                            .7
                                                                                                                                .w                                                l l
l                              g
                                '~
g The auxiliary pressurizer spray is required to depressurize the RCS by cool 1
                            /ing  the
                          / The auxiliary  pressurizer steam space to permit the plant to enter shutdown cooling.                                                              ;
pressurizer spray is required during those periods when normal                                                              1
                        ' i pressurizer spray is not available, such as during natural circulation and during ;
the later stages of a nomal RCS cooldown. The auxiliary pressurizer spray also j distributes boron to the pressurizer when normal pressurizer spray is not avail /
                          , able. Use of the auxiliary pressurizer spray is required during the recovery                                                                      /
                          ' from a steam generator tube rupture and a small loss of coolant accident.                                                                      '
                                  ~ .        - . -                    _ , _ _ . .    . .      . , . . . . . _ . . - .
                                                                                                                                          - ._ _ ., _ _, /
                                                                                                                                                                        /j I
 
3/4.4.5.2          OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value. A threshold value of 1 gpm is sufficiently low to ensure early detection of additional leakage.
The 10 gpm IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.
The total steam generator tube leakage limit of 1 gpm, for both steam genera-tors ensures that the dosage contribution from the tube leakage will be limited to less than Part 100 guidelines for infrequent and limiting fault events.
PRESSURE B0UNDARY LEAKAGE of any magnitude may be indicative of an impending failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.
3/4.4.6          CHEMISTRY The limitations on Reactor Coolant System chemistry ensures that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protec-tion to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, far the speci-fied limited time intervals without having a significant effect on the struc-tural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentration to within the Steady State Limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.7          SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour doses at the site boundary will not exceed an appropri-ately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state prfesry-to-second-ary steam generator leakage rate of 1.0 gpm and a concurrent loss of offsite l  electrical power.
Amendment Number 9 February 27, 1984 B 3/4 4-3
 
l The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 micro-curie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accomodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER. Operation with specific activity levels exceeding 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be. restricted to no more than 800 hours per year l
(approximately 10 percent of the unit's yearly operating time) since the l    activity levels allowed by Figure 3.4-1 increase the 2 hour thyroid dose at the site boundary by a factor of up to 20 following a postulated steam genera-tor tube rupture. The reporting of cumulative operating time over 500 hours in any 6 month. consecutive period with greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 will allow sufficient time for Commission evaluation of the circumstances prior to reaching the 800 hour limit.
Reducing Tc to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift). pressure of the atmospheric steam relief valves. The surveil-lance requireO- ments provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.8        PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Chapters 3 and 5 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic opera-tion, at 1W N w waQ5 During heatup, the thermal gradients in the reactor vessel wall produce thennal stresses which vary from compressive at the inner wall to tensile at the outer wall. There thermal induced compressive stresses / tend to alleviate the tensile stresses induced by the internal pressure.J Thererore, a pressur (temperature curve cased en 5teady state conoitions (i.e., no thermal stresses represents a lower bound of all similar curves for finite heatup rates when Ltha 4nnge wm11 nf tha yeg gd -j 3 (,ggigf-ap-the nnverning 10 Cation.
(The heatup analysis also cuven ilie deienninetio~n of pressure-temperatuPirm limitations for the case in which the outer wall of the vessel becomes the (controllinglocation. The thermal nradine,?s attahliched durino heatup produc
    \ tensile stressesQt the outer wall of the vessel $ f.These stresses are additive to the pressure induced tensile stresses which are already present. The the .:1 %duced :tre: e: at the cuter =11 ef the vc :el are tensile and cr dependentonboththerateofheatupandthetimealongtheheatuprampt.
  @  therefc 0, 2 lower hn"ad curse similar te thet de:;ribed Tc, the heeLur ei Gre g % g. g Igwm  g          g                                                Amendment Number 9 February 27, 1984 B 3/4 4-4
 
inna waii wnnet. bc defined. Consequently, foi the ce2cs                                in  hich th: =tcr vc:0:1 5:== the stres: contielling 10::tice each heatu wc['icfthinterest o                          must be analyzed on an individual basis 9r bM M waa. @p rate Mut edw Ge y4t) aaaor vmel wat .Aup*v.a .
The heatup and cooldown limit curves (Figure 3.4-2) are composite curves which were prepared by determining the most ccnservative case, with either the inside or outside wall controlling, for any heatup or cooldown rates of up to 100'F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figure 3,4-2.
                                                                                  %tzd
    @ The reactor vessel materials -for.kast                  0: 9Eylic nt are testd to determine their initial RT            the results of these tests will be shown in the Applicant's SAR. Reach; operation and resultant fast neutron (E > 1 Mev) irradiation will cause an increase in the RT                                Therefore, an adjusted reference temper-ature, based upon the fluence rcaNDde. predicted using Chapter 5, Figure 5.3-6 and the recomendations of Regulatory Guide 1.99 " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials." The heatup and cooldown limit curves shown ci Figure 3.4-2 include predicted adjustments for this shift in RT          at the en of the applicable service period, as well as n the pressure and tem erature sensing
    @ adjustments instruments.          for hIsible errors                  om.d. Ar.sMund '      e.enlem.t The actual shift in RT                        of the vessel material will be established period-ically during operatioNDdy removing and evaluating, in accordance with ASTM E185-73, and Appendix H of 10CFR50, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART      determined from the surveillance capsule is d!'ferent #r~n the calcul-ategDIRT      for the equivalent capsule radiation exposure, l.                                  m a.u, NDT The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criti-cality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.                                                                                                    =
The maximum RT H
for all reactor coolant system pressure-retraining mater-ials with the Nception                  of the reactor pressure vessel, has been assumed to b      F. The Lowest Service Temperature limit line shown on Figure 3.4-2 is ased upon this RT                  since Article NB-2332 (Summer Addenda of.1972) of
    @    Section III of the gME  N Boiler and Pressure Vesse1 Code requires the Lowest Service Temperature to be RT                                100*F for piping, pumps and valves. Below thistemperature,thesystemke+ssuremustbelimitedtoamaximumof20%of N
the system's hydrostatic test pressure of 3125 psia.
The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.
Amendment Number 9 February 27, 1984 B 3/4 4-5
 
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.
The OPERABILITY of two shutdown cooling suction line relief valves, one located in each shutdown cooling suction line, while maintaining the limita-tions imposed on the RCS heatup and cooldown rates, ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to * *F during cooldown or * *F during heatup. Either one of the two SCS suction line relief valves provide adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam                                      "F above the RCS cold leg temper-
  @  atures or fenerator less than or equal to2) the inadvertent safety injecti n actu injecting into a water solid RCS with full ch trging capacity and with letdown isolated. These events are the most limiting energy and mass addition trans-ients, respectively, when the RCS is at low t:emperatures.
3 ksed A                                          (g 3/4.4.9          STRUCTURAL INTEGRITY        -
asam%      a.,L tn                                  & *W,& f The ainspection, progr%ams for t!ie safety-re a [ECodeClass1,2and3
    - components ensure that the structural integrilyfof these components will be maintained at an acceptable level throughout the life of the plant. To-the-extent ;p?F M !e, the in:p:: tier pr gra for th::: cicaent: is in : =pM-w e with Secti:n XI of the ASME Buile, m4 Prc::ur: '.'es3el Cede.
        %g 0    a pb p a lu                accorda 4ct. Mth %N% x1 ss a O mel & dz u A a        e
                                                                    % AsNg'6dits lo cF(R Go.5Ta O3) n etpt erks        speci u.
Acppa, vidttradit. Oddt"d+
h o        gra Gd p (W, p uuua.uf Tb 10 CFR Co.CC a, (3) t) (j).
See Applicant's SAR.
Amendment Number 9 February 27, 1984 8 3/4 4-6
 
INSERT A The limitations imposed on the RCS heatup and cooldown rates are provided to assure low temperature overpressure protection (LTOP) with the two shutdown cooling suction line relief valves operable. ''; : " ''' ' 2 , __ :.. n .
' ' -1 : : ;f f,;; :i- 0 :: 'a """ "; c t 5 0 ' - "- 4 ~; : ';;;_; _c.: .;;'f;. ;;...
At low temperatures with the relief valves aligned to the RCS, it is necessary to restrict heatup and cooldown rates to assure that the P/T limits are riai-exceeded. During worst case transients, RCS peak pressures can reach the relief valve setpoint, se psig, plus accumulation. At temperatures greater than 9e *F during cooldown and 9e *F during heatup, the heatup and cooldown rate limitations assure the limits of Appendix G to 10 CFR 50 will not be exceeoed with overpressure protection provided by the primary safety valves.
Y be.e.      Mppl iceAY '.5 54 R, .                                                .
i l
i
 
REACTOR COOLANT SYSTEM BASES
(] ) 3/4.4.10  REACTOR COOLANT SYSTEM VENTS Reactor Coolant System-vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head ensures the capability exists to perform this function. In addition, Branch Technical Position RSB 5-1 requires that a reactor vessel head vent path is OPERABLE in order to achieve a safe shutdown.
The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path.
The function, capabilities, and' testing requirements of the Reactor Coolant System vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737.
O e
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3/4.5          EMERGENCY CORE COOLING SYSTEM (ECCS)
BASES 3/4.5.1          SAFETY INJECTION TANKS ekisu Sy%
h    The OPERABILITY of each of              safety injection tanks (SIT) ensures that a sufficient volume of borated water will be immediately forced into the reactor through each of the cold legs in the event the RCS pressure falls below the pressure of the safety injection tanks. This initial surge of water into the RCS provides the initial cooling mechanism during large RCS pipe ruptures.
The limits on safety injection tank volume, boron concentre tion and pressure                      bpC ensure that the safety injection tanks will adequately perform their function in the event of a LOCA in MODES 1, 2, 3, or 4.
3                                3 minimum of 25% (1790 ft ) and a maximum of 75% (1927 ft ) on the SIT narrow h      range level instruments of borated rater at normal operating conditions are used in the safety analysis3as the volume in t e SIT's. To allow for instru -
ment accuracy, 28% (1802 ft ) and 72% (1914 ft ) on-the narrow range instru-ment, are specified in the technical specification.
A minimum of 593 psig and a maximum pressure of 632 psig at nomal operating conditions are used in the safety analysis. To allow for instrument accuracy 600 psig minimum and 625 psig maximum are specified in the technical speci 1-cation.                                                                                $ set g  f minimum of(ForMODES3and4operaft}
57%{1361      ) on the SIT wideion  with range    pressurizer-level instrument pressureand a maximum less than 1750 ps of 75% (1927    ft ) on the SIT narrow range level instrument of borated water per tank is reguired when 3 safety injection tanks are operable and a minimum of36%(g08ft)ontheSITwiderangelevelinstrumentsandamaximumof75%
(1927 ft ) on the SIT narrow range instruments is required when 4 safety injectiontanksareoperableatamjnimumpressureof235psig. To allow for instrumegtinaccuracy, 60%(1415 ft ) on the wide range instrument and 72%
(1914 ft ) on the narrow $ nge instrument when -3 safety injection tanks age operable, and 39% (962 ft on the wide range instrument and 72% (1914 ft )
the narrow range instrument when 4 SIT's are operable, are specified in the Technical Specifications. To allow for instrument inaccuracy 254-psig is cified in the Technical RnacMientinnt.
      ' A boron concentration of 4000 ppm minimum and 4400 ppm maximum are used in the safety analysis.                                                                                              I l
The SIT nitrogen vent valves are not single failure proof against depressuriz-ing the SIT's by spurious opening. Therefore, power to the valves is removed
      . while they are closed to ensure the safety analysis assumption of four pressur-ized SIT's.
All of the SIT nitrogen vent valves are required to be operable so that, given a single failure, all four SIT's may still be vented during post LOCA long tem cooling. Venting the SIT's provides for SIT depressurization capability which ensures the timely establishment of shutdown cooling entry conditions as assumed by the safety analysis for small break LOCA's.                        Amendment Number 9 February 27, 1984 i
 
                                  /                                                                                                                                  1730                    X For MODES 3 and 4 operation with pressurizer pressure less than@ psia 1 the Technical Specifications require a minimum of_57% wide range corresponding to 1361 cubic feet and a maximum of 755 narrow range corresponding to 1927 cubic feet of borated water per tank, when three safety injection tanks are operable                                                                                  ,
                              /andaminimumof36Xwiderangecorrespondingto908cubicfeetandamaximum                                                                                                  i
                              ' of 755 narrow range corrisponding to 1927 cubic feet per tank, when four safety'                                                                                \
                            , { injection tanks are opsc ele at a minimum pressure cf 235 psig and a naxime                                                                                        t
                          / pressure instrument of corresponding 625 psig. To allow      to 1415              forcubic      instrument      feet, and inaccuracy, 725 narrow605      wide range    range instrument                        \,
i        corresponding to 1914 cubic feet, when three safety injection tanks are oper-I        able, and 395 wide range instrument corresponding to 962 cubic feet, and 725                                                                                        '!
I        narrow range instrument corresponding to 1914 cubic feet, when four SITS are                                                                                        ,j
                        ,      operable, are specified in the Technical Specifications. To allow for instru-                                                                                        '
                          \ ment inaccuracy 254 psig is specified in the Technical Specifications.                                                                                                                        '
                            \                              _                                                                  . . _ - -
Aminimumof25Ynarrow ange corresponding to 1790 cubic feet and a'                                                                                            '
maximum of 75% narrow range corresponding to 1927 cubic feet of borated water are used in the safety analysis as the volume in the SITS. To allow for                                                                                              i instrument accuracy, 28% narrow range corresponding to 1802 cubic feet and 72%
i                                      narrow range corresponding to 1914 cubic feet, are specified in the Technical Specification.
C '                                                                                              - - -
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  . . _ . . - ,- .--. . . . - .              , ,      _...._._,,,__._r. _ , _ . _ , _ . _ . . . _ , . - . . - . . _ . , ,              ._.m__..,_--,..    ._
                                                                                                                                                                        . - - . _ _ , - , , .                ,.. - - ~
 
r 1
)                                      The SIT isolation valves are not single failure proof; therefore, whenever the valves are open power shall be removed from these valves and the switches I                                      keylocked open. These precautions ensure that the SIT's are available during
                                      .a Limiting' Fault.
,                                      The limits of operation with a safety injection tank inoperable for any reason i                                      except an isolation valve closed minimizes the time exposure of the plant to a
!                                      LOCA event occuring concurrent with failure of an additional safety injection L                                      tank which may result in unacceptable peak cladding temperature. If a closed isolation valve cannot be immediately opened, the full capability of one safety injection tank .is not available and prompt action is required to place the reactor in a MODE where this capability is not reouired.
l                                                                                                                                                    Hones I awl 1              d No06 3 caf*
!                                      3/4.5.'2 and 3/4.5.3                                      ECCS SUBSYSTEMS
                                                                                                                                                      %m            cess h p k M A The OPERABILITY of two sesarate and independent ECCS subsystembth the RCh c w wrature eauai to or a)ove 350*hensures that sufficient emergency core
:                                      cool ng capability will be available in the event of a LOCA, assuming the loss
!:                                      of one subsystem through any single failure consideration. Either subsystem
!                        g              inconjunctionwiththesafetyinjectiontanks-iscapableofsupplyingsuffi-k i"
cient core cooling to limit the per cladding temperature within acceptable                                                                                    "
limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe-downward.                                                          In addition, each ECCS subsystem provides a long term core cooling capability in the recirculation mode during i'
(Ino5NEdoS*EN'[d75[s$hn rrsopsia ed Hobb f<With the RCS tempersture below 350done OPERABLE ECCS' subsystem is accept-able without single failure consideration on the basis fo the stable reactiv-L                                      ity condition of the reactor and the limited core cooling requirements.
The Surveillance Requirements provided te ensure OPERABILITY                                                                              component
                    @                  ensure that at a minimum, the assumptions used in the :::'i...t' analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance
!                                      that proper ECCS flows will be maintained in the event of a LOCA*. Mainten-ance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding
:                                      runout conditions when the system is in its minimum resistance configuration, (2) provide the' proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.
                                        *The following test conditions apply during flow balance tests ensure that the ECCS subsystems are adequately tested.
F                                      1)          The pressurizer pressure is 15 psia.
  ;                                    2)          The mini flow bypass regirculation lines are aligned for injection.
: 3)          For the LPSI system, { add /                                      }3.2gpm(f/                      the 2450 gpm requirement for every foot by whg '                            a6ffference of E) water level above the RWT
                                        $          RAS                    setpoint level {be.ess    l    than) the difference of RCS water level above
  ;                                                the6pidgegcenterl February 27, 1984 B 3/4 5-2 4
i vi<.,e-  r -my-      rm-w,,.+    ,,,,--,,..-.,m_.,w.-.4....,..-        --m., .,,.--.y      ,,..-,.,..+---e-,-r-e-.--,....-w                  -..,m----- --.,w-sm-.-    ,,..-,wm-.*--    -
 
The tem " minimum bypass recirculation flow", as used in paragraphs 4.5.2.e.3 and 4.5.2.f. refers to that flow directed back to the RWT from the ECCS pumps for pump protection. Testing of the ECCS pumps under the condition of minimum bypass recirculation flow in paragraph 4.5.2.f verifies that the perfomance of the ECCS pumps supports the safety analysis minimum RCS pressure assumption at zero delivery to the RCS.
3/4.5.4        REFUELING WATER TANK The OPERABILITY of the Refueling Water Tank (RWT) as part of the ECCS ensures that a sufficient supply of borate 1 water is available for injection by the ECCS in the event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that (1) sufficient water plus 10 percent margin is available to permit 20 minutes of Engineered Safeguard Features Pump operation, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWT and the RCS water volumes with all control rods inserted except for the most reactive control element assembly.. The limit on the RWT solution temperature ensures that the assumptions used in the LOCA analysis remain valid., These assumptions are consistent with the LOCA analysis,
  &                'NO
                      %.      m    na  & VM AEnit YA m w g m&u w absu)ss.t.L.
fo*Hm          Gt.ch y mt h .a                        g u
i I
a Amendment Number 9      l February 27, 1984        j B 3/4 5-3
 
e THIS PAGE INTENTIONALLY BLANK l
 
I i
3/4.6        CONTAINMENT SYSTEMS BASES 3/4.6.1        PRIMARY CONTAINMENT See Applicant's SAR.
3/4.6.2        DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1        CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that containment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and re:ultant lower containment leakage rate are consistent with the assumptions used in the safety analyses.
The containment spray system and the containment cooling system are redundant to each other in providing post accident cooling of the containment atmosphere.
However, the containment spray system also provides a mechanism for removing iodine from the containment atmosphere and therefore the time requirements for restoring an inoperable spray system to OPERABLE status have been maintained consistent with that assigned other inoperable ESF equipment.
3/4.6.2.2          IODINE REMOVAL SYSTEM The OPERABILITY of the Iodine Removal system ensures that sufficient N H is added tp the containment spray in the event of a LOCA. Thelimitson$
volumeandconcentrationensureadequatechemicalavailabletoremoveibine from the containment atmosphere following a LOCA.
3/4.6.2.3          CONTAINMENT COOLING SYSTEM See Applicant's SAR.
3/4.6.3        CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the contain-ment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressuriza-tion of the containment and is consistent with the requirements of GDC54 through GDC57 of Appendix A to 10CFR50. Containment isolation within the time liciits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA, Unless otherwise stated no containment purging should be done while operating at 100% of RATED THERMAL POWER. This is necessary as it is one of the assumptions used in the LOCA Safety analyses.
3/4.6.4        COMBUSTIBLE GAS CONTROL See Applicant's SAR.
3/4.6.5        IODINE CLEANUP SYSTEMS
} See Applicant's SAR.
Amendment Number 9 February 27. 1984 8 3/4 6-1 c, , ,
 
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3/4.7              PLANT SYSTEMS BASES 3/4.7.1              TURBINE CYCLE 3/4.7.1.1              SAFETY VALVES
                                                          @5ffb b a- The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% (1382 psig) of its design pressure of 1255 psig during the most severe anticipated system opera-tional transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).
The specified valve lift settings and relieving capacities are in acccrdance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code, 1974 Edition. The total relieving capacity for all valves on all of the steam lines is 18,660,000 lbs/hr which is 104.6 percent of the total secondary steam flow of 17,830,000 lbs/hr at 100% RATED THERMAL POWER plus 2% uncertain-ity.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION recuirements on the basis of the reduc-tion in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:
For two loop, four pump operation SP = (        ) x 104.6 where:
SP = reducen allowable steady state power level in percent of RATED THERMAL POWER. This is a ratio of the available relieving capacity over the total steam flow at rated power.
10 = total number of secondary safety valves for one steam generator.
N = the number of inoperable secondary safety valves on the steam generator with the greater number of inoperable valves.                                I 104.6 = the ratio of the total relieving capacity of all twenty (20) second-ary safety valves 18,660,000 lb/hr at 1382 psig - 50 psi pressure drop to the inlet of the safety valves) over the secondary steam                        i flow at 100% Rated Thennal Load plus 2% uncertainity (17,830,000                        l lbs/hri.                                                                                '
3/4.7.1.2              EMERGENCY FEEDWATER SYSTEM See Applicant's SAR.                                                                  '
B 3/4 7-1                  Amendment Number 9 February 27, 1984
 
t 0
    %d h k (Rof 1
system The  0. :"f 51LJi pressure        ef tM main step sgfety valves (MSSVs) limit secogary110%(1397$sid)o to within during the most severe anticipated operational transient. For design purposes the~ valves are (zedtopassaminimumof102%oftheRATEDTHERMALPOWERat 102% of design          . The adequacy of this relieving capacity is demonstrated by maintaining the Reactor Coolant System pressure below NRC acceptance criteria (120% of design pressure for large feedwater line breaks, CEA ejection and 110%
of design pres:ure for all overpressurization events).
The specified valve lift settings and relieving capacities are in accord-ance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. 1"7' Editi = *a'!"dia; t%        S'--- 1975 AddaaM.
The total relieving capacity for all twenty MSSVs at 110% of sys            design pressure (adjusted for 50 psi pressure drop to valves inlet) is              x 108 lbm/hr. This capacity is less than the total rated capacitygas the MSSVs are operating at an inlet pres-sure below rated conditions. At thhse same secondary pressure conditions, the totalsteamflowat102%(2%uncerta%y)of3817MWt(RATEDTHERMALPOWERplus 17 MWt pump heat input) is 17.83 x 10s ibs/hr. The ratio of this total steam flow to the total capacity is 1          .
N cf t9 e tob b/h,-pued FSA R%M.14.13 .1.
STARTUP and/or POWER OPERATION is allowable with MSSVs inoperable if the maximum allowable power level is reduced to a value equal to the product of the ratio of the number of MSSVs available per steam generator to the total number of MSSVs per steam generator with the ratio of the total steam flow to available relieving capacity.
: 4. 6 Allowable Power Level = (10-N)10    *1 The ceiling on the variable over power reactor trip is also reduced to an amount over the allowable power level equal to the BAND given for this trip in Table 2.2-1.
SP = Allowable Power Level + 9.8 where:
SP    =    reduced reactor trip setpoint in percent of RATED THERMAL POWER. This is a ratio of the available relieving capacity over the total steam flow at rated power.
i l
 
c _
          $sralA (e**d) 10    =            total number of secondary safety valves for one steam generator.
N    =            number of inoperable main steam safety valves on the steam generator with the greater number of inoperable valves.
d, 6 109ta =              ratio of main steam safety valve relieving capacity of 110%
steam generator design pressure to calculated steam flow rate at 100% plant power + 2% uncertainty (see above text) 9.8 =                BAND between the maximum thermal power and the variable over-power trip setpoint ceiling 4
    , -          .m._.  .__ ._
 
I l
l A          !
3/4.7.1.3        CONDENSATE STORAGE TANK S ,
[The OPERABILITY of the condensate storage tank with the minimum water volumde ensures that sufficient water is available to maintain the RCS at H0T STANDBY    '
00,      conditions for 8 hours with steam discharge to atmosphere with concurrent total loss of off-site power. TheOPERABILITYofthecondensatestoragetanN with the minimum water volume also ensures that sufficient water is available to maintain the RCS at H0T STANDBY conditions for 4 hours followed by an i orderly cooldown to the shutdown cooling entry temperature (350'F) with (concurrenttotallossofoff_-sitepower.
Alan.u. c Aciturd The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electric-al power. These values are consistent with the assumption used in the safety analyses.
3/4.7.1.5        MAIN STEAM ISOLATION VALVES The OPERABILITY of the main steam isolation valves nsures that no more than one steam generator will blowdown in the event of a            steam line rupture.
This restriction is required to (1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves        '
@        within the closure times of the surveillance requirements are consistent with the assumptions used in the safety analyses.
    -> Tssert B 3/4.7.2        STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses'in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitation to * 'F and
* psig
    . are based on a steam generator RTNOT      f*  F and are sufficient to prevent brittle fracture.
3/4.7.3        COMPONENT COOLING WATER SYSTEM See Applicant's SAR.
3/4.7.4        SERVICE WATER SYSTEM See Applicant's SAR.
ULDH A t'l HEAF GINV-3/4.7.5        EMERGENCY COOLING FONO See Applicant's SAR.
3/4.7.6        FLOOD PROTECTION See Applicant's SAR.
* See Applicant's SAR.                                        Amendment Number 9 February 27, 1984 8 3/4 7-2
 
n ,--+ -      .-                                                                            -
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERA 8ILITY of the c'o2 ens ~ite stifrige tank ensures-that-a-minists f water volume of 300,000 gallons is available to maintain the Reactor Coolan b System at HOT STAN08Y for 8 hours followed by an orderly cooldown to the shut- \
down cooling entry (350*F) temperature with concurrent total loss of-site power, and also ensures that sufficient water is available to maintain the          ,
RCS at HOT STANOSY conditions for 8 hours with steam discharge to atmosphere concurrent with total loss-of-offsite power. The contained water volume limit      /
includes an allowance for water not usable Locause of tank discharge line loca-tion or other physical characteristics.
 
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y4                          4as    m . eme.              y N*'-
* g                                                                    w      e-e    -                            e                                        W 6hN e als-@        a+-4  e      ae
                                                                                                                        ,g,      ,
gem                                                                                                          m66                6@ m              e9 =>          WM    '              6                          -
a        w-@
_ mm -                          +,.        .my-                                                    w-M@-            w eh                  e                ~ght      o        ae e
 
w 3/4.7.7        CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM See Applicant's SAR.
3/4.7.8        ECCS PUMP ROOM EXHAUST AIR CLEANUP SYSTEM See Applicant's SAR.
3/4.7.9        SNUBBERS
        @ The t,W.e.lic :: it:r: in:1
          . Tele 3.7-3e Of S;;ti : 3/4.7.S d H thia the :::;; cf CESS *" :r: id::ti'i d en CLll Th: hydr;;lis snubbers are required to be OPERABLE to (nsure that the structur-
        @ al integrity of the reactor coolant system and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Lert A The hydraulic snubbers included in the CESSAR scope are provided with se.ls fabricated from materials which have been demonstrated compatible with their operating environment. Inspectifon program intervals and acceptance criteria for the CESSAR scope snubbers will be included in the Applicant's SAR, consis-tent with the program set-up by the Applicant for other snubbers in the plant.
3/4.7.10        SEALED SOURCE CONTAMINATION See Applicant's'SAR.
3/4.7.11        FIRE SUPPRESSION SYSTEMS See Applicant's SAR.
:P 3/4.7.12        FIRE BARRIER PENETRATIONS See Applicant's SAR.
3/4.7.13        AREA TEMPERATURE MONITORING See Applicant's SAR.
3/4:7.14        SHUTDOWN COOLING The OPERABILITY of two separate and independent shutdown cooling subsystems                  j i        ensures that the capability of initiating shutdown cooling in the event of an                1 accident exists even assuming the most limiting single failure occurs. The safety analysis assumes that shutdown cooling can be initiated when conditions permit.
The limits of operation with one shutdown cooling subsystem inoperable for any reason minimize the time exposure of the plant to an accident event occuring
-l        concurrent with the failure of a component on the other shutdown cooling
:        subsystem.
Amendment Number 9 February 27, 1984 8 3/4 7-3                                              l
 
                                                                                          's Snubbers excluded from this inspection program are those installed on nonsafety-related          ,
systems and then only if their failure or failure of the system on which they          '
are installed, would have no adverse effect on any safety-related system.            '
N__..___.
t
 
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1 6
 
3/4.8'        ELECTRICAL POWER SYSTEMS BASES 3/4.8.1            A.C. SOURCES See Applicant's SAR.
3/4.8.2            D.C.' SOURCES See Applicant's SAR.
3/4.8.3            ONSITE POWER DISTRIBUTION See Applicant's SAR.
  .          3/4.8.4            ELECTRICAL EQUIPMENT PROTECTIVE DEVICES See Applicant's SAR.
I t
Amendment Number 9    !
February 27, 1984 8 3/4 8-1
 
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3/4.9                        REFUELING OPERATIONS BASES 3/4.9.1-                      BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that: (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a uniform BORON CONCENTRATION is maintained for reactivity control in the water volume having direct access to the reactor vessel. These limitations are consistent
        @                          with the initial conditions assumed for the boron dilution incident in the g ...;i nt analyses.                                                                                            includes a 15 delta k/k conservative allowance              Theforvalue            of 0.95Similar1 uncertainties.      or less the                      forboron          K) f concentration value of 2150 ppm or greater also includes a conservative uncertainty allowance of 50 ppm boron.
3/4.9.2                        INSTRUMENTATION The OPERA 8ILITY of the source range neutron flux monitors ensures that redun-
+
dant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3                        DECAY TIME The minimum requirement for reactor subcriticality prior to movement of
,                                  irradiated fuel assemblies in the reactor pressure vessel ensures that suffi-cient fissiontime    products. has elapsed      to allow This decay        timethe  radioactivewith is consistent          decay      the of                  the short lived,d assumptions                    use in l                                  the safety analyses.
3/4.9.4                        CONTAINMENT BUILDING PENETRATIONS See Applicant's SAR.
!                                  3/4.9.5                        C0mVNICATIONS See Applicant's SAR.
3/4.9.6                        REFUELING MACHINE OPERABILITY See Applicant's SAR.
3/4.9.7                        CRANE TRAVEL-SPENT FUEL POOL BUILDING See Applicant's SAR.
3/4.9.8                        SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling loop be in operation and t
circulating reactor coolant at a flowrate of > 4000 gpm ensures that (1)_                                                                                                      l sufficient cooling capacity is available to remove decay heat and maintain the i                              water in the reactor pressure vessel below 135'F as required during the l                                  REFUELINGMODE,(2)sufficientcoolantcirculationismaintainedthroughthe
,                                                                                                                                                                      Amendment Number 9 i                                                                                                                                                                      February 27, 1984 8 3/4 9-1 1
        , - - , . - . , + . . , .    . - - . _ - _ . - - _ _ -                . _ , _          _ _ _                      . . . _ - - , , . , _ _ _ . . . . _ , , - - - .              . . . - . - -
 
reactor core to minimize the effects of a boron dilution incident and prevent boron stratification, and (3) that the AT across the core will be maintained less than 75'F.
Without a shutdown cooling train in operation steam may be generated; there-fore, the containment should be sealed off to prevent escape of any radioacti-vity, and any operations that would cause an increase in decay heat should be secured.
ht The redirement to have two shutdown cooling loops OPERABLE when there is less than 230 of water above the reactor pressure vessel flange, ensures that a
    @ single failure of the operating shutdown cooling loop will not result in a complete loss ofgcay heat removal capability. With the reactor vessel head removed and 23Tof water above the reactor pressure vessel flange, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling loop, adequate time is provided to initiate emer-gency procedures to cool the core.
3/4.9.9              CONTAINMENT PURGE VALVE ISOLATION SYSTEM See Applicant's SAR.
3/4.9.10 and 3/4.9.11                WATER LEVEL - REACTOR VESSEL AND WATER LEVEL -
STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from
    @      the rupture of an irradiated fuel assembly. The min _imum water depth is annsistent with the assumptions of the g                          t analys1 3/4.9.12 p a max 4mu4x. h M STORAGE POOL AIR CLEANUP SYSTEM
!          See Applicant's SAR.                                                      F"              3    "I            b7 Amendment Number 9 February 27, 1984 B 3/4 9-2 m _ ___.              . _ . . _  _ _ _ _      _ _ . _ _ _ _ _ _ _ _ _            _ _ _ _ _ _ _          __  _____    .
 
E E
I 3/4.10        SPECIAL TEST EXCEPTIONS f            BASES p                                                                                                                            -
3/4.10.1          SHUTDOWN MARGIN Y
This special test exception provides that a minimum amount of CEA worth is l-innediately available for reactivity control when tests are performed for CELs worth measurement. This special test exception is required to permit the
  ;        periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel ournup or fuel cycling (,perations. Although
  '          testing will be initiated fron MODE 2, temporary entry into MODE 3 is neces-sary during some CEA worth measurements. A reasonable recove                  time is e
V'      O available for return to MODE 2 in order to continue Mb                          %rg.
ii          3/4.10.2          MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DI5TRIBUTION LIMITS i
j          This special test exception permits individual CEAs to be positioned outside of their nomal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure CEA worth. 2) determine the reactor stability index and damping factor under xenon oscillation conditions.
: 3) determine power distributions from non-nonnal CEA configurations. 4) meas-j        ure rod shadowing factors, and 5) measure temperature and power coefficients.
g
-r This special test exception permits the MTC to exceed the limits of Specifica-f          tion 3.1.1.3 during the performance of physics tests.
3/4.10.3          REACTOR COOLANT LOOPS i
[        This special test exception permits reactor criticality with less than four p
reactor coolant pumps in operation and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.
3/4.10.4          CEA POSITION REGULATING CEA INSERTION LIMITS AND ReAcret. coeuwt coto
,      @  This special test exception pennits the CEA's to be positioned beyond the LE6 TEMPEAMuRE i          insertion limits, during PHYSICS TESTS required to determine the isothermal E          temperature coef engajdpgegf                e                          ,
    @        4.10.) 6      SAFETY INJECTION TANKS This special test exception permits testing the low pressure safety injection l        system check valves. The pressure in the injection header must be reduced E
below the check thevalves.
head Theof thesafety  lowinjection pressuretankinjection  p(ump)in SIT isolation  valve mustorder                    beto get flow through 2
closed in order to accomplish this. The SIT isolation valve is still capable
;          of automatic operation in the event of a SIAS, therefore, system capability
_          should not be affected.
c
:                                                                                    Amendment Number 9 February 27. 1984 B 3/410-1 L
~
                                                                          ..  .. .      .    .                    .                    l
 
ht N w ,                          ANo PREssue6 3/4.10.5 ' MINIMUM TEMPERATURE.FOR CRITICALITY Thih special test exception permits reactor criticality at low THERMAL POWER levels with T below the minimum critical temperature during PHYSICS TESTSwhicharereq6TNdtoverifythelowtemperaturephysicspredictionsand to ensure the adequacy of design codes for reduced temperature conditions.
s.
perform The Low                the following Power    tests:Physics                          Testing Program                                      2  at low temperature (3
: 1.                Biological shielding survey test                                                                              aes. a pr2um Q
: 2.                Isothermal temperature coefficient tests                                                                      g bg
: 3.                CEA group tests
: 4.                Soron garth tests s
: 5.              ' Critical. configuration boron concentration
(,'                -
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3/4.10.h        SAFETY INJECTION TANK PRESSURE 3Ao This special test exception allows th      erformance of PHYSICS TESTS at low pressure / low temperature (600 psig,      F) conditions which are required to verify the low temperature physics predictions and to ensure the adequacy of design codes for reduced temperature conditions.
l l
1 l
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!                                                                    Amendment Number 9 February 27, 1984 8 3/4 10-2 l
 
        @    TECHNICAL SPECIFICATIONS SECTION 5.0
                  , DESIGN FEATURES 1
l l
l Amendment Number 9 l                                                        February 27, 1984 l
t..  ..
                                                                                )
 
i THIS PAGE INTENTIONALLY BLANK.
m +
_    .. _a
 
5.0            DESIGN FEATURES 5.1            SITE                                                                                                        ,
See Applicant's SAR.
5.2            CONTAINMENT See Applicant's SAR.
5.3            REACTOR CORE 5.3.1            FUEL ASSEMBLIES The reactor core shall contain 241 fuel assemblies with each fuel assembly containing a maximum of 236 fuel rods or burnable poison rods clad with Zircaloy-4. Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight of approximately 1900 grams uranium.
Each burnable poison rod shall have a nominal active poison length of 136 inches. The initial core loading shall have a maximum enrichment of 3.30 weight percent U-235. Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.0 weight percent U-235.
5.3.2            CONTROL ELEMENT ASSEMBLIES The reactor core shall contain 76 full length and 13 part-length control element assemblies.
5.4            REACTOR COOLANT SYSTEM 5.4.1            DESIGN PRESSURE AND TEMPERATURE The reactor coolant system is designed and shall be maintained in accordance with the code requirements specified in Section 5.2 of the FSAR with allowance for normal degradation pursuant of the applicable Surveillance Requirements, and is also designed as follows:
: a.        For a pressure of 2500 psia, and
: b.        For a temperature of 650'F, except for the pressurizer which is .
700*F.
5.4.2            VOLUME The total water and steam volume of the reactor coolant system is 13,900 +
300/-0 cubic feet at a nominal T,yg of 593*F.
5.5            METEOROLOGICAL TOWER LOCATION See Applicant's SAR.
5.6            FUEL STORAGE Amendment Number 9 i    See Applicant's SAR.                                                                February 27, 1984                        l 5-1 L
 
5.7      COMPONENT CYCLIC OR TRANSIENT LIMITS Q f II  The components identified in Table 5.7-1 are designed and shall be maintained
      ^ within the cyclic or transient limits of Tabic (5.7-1, aud C.7  ,1.              l l
Amendment Number 9 February 27, 1984      1 5-2 l
1
 
TABLE 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS I
Cyclic or                                        Design Cycle Component                  Transient Limit                                    or Transient                            .
48*4 4a%44,  %.
:                  Reactor Coolant System    500 system heatup and cooldown          Heatup cycle -            f          F to  65'F; i                                            cycles at rates $ 100* F/hr.            cooldown cycle -              fr  565'F to        F.
                                                                                                            .=    ,W 500 pressurizer heatup and              Heatup cyc e - Pressurizer temperature from cooldown cycles at rates <  -
7o : 2^^ F t          653*F; cooldown cycle - Pressurizer 200*F/hr.                              temperature                            F.
romb 653*F to Wfo i                                            10 hydrostatic testing cycles.          RCS pressurized to 3125 psig with RCS temperature j                                                                                    between 100*F and 400*F.
I 480 reactor trip cycles, turbine        Includes combinations of reactor trips due to m                                    trip cycles, and. loss of re-          operator errors, equipment malfunctions, and  _
a                                    actor coolant flow.                    total loss of reactor coolant flow.
200 seismic stress cycles.              Subjection to a seismic event equal to one half            f the design basis earthquake (DBE).
                                  \          l complete loss of secondary pressure cycle.
Loss of secondary pressure from either' steam generator due to a complete double-ended break of a steam generator steam or feedwater nozzle.
15,000 power change cycles                      s from 15% to 100% ful'1 load, at a rate %
5% per minute, either increasing or decreasing.        '
pg                                                                      (30,000 cycles total) l            erto i
jh                              106 step changes of 100 psi and        Pressure variations between the pressurizer-qg                              10*F (20*F for surge line)              pressure setpoint for backup heater actuation j                                                                      and spray valve opening. Temperature varia-N :=                                                                  tions due to CEA controller; 2000 step chan
;            -5                                                                      of 10% full power.
j            M Q        @                    DCL) %*p % len ha!-                    Oap. tut Pri                CysOwn oA a- fawo Q st c p u h. f                        of x%o si          al a temptuAL $w ^ lWY it 4 66*
 
TABLE 5.7-1 (Cont'd)
COMPONENT CYCLIC OR TRANSIENT LIMITS                                                                                            ,
Cyclic or                                  Design Cycle Component                  Transient Limit                              or Transient Pressurizer Spray Nozzle                              7 Main Spray (4 pumpMng)-
Main Spray (Les    han 4 pumps operating) -
{ Unlimited number of cycles          with fluid AT,        F.
                              .L                                                        on Auxiliary spray at various initial fluid temperatures with fluid AT, N F.
                                                                                                      .2 Main spray (less than 4 pumps operating)
Calculate usage factor per          with fluid AT* > +* F.
Table 5.7-2 Auxiliary spray with fluid AT a
                                                                                                        >                              F.
m                                                                                                        goo 1.
AT,  =  The difference in temperature between the pressurizer and main spray water as adjusted by the instrument                                                      -
correction factor.**
AT a
            =  The difference in temperature between the pressurizer and Auxiliary spray water as adjusted by the instrument correction factor.**
* See Applicant's SAR
(( **    For instrument correction factors, see Applicant's SAR.
2 "$
N
 
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  @    if av 1
Amendment Number 9          ;
February 27, 1984          '
5-5
            .---,r.    - - . - - .-              -
 
1 THIS PAGE INTENTIONALLY BLANK.
9
  , - ,      -  n
 
g ___.)  g h_HNICALSPECIFICATIONS SECTION 6.0 ADMINISTRATIVE CONTROLS I
Amendment Number 9 February 27, 1984 1
 
THIS PAGE INTENTIONALLY BLANK, i
 
6.0        ADMINISTRATIVE CONTROLS 6.1        RESPONSIBILITY See Applicant's SAR.
6.2        ORGANIZATION See Applicant's SAR.
6.3        UNIT STAFF QUALIFICATION See Applicant's SAR.
6.4        TRAINING See Applicant's SAR.
6.5        REVIEW AND AUDIT See Applicant's SAR.
EUGNT h  6.6        REPORTABLE-066tfRRENEE ACTION See Applicant's SAR.
6.7        SAFETY LIMIT VIOLATION See, Applicant's SAR.
6.8        PROCEDURES See Applicant's SAR.
6.9        REPORTING RE0VIREMENTS See Applicant's SAR.
6.10        RECORD RETENTION See Applicant's SAR.
6.11        RADIATION PROTECTION PROGRAM See Applicant's SAR.
l 6.12        HIGH RADIATION AREA See Applicant's SAR.                                            -
6-1 Amendment Number 9 February 27, 1984
 
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>    r ENCLOSURE E TO LD-85-024
                                                                        ,t.
Palo' Verde Technical Specifications-With Differences From System 80
                                        -' Technical Specifications Identified
                                                                                                -4 5
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                                                      .          1
                                                    ~
CONTROLLED BY USER CliAff62. !b 5'/516M SC - fsAE. C6sSAR.
T6LH A IcAL      SPEci FtcP.ficNS l
                                                            .l w  .                    ;
TECHNICAL SPECIFICATIONS                        i N
PALO VERDE NUCLEAR GENERATING STATION UNIT NO. 1                  \,
DOCKET NO. 50-528 DEC    1984
                                        /-                  <
                                                        *4 P A AITD AI I Er% DV I IP Ert
 
I CONTROLLED BY USER                                                                                                !i, 1
INDEX DEFINITIONS SECTION                                                                                                              PAGE
: 1. 0 DEFINITIONS 1.1    ACTI0N............................'..........................                                                    1-1
: 1. 2  AX I A L S HAP E I N0 EX. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1
: 1. 3    AZIMUTHAL POWER TILT........................................                                                    1-1
: 1. 4    CHANN EL CA LI B RATIO N . . . . .' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .      1-1
: 1. 5    CHANNEL CHECK...............................................                                                    1-1
: 1. 6    CHANNEL FUNCTIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                1-2 1.7    CONTAINMENT    INTEGRITY.......................................                                                  1-2 1.8    CONTROLLED LEAKAGE..........................................                                                    1-2 1.9    CORE ALTERATION.............................................                                                    1-2 1.10 DOSE EQUIVALENT I-131.......................................                                                        1-3
: 1. n I-AVERAGE DISINTEGRhTION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . .                          1-3 1.12 ENGINEERED SAFETY FEATURES RESPONSE TIME....................                                                      1-3 1.13 FREQUENCY N0TATION..........................................                                                      1-3 fd                                    SYSTNMt....................................                                              1-3
  @e$) 1.15    G IDENTIFIED T4 GASEOUS              RADWASTE LEAKAGE..........................................                                                1-3
          '1716 MEMBER (5) 0F THE PUBLIC h .............................                                                          1-4 (gers) 1.17 0FFSITE 00SE CALCULATION MANUAL . . . . . . . . . . . . . . . . . . . . . . . . . . . .                              1-4 1.18 OPERABLE - OPERABILITY......................................                                                      1-4 1.19 O P E RATIONA L MOD E - M0 D E. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .          1-4 1.20 PHYSICS TESTS...............................................                                                      1-4 1.21 PLANAR RADIAL PEAKING FACTOR - F                                                                                  1-4
                .                                                      xy..........................
1.22 PRESSURE BOUNDARY LEAKAGE...................................                                                      1-4 8dP    '1.23 PROCE35 CONTROI 7RO'GRXD. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .                    1-5 NM t
1.24 FURGE - PURGING.....      3..'....................................                                              1-5 1.25 RATED THERMAL P0WER.........................................                                                      1-5          -
1.26 REACTOR TRIP SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . .                          1-5
                                                                                                                                -  1''
  @gs) c
                > 1dk"VM1AESr 1.28  SHUTDOWN MARGIN.............................................
1-5 gce    _.29 SITE B00NDARR..............................................                                                  1-6 O        1.30    S0FTWARE....................................................                                                  1-6 PALO VERDE - UNIT 1                  -
I f, @%i sol 1:.          .        M. . . W ; I .-T 5 D                          2  .
L
 
CONTROLLED BY USER INDEX DEFINITIONS SECTION                                                                      PACE C31      SOLIDIFICATION..............................................      1-6 6 ) ( Q URCE CHECX            3    ,2 ............................................ 1-6 (6      1.33 STAGGERED TEST BASIS........................................            1-6 1.34 THERMAL P0WER...............................................            1-6 1.35 UNIDENTIFIED LEAKAGE........................................            1-6
                            ~                      ~
[1.36 UNRESTRR TED AREA............. 2 x .......................                1-6 a                                                    s 1.37                                                                          1-7 60Y              VENTILATIONEXHAUSTTREATMENTSYSTEM..').....................
gtD 1.38 VENTING..............                          d...................... 1-7 (g                                        .....  . _ .
PALO VERDE - UNIT 1                            II CONTRoli FD RY I KFR
 
l i
CONTROLLED BY USER INDEX 1
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS                                                                              1 SECTION                                                                                                            PAGE 2.1    SAFETY LIMITS 2.1.1      REACTOR C0RE.............................................                                                2-1 2.1.1.1    DNBR.....................................................                                                2-1 2.1.1.2    PEAK LINEAR HEAT RATE....................................                                                2-1 2.1.2      REACTOR COOLANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . .                      2-1 2.2 LIMITING SAFETY SYSTEM SETTINGS 2.2.1    REACTOR TRIP SETP0lNTS.....................................                                                2-2 2.2.2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS. . . . . . . . . . .                                        2-2 BASES SECTION                                                                                                            PAGE 2.1    SAFETY LIMITS 2.1.1    REACTOR C0RE...............................................                                              B 2-1 2.1.2 REACTOR COO LANT SYSTEM PRESSURE. . . . . . . . . . . . . . . . . . . . . . . . . . . .                      B 2-2
: 2. 2 LIMITING SAFETY SYSTEM SETTINGS
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2.2.1    R EACTO R TR I P S ETP0 I NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 2-2 2.2.2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS...........                                                  B 2-7 PALO VERDE - UNIT 1                                    III CONT 12OII Fn RY I KFD
 
l CONTROLLED BY USER INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                                                  PAGE        !
3/4.0  APPLICABILITY..............................................                                  3/4 0-1          l 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1    BORATION CONTROL                                                                                          !
SHUTDOWN MARGIN - Tcold  >210*F........................                                3/4 1-1          ,
SHUTDOWN MARGIN - Tcold  <210*F........................                                3/4 1-3          '
MODERATOR TEMPERATURE COEFFICIENT.....................                                  3/4 1-4 MINIMUM TEMPERATURE FOR CRITICALITY...................                                  3/4 1-6 3/4.1.2    BORATION SYSTEMS FLOW PATHS - SHUTD0WN.................................                                  3/4 1-7 FLOW PATHS - 0PERATING................................                                  3/4 1-8 CHARGING PUMPS - SHUTD0WN.............................                                  3/4 1-9 CHARGING PUMPS - 0PERATING............................                                  3/4 1-10 BORATED WATER SOURCES - SHUTD0WN......................                                  3/4 1-11 BORATED WATER SOURCES - OPERATING.....................                                  3/4 1-13 BORON DI LUTION A LA RMS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-14    1 3/4.1.3    MOVABLE CONTROL ASSEMBLIES CEA P0SITION..........................................                                  3/4 1-21 POSITION INDICATOR CHANNELS - OPERATING...............                                  3/4 1-25 POSITION INDICATOR CHANNELS -      SHUTDOWN................                            3/4 1-26 CEA DROP TIME.........................................                                  3/4 1-27 SHUTDOWN CEA INSERTION LIMIT..........................                                  3/4 1-28 REGULATING CEA INSERTION LIMITS.......................                                  3/4 1-29 PALO VERDE - UNIT 1                      IV u
CONTDoll En RY IICED
 
I CONTROLLED BY USER INDEX                                            .
i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1    LINEAR HEAT  RATE........................................      3/4 2-1 3/4.2.2    PLANAR RADIAL PEAKING FACT 0RS........................... 3/4 2-2 3/4.2.3    AZIMUTHAL POWER TILT....................................      3/4 2-3 3/4.2.4    DNBR MARGIN.............................................      3/4 2-5                  l
            .f4.2.5 RCS FLOW RATE...........................................      3/4 2-8 3/4.2.6    REACTOR COOLANT COLD LEG TEMPERATURE.................... 3/4 2-9                        j 3/4.2.7    AXIAL SHAPE INDEX.......................................        3/4 2-11                l 3/4.2.8    PRESSURIZER PRESSURE.................................... 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1    REACTOR PROTECTIVE INSTRUMENTATION......................      3/4 3-1                  l 3/4.3.2    ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION....................................... 3/4 3-17 3/4.3.3    MONITORIC INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION.................      3/4 3-37 INCORE  DETECTORS..................................... 3/4 3-41 SEISMIC INSTRUMENTATION.............................. 3/4 3-42 METEOROLOGICAL INSTRUMENTATION....................... 3/4 3-45 h      REMOTE SHUTDOWN SYSTEM.3tr.4fe.M.C.M %.E.N............ 3/4 3-48 2 POST-ACCIDENT.
C.HtC W E W K % NMONITORING WM      pSTRUMENTATION............. 3/4 3-57
              )
(                  FIRE DETECTION INSTRUMENTATION....................... 3/4 3-61 LOOSE-PART DETECTION __ INSTRUMENTATION................. 3/4 3-69 o9 R54- 27 /NRADI0
                          -IN ACTIVE GASEOUS EFFLUENT MONITORINGi                    3/4 3-71 b      --> > fu aa:STRUMENTATION.
u mswSTcoreew                . . . . . . . . . . . . .&. . . .2 ............ i L 3'/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER  OPERATION.............................      3/4 4-1 HOT STAND 8Y.............................................      3/4 4-2                  J HOT SHUTD0WN............................................      3/4 4-3                  I 1
COLD SHUTDOWN - LOOPS FILLE 0............................      3/4 4-5
                                                                                                                ~
COLD SHUTDOWN - LOOPS NOT FILLED........................      3/4 4-6 I
1
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PALO VERDE - UNIT 1                        V CONTROII FD RY IKFR                                                        .
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CONTROLLED BY USER INDEX                                              I LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                            PAGE 3/4.4.2    SAFETY VALVES SHUTD0WN.............................................            3/4 4-7 0PERATING............................................            3/4 4-8 3/4.4.3    PRESSURIZER PRESSURIZER..........................................            3/4 4-9 AUXILIARY SPRAY......................................            3/4 4-10 3/4.4.4    STEAM GENERATORS........................................            3/4 4-11 3/4.4.5    REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................            3/4 4-18 OPERATIONAL LEAKAGE..................................            3/4 4-19 3/4.4.6    CHEMISTRY...............................................            3/4 4-22 3/4.4.7    SPECIFIC ACTIVITY.......................................            3/4 4-25 3/4.4.8    PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM...............................            3/4 4-29 PRESSURIZER HEATUP/COOLDOWN LIMITS...................            3/4 4-32 OVERPRESSURE PROTECTION SYSTEMS......................            3/4 4-33 3/4.4.9    STRUCTURAL INTEGRITY....................................            3/4 4-34 3/4.4.10 REACTOR COOLANT SYSTEM VENTS............................                3/4 4-35 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1                                                                          3/4 5-1 SAFETY INJECTION TANKS.........d.M                            4....
                                                            ........ 5(f duV.
r43 3/4.5.2    ECCS SUBSYSTEMS -35h          h ). . 5f6N    8Y,
                                                  . P(s.WW.  . 9ff.E!TP.C.O  3/4 .5-3 3/4.5.3    ECCS SUBSYSTEMS -                      .P.of 5.H.ufoo,yN,,A,N,q ,M0f 3/4  -7 3/4.5.4    REFUELING WATER TANK....................................            3/4 5-8 PALO VERDE - UNIT 1                        VI COFTROM.ED BY USR
 
l CONTROLLED BY USER INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS d                        C-ti Mnunnex *f'rc cMra:ntut:r
(            SECTION                                                                                                                          PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1    PRIMARY CONTAINMENT CONTAINMENT  INTEGRITY................................                                          3/4 6-1 CONTAINMENT  LEAKAGE..................................                                          3/4 6-2 h0                  3  CONTAINMENT        AIR LOCKS./Au/c cenwr*.Er4 isetAuct.    '
ANS aiAt;r.rt, Wyu) 3/4 6-4 (0                      INTERNAL PRESSURE.[.I'.C 9 .# D .#I..Y ff7.0............
                                                      .    .        .                                                        3/4 6-6 AIR  TEMPERATURE......................................                                          3/4 6-7 h                  CONTAINMENTd/ESSEDSTRUCTURAL INTEGRITY. . . . . . . . .3/4                                            . . . 6-8      ..
('end                CONTAINMENT VENTILATION      SYSTEM.......................                                    3/4 6-14 l              3/4.6.2    DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY    SYSTEM.............................                                      3/4 6-15 IODINE REMOVA L SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .        3/4 6-17 g                      > Ctt>fAtNt % th' C c u t h Sy5761A 75) 3/4.6.3    CONTAINMENT ISOLATION VALVES............................                                          3/4 6-19                        j 3/4.6.4    COMBUSTIBLE GAS CONTROL
                                        ..~ f.ti A LWti(L5 HYDROGEN;MONIJORJ....................................                                        3/4 6-36
(' ' N                  ELECTRIC HYDROGEN    RECOM8INERS........................                                    3/4 6-37 g'
          .g JMlE4_
HYDROGEN PURGE _CLEARNt#D SYSTEM::. . . . . . . . . . . . . . . . . . . . . . .              3/4 6-38 (C ('g3)              f 6thxgAnse; f_xM dxpAug lyIL cuffh uF fy&f&!
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PALO VERDE - UNIT 1                          VII CONTRoli En RY IICED                                                              _ _ _ _ _ _ _ _ _ _ _ _ _ _
 
                                                                                                            ~
CONTROLLED BY USER INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                                        PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1    TURBINE CYCLE SAFETY  VALVES........................................                    3/4 7-1
          @          Et4E% ENC /duX f LI ARD FEEDWATER SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 7-4 CONDENSATE STORAGE TANK..............................                      3/4 7-6 ACTIVITY.............................................                      3/4 7-7 MAIN STEAM LINE . ISOLATION VALVES. . . . . . . . . . . . . . . . . . . . . 3/4 7-9 ATMOSPHERIC DUMP  VALVES..............................                    3/4 7-10 3/4.7.2      STEAMGENERATORyRESSURE/TEMPERATURELIMITATION......... 3/4 7-11 (9(GfS) 3/4.7.3 655EETfAbC LbHGWATERSYSTEM.......................... 3/4 7-12                                                    ,
Q (sts) 3/4.7.4 MTAf < DRAY                            .W@P.W. gfE,R. . . . . . . . . . . . 3/4 7-13                              !
3/4.7.5      ULTIMATE HEAT SINK......................................                      3/4 7-14 O (sf5) 3/4.7.6 (SSENTIAL CHILLED WATER SYSTEM 3.799.8.P@0fEC1rlot4 ..... 3/4 7-15    .
Q (.5T5) 3/4.7.7        ONp0L ROOM @ENTIAL FILTRAT                    .      . NRP. 3/4 7-16
    @ (GTs) 3/4.7.8            PUMP ROOM AIR EXHAUST CLEANUP SYSTEM................ 3/4 7-19 3/4.7.9      SNUBBERS................................................                      3/4 7-21 3/4.7.10 SEALED SOURCE CONTAMINATION.............................                          3/4 7-27 3/4.7.11 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM........................                      3/4 7-29 SPRAY AND/OR SPRINKLER  SYSTEMS.......................                      3/4 7-32 SYSTEMS..........................................                    3/4 7-35                '
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                  @(STS) FIRE  42 HOSE STATIONS...................................                      3/4 7-37 3/4 7-40 i{HALON YARD      FIRE HYDRANTS AND HYDRANT 3/4 SYSTEMS........................................                    HOSE  7-42    HOUSES. . . .
G CSTsT 3/4.7.12 FIRE @ED ASSEM8M. 66@fM.R. .P.NN.C.NS. . . . . . . . . . . 3/4 7-43 h/4.7.13 SHUTDOWN COOLING AREA SYSTEM.................................                    3/4 7-45 Q(STS)l L3/4.7.14 C6 TROL ROOM AEIDTEMPERATURE. .Y.N.$. . . . . . . . . . . . . . .                  3/4 7-46 3/4.8 ELECTRICAL POWER SYSTEMS                                ,
3/4.8.1    A.C. SOURCES OPERATING............................................                      3/4 8-1 SHUTD0WN.............................................                      3/4 8-8
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  .                                                                                                                                    l PALO VERDE - UNIT 1                        VIII                                                                          ;
CON"RO'.L''D BY UMR 1
 
CONTROLLED BY USER II INDEX                                                                              l i!
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS                                                                      l SECTION                                                                                                            PAGE            4 ELECTRICAL POWER SYSTEMS (Continued) 3/4.8.2      D.C. SOURCES OPERATING..........................:.................                                        3/4 8-9 SHUTD0WN.............................................                                        3/4 8-13 3/4.8.3      ONSITE POWER DISTRIBUTION SYSTEMS                                                                                      -
0PERATING............................................                                        3/4 8-14 SHUTD0WN.............................................                                        3/4 8-16 3/4.8.4      ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.................................                                        3/4 8-17 MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES.................................                                        3/4 8-40 3/4.9 REFUELING OPERATIONS 3/4.9.1      BORON  CONCENTRATION.....................................                                        3/4 9-1 3/4.9.2      INSTRUMENTATION.........................................                                        3/4 9-2 3/4.9.3      DECAY TIME..............................................                                        3/4 9-3 l
l 3/4.9.4      CONTAINMENT BUILDING PEN ETRATIONS. . . . . . . . . . . . . . . . . . . . . . .                  3/4 9-4 l      3/4.9.5    COMMUNICATIONS..........................................                                          3/4 9-5 3/4.9.6    REFUELING MACHINE.......................................                                          3/4 9-6 3/4.9.7    CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.. .. . . .. . 3/4 9-7 3/4.9.8      SHUTDOWN COOLING AND COOLANT CIRCULATION HIGH WATER LEVEL.....................................                                        3/4 9-8 LOW WATER LEVE L. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-9 3/4.9.9    CONTAINMENT PURGE VALVE ISOLATION SYSTEM. . . . . . . . . . . . . . . . 3/t. 9-10 3/4.9.10 WATER LEYEL - REACTOR VESSEL                                                                                            .
F0 E L AS S EMB LIE S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 9-11 IEAs.................................................                                        3/4 9-12 3/4 9-13 3/4.9.11 WATER LEVEL - STO RAGE P00 L. . . . . . .mult46          -            . . . . . . u.  . . . . . . . . . . . is. . . . .
3/4.9.,12 FUEL BUILDING ESSENTIAL VENTILATION SYSTEM                        2          . 8;Rc.u?Muf 3/4 9-14 (s@.rs) 3/4.10 SPECIAL TEST EXCEPTIONS sys(en 3/4.10.1 SHUTOOWN MARGIN.........................................                                            3/4 10-1 3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS.............                                        3/4 10-2 3/4.10.3 R EACTO R COO LANT L00 P S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 10-3 PALO VERDE - UNIT 1                              IX                                                  -
COWRO1.MD B'f Um
 
CONTROLLED BY USER INDEX                                    -
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION                                                                              PAGE 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE................            3/4 10-4 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY........                  3/4 10-5 3/4.10.6 SAFETY INJECTION TANKS..................................                  3/4 10-6 e Y 374.10.7 SPENTFUEL_POOLLEVEQ.................................                              3/4 10-7 3/4.10.8 SAFETY INJECTION TANK PRESSURE..........................                  3/4 10-8  -
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7 3/4.11 RADIOACTIVE EFFLUEk<TS                                                      ~
      /
N 3/4.11.1 SECONDARY SYSTEM LIQUID WASTE DISCHARGES TO ONSITE                                N EVAPORATION PONDS CONCENTRATION...........................................            3/4 11-1 00SE....................................................            3/4 11-5 LIQUID HOLDUP TANKS.....................................            3/4 u-6 3/4.11.2 GASEOUS EFFLUENTS DOSE          RATE............................................... 3/4 n-7 h,                          DOSE - NOBLE GASES......................................              3/4 11-11    .
7                          DOSE - IODINE-131, 100INE-133, TRITIUM, AND RADIONUCLIDES IN PARTICULATE F0RM.....................      3/4 11-12 GASEOUS RADWASTE              TREATMENT.............................. 3/4 11-13 EXPLOSIVE GAS              MIXTURE................................... 3/4 11-14 GAS STORAGE              TANKS....................................... 3/4 11-15 3/4.11.3 SOLID RADI0 ACTIVE WASTE.................................                  3/4 11-16 3/4.11.4 TOTAL                  00SE..............;............................... 3/4 11-18 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM......................................                  3/4 12-1 3/4.12.2 LAND USE CENSUS.........................................                  3/4 12-11 3/4.'12.3 INTERLABORATORY COMPARISON PR0 GRAM......................                3/4 12-12 w-%-.  , , , ,  , , _ .
PALO VERDE - UNIT 1                                    X CONTROLLED T' US3R
 
CONTROLLED BY USER                                                                                        L INDEX                                                                        i BASES SECTION PAGE            j 3/4.0    APPLICABILITY..............................................                                        B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1    BORATION CONTR0L....................,...................
B 3/4 1-1 3/4.1.~2    BORATION  SYSTEMS........................................                                        B 3/4 1-2 3/4.1.3    MOVABLE CONTROL  ASSEMBLIES.............................. B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1    LINEAR HEAT  RATE........................................                                      B 3/4 2-1 3/4.2.2    PLANAR RADIAL PEAKING FACT 0RS...........................                                      B 3/4 2-2 3/4.2.3    AZIMUTHAL POWER  TILT....................................                                      B 3/4 2-2 3/4.2.4      DNBR  MARGIN.............................................                                      B 3/4 2-3 3/4.2.5      RCS F LOW RATE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 2-4 3/4.2.6      REACTOR COOLANT COLD LEG TEMPERATURE....................                                        B 3/4 2-4 3/4.2.7      AXIAL SHAPE  IN0EX.......................................                                      B 3/4 2-4 3/4.2.8      PRESSURIZER  PRESSURE....................................                                      B 3/4 2-4 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR PROTECTIVE AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION...............                                        B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION..............................                                            B 3/4 3-2
  &  3/'l . 3. 4 <upeme ovGRsPEED PRofEc4tDN (s&
PALO VERDE - UNIT 1                              XI CONTROtt?D BY UMP
 
CONTROLLED BY USER INDEX
{
BASES f'i!
SECTION PAGE r
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1        REACTOR COOLANT LOOPS AND COOLANT CIRCULATION...........                                B 3/4 4-1                                '
3/4.4.2        SAFETY  VALVES...........................................                                B 3/4 4-1 3/4.4.3        PRESSURIZER.............................................                                B 3/4 4-2 3/4.4.4        STEAM GENERATORS........................................
B 3/4 4-3
          ~3 /4.4.5        REACTOR COOLANT SYSTEM LEAKAGE........ ................. B 3/4 4-4 3/4.4.6        CHEMISTRY...............................................                                B 3/4 4-4 3/4.4.7        SPECIFIC ACTIVITY.......................................                                B 3/4 4-5 3/4.4.8        PRESSURE / TEMPERATURE LIMITS.............................                              B 3/4 4-6 3/4.4.9        STRUCTURAL INTEGRITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-11 3/4.4.10 REACTOR COOLANT SYSTEM        VENTS............................                                B 3/4 4-12 3/4.5 EMERGENC'Y CORE COOLING SYSTEMS (ECCS) 3/4.5.1        SAFETY INJECTION    TANKS..................................                              B 3/4 5-1 1
3/4.5.2 and 3/4.5.3          ECCS  SUBSYSTEMS............................. B 3/4 5-2                                                                        l l          3/4.5.4        REFUELING WATER    TANK....................................                              B 3/4 5-3 CE-ATMOSPHERIC TYPE CONTAINMENT 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1        PRIMARY CONTAINMENT..................................... B 3/4 6-1 3/4.6.2        DEPRESSURIZATION AND COOLING    SYSTEMS....................                            B 3/4 6-3 3/4.6.3        CONTAINMENT ISOLATION  VALVES............................                              B 3/4 6-4 3/4.6.4        COMBUSTIBLE GAS    CONTR0L.................................                            B 3/4 6-4
        @ 3/et,G.E          %NNE cusefauf sy&f EMS
    ' bG f
i PALO VERDE - UNIT 1                          XII f O N, TPv G. l @. 9.VI                            s %...R.
 
CONTROLLED BY USER INDEX BASES                                                      t                            '
SECTION
_                                  PAGE 3/4.7 PLANT SYSTEMS                -
3/4.7.1    TURBINE  CYCLE....................................l......                                            B 3/4 7-1 3/4.7.2      STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.........                                            B 3/4 7-3
\                        contenent b
;    ( c) 3/4.7.3 (l_SSENTTAL
                          ,_          COOLING waste        vuree      WATER SYSTEN. . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 7-3                      .
                                                                                                                                                              }
l                                                                                                                                                            '
l    (@f45) 3/4. 7. 4 CESSENTIAL SPRAY PON0'> SYSTEM. . . . . . . . . . . . . . . . . . . .
3/4.7.5 ULTIMATE HEAT SINK......................'...............                                                                                    l P                                                                                                                        B 3/4 7-4                      '
                                                                  . FLC@ ?Rt%cDcn                                                                            ;
3/4 7.6 (ESSENTIAL CHILLED WATER SYSTEM 3........................                                                  B 3/4 7-4 3/4.7.7    CONTROL ROOM ESSENTIAL FILTRATION                      SYSTEM................                          B 3/4 7-5 3/4.7.8    ESF PUMP ROOM AIR EXHAUST CLEANUP                      SYSTEM................ B 3/4 7-5 3/4.7.9    S NU B B E RS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
* B 3/4 7- 5 i
3/4.7.10 SEALED SOURCE CONTAMINATION............................. B 3/4 7-7 3/4.7.11 FIRE SUPPRESSION        SYSTEMS................................                                          B 3/4 7-7 3/4.7.12 FIRE-RATED ASSEMBLIES...................................                                                  B 3/4 7-8 f f ~/4.7.13 3          SHUTDOWN COOLING SYSTEM.................................                                                B 3/4 7-8 s 3/4.7.14 GIiNTROL R00'M' AIR            PERATURE,.9.'.". 0.9.#$................                                    B 3/4 7-8 (M9                                                          '
3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3              A.C. SOURCES, D.C. SOURCES, and ONSITE POWER DISTRIBUTION SYSTEMS................                                            B 3/4 8-1 3/4.8.4    ELECTRICAL EQUIPMENT PROTECTIVE                    DEVICES.'................ B 3/4 8-3                                                  f, 3/4.9 REFUELING OPERATIONS 3/4.9.1    BORON  CONCENTRATION..................................... B 3/4 9-1 3/4.9.2    INSTRUMENTATION......................................... B 3/4 9-1 3/4.9.3    DECAY  TIME..............................................                                              B 3/4 9-1 3/4.9.4    CONTAINMENTcBjILQNGPENETRATIONS.......................                                                  B 3/4 9-1 3/4.9.5    COMMUNICATIONS.......................................... B 3/4 9-1 PALO VERDE - UNIT 1                                  XIII 1
l COPROM.ED B'f USE                                                                                                              l 1
 
l CONTROLLED BY USER INDEX                                                                    .
BASES SECTION                                                                                                                  i PAGE                  i 3/4.9.6      REFUELING MACHINE.......................................                        B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING.........                        B 3/4 9-2 3/4.9.8                                                                                                                  ;
SHUTDOWN COOLING AND COOLANT CIRCULATION................
B 3/4 9-2                I 3/4.9.9        CONTAINMENT PURGE VALVE ISOLATION SYSTEM................
t B 3/4 9-3 3/4.9.10 and 3/4.9.11      WATER LEVEL - REACTOR VESSEL a                                                              -
STORAGE POOL
                                        .................g g .g. g .nd .
                                                                                      .g .                  B 3/4 9-3                i j  ) 3/4. 9.12 i[ijEL BUILDING ESSENTIAL VENTILATdNT SYSTEM. . . . . . . . . . . . . .B 3/4 9-3 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1      SHUTDOWN                                                                                                      !
MARGIN.........................................                        B 3/4 10-1                  (
3/4.10.2 MODERATOR TEMPERATURE COEFFICIENT, GROUP HEIGHT, INSERTION, AND POWER DISTRIBUTION LIMITS................                        B 3/4 10-1 3/4.10.3      REACTOR COOLANT L00PS...................................                        B 3/4 10-1 3/4.10.4 CEA POSITION, REGULATING CEA INSERTION LIMITS AND REACTOR COOLANT COLD LEG TEMPERATURE................                          B 3/4 10-1 3/4.10.5 MINIMUM TEMPERATURE AND PRESSURE FOR CRITICALITY........                          B 3/4 10-1 3/4.10.6 SAFETY INJECTION        TANKS.................................. 8 3/4 10-2
    @ C 3/4.10.7 SPENLFUELP00klM . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/
3/4.10.8 SAFETY INJECTION TANK        PRESSURE.......................... 5 3/4 10-2
              '4.11  RADIOACTIVE EFFLUENTS                                                                      . _ , . . .
b LN!- l3/4.11.1 SECONDARY EVAFORATION SYSTEM LIQUID WASTE DISCHARGES TO ONSITE P0NDS.......................................                        B 3/4 11-1 f3/4.11.2        GASEOUS EFFLUENTS.........'..............................                        B 3/4 11-2 3/4. n.3 So u D RADI0 ACTIVE WASTE. . . . . . . . . . . . . . . . . . . . . . . . . . . . .; . . . . B 3/4 11-5 3/4.11.4 TOTAL 00SE..............................................                        B 3/4 11-5 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING                                                                        3 3/4.12.1 MONITORING PR0 GRAM...................................... B 3/4 12-1 3/4.12.2 LAND USE        CENSUS.........................................                      B 3/4 12-2 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM......................                            B 3/4 12-2
        . PALO VERDE - UNIT 1                          XIV - -
_s'
          'N f.QNc T',?Q.T.:.D.
                                            .  ~            'Q.V. 1 M~W      .
L
 
                                                                                            ~
CONTROLLED BY USER i i l
INDEX                                                      -
DESIGN FEATURES
                                ~
l SECTION                                                                              PAGE l
51 SITE
                                          ._ - ~._ _N                                                              j 5.1.1    SITE AND EXCLUSION BOUNDARIES...........................                    5-1                i l
    ?    5.1. 2    LOW POPULATION 20NE. . . . . . ...........................
                                                      . . . .'i                                5-1 YhM    5.1.3    GASEOUS RELEASE POINTS......
(
5.2 CdNTAINMENT
                                      ,        /
5-1
_ . . ~ -        - . - -
i 5.2.1      CONFIGURATION.................'.'.........................                  5-1 5.2.2      DESIGN PRESSURE _AND-~-TEMPERATURE ,..                                      5-1 5.3 REACTOR CORE 5.3.1      FUEL      ASSEMBLIES.........................................              5-5                l 5.3.2    CONTROL ELEMENT ASSEMBLIES..............................                    5-5 5.4 REACTOR COOLANT SYSTEM i
5.4.1      DESIGN' PRESSURE AND TEMPERATURE.........................                  5-5 l
l 5.4.2
                                                                                                                  )
5-5 l                  V0LUME...............................        ..................
5.5 METEORLOGICAL TOWER  LOCATION.................................                    5-6 5.6 FUEL STORAGE j ,
5.6.1      CRITICALITY.............................................                    5-6 5.6.2      0RAINAGE................................................                    5-6 1
5.6.3      CAPACITY................................................                    5-6 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT..........................                      5-6 l
f PALO VERDE - UNIT 1                      XV CONTROll.ED BY USER
 
CONTROLLED BY USER INDEX                                                .
ADMINISTRATIVE CONTROLS SECTION PAGE 6.1    RESPONSIBILITY...............................................                      6-1 6.2 ORGANIZATION
                      -                      ~
6.2.1 0FFSITE.... U .......Z ............ D ....................                          6-1 4
                    ,I 6.2.2 UNIT          STAFF............................................../..                6-1 I 6.2.3        INDEPENDENT SAFETY ENGINEERING GROUP
  ?'c i
* FUNCTION...................................................                      6-6 COMPOSITION................................................                      6-6 RESP 0NSIBILITIES...........................................                    6-6 l              AUTHORITY..................................................                      6-6 i
RECOR05....................................................                      6-6
(. 6.2.4        ' SHIFT TECHNICAL ADVIS0R....................................
6-6 6.3 UNIT STAFF QUALIFICATIONS....................................                        6 i 6.4  TRAINING.....................................................                      6-7 t
6.5 REVIEW AND AUDIT
              ;        6.5.1 PLANT REVIdBOARD
            ,I FUNCTION...................................................
6-7 p5                            COMPOSITION................................................                      6-7 i                      ALTERNATES.................................................                      6-7 MEETING        FREQUENCY..........................................              6-7 QU0 RUM.....................................................                      6-8 RESP 0NSIBILITIES...........................................                      6-8 AUTHORITY..................................................                      6-8
    .,                          RECORDS....................................................                      6-9 6.5.2 TECHNICAL REVIEW AND CONTR0L...............................
6-9 N ~_            _ _ _
PALO VERDE - UNIT 1                        XVI CONTROLLED BY USER
 
CONTROLLED BY USER                                                          l INDEX                                                .
l ADMINISTRATIVE CONTROLS                                                                        '
SECTION                                                                      PAGE r,
E.5.3 NUCLEAR SAFETY GROUP FUNCTION...................................................        6-10    s COMPOSITION................................................        6      b      !,            CONSULTANTS................................................        6-10        i REVIEW.....................................................        6-10 i          AUDITS.....................................................        6-11      ,
                \          AUTHORITY..................................................        6-12    ,/
(      RECORDS....................................................
                                                                            ~
6-12 J
6.6 REPORTABLE EVENT ACTI0N......................................            6-12 6.7 SAFETY LIMIT VIOLATION.......................................            6-13 6.8 PROCEDURES AND PR0 GRAMS......................................          6-13 6.9 REPORTING REQUIREMENTS N
6.9.1 ROUTINE ~ REPORTS............................. m . _ y .....          6-16 p; i i              STARTUP REP 0RT...................................                  6-16 s
          !                ANNUAL REP 0RTS......................                              6-17 MONTHLY OPERATING REP 0RT...................................        6h17
          \
ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT.........        6 17 i
            ;            SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT.............        6-l18 i
j 6.9.2 CPECIAL REP 0RTS............................................        / -20 6
l
                                                                      ~                d                          <
l 6.10 RECORD RETENTION............................................        ,
6-20 6.11 RADIATION PROTECTION PR0 GRAM................................          6-22 6.12 HIGH RADIATION AREA.........................................            6-22 i
N      ~~  6 .13 PROCESSCONTROLPR0 GRIM.l...................................            6-23                {
GC45)-                                      ~
l l
PALO VERDE - UNIT 1                      XVII CON" ROLLED BV UMR
 
I
                                                                                ~
CONTROLLED BY USER                                                      l 1
INDEX                                        .  ,l II ADMINISTRATIVE CONTROLS SECTION PAGE
                                      ~ ~ ~
6.14 0FFSITE DOSE CALCULATION      MANUAL................'-                    6-23      l
($
A 6.15 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS, AND SOLIO WASTE TREATMENT      SYSTEMS...............,............        6-24 6.16 PRE-PLANNED ALTERNATE SAMPLING PR0 GRAM. . . . . . . . . ( . . . . . . . .6-25 l
mwe. as
* l l
e-PALO VERDE - UNIT 1                      XVIII CONTROLLED BY USER
 
e-CONTROLLED BY USER INDEX LIST OF FIGURES                                                                                        l PAGE 3.1-1      ALLOWABLE MTC MODES 1 AND 2............................ 3/4 1-5, 3.1-2      MINIMUM BORATED WATER V0LUMES.......................... 3/4 1-12'
            @ 3.1-2A                          CEA INSERTION LIMIT VS. THERMAL POWER...... 3/4 1-2's (3.1-28      CORE POWER IR IT AFTER CEA DEVIATION    ................. 3/4 1-24 3.1-3      CEA INSERTION LIMITS VS THERMAL POWER ct                      (TCOLSS IN SF R E)...................    ................. 3/4 1-31
                  '3.1-4 L              CEAINSERTIONLIMITSVSTHERMALPOW5./
(COLSS OUT OF SERVICE).                                    3/4 1-32 3.2-1      DNBR MARGIN OPERATING LIMIT BASED ON COLSS                                            ,
3/4 2-6 (COLSS IN SERVICE).....................................
3.2-2      DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR (COLSS OUT OF    SERVICE)...................... 3/4 2-7 3.2-3      REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVEL.................................................. 3/4 2-10 3.3-1    C DsBR MARGIN OPERATING LIMIT BASED ON COLSS
      @                    QRBOTHCEAC'SINOPERABLE..........._,.,1._j.............
3/4 3-10 3.4-1      DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY
                                > 1.0 pCi/ GRAM DOSE EQUIVALENT I-131................... 3/4 4-28 3.4-2'      REACTOR COOLANT SYSTEM PRESSURE _IEMP_ERAIURE r                      LIMIIATIONS YdF070107 EARS OF FULL POWER]
C0PERATION................... _... w.........'........... 3/4 4-30
( 4!751 SAMPLING PLAN FOR SNUBBER FUNCTIONAL TEST..............
N 3/4 7-26 N
              /
B 3/4.A-1 NIL-DUCTILITY TRANSITION TEMPERATURE INCREASE AS A                            }  .
            ,I                FUNCTION OF FAST (E > 1 MeV) NEUTRON FLUENCE                                      l
    -      t                  (550*F    IRRADIATION).................................... B 3/4 4-10i
  & li  '
5.1-1      SITE AND EXCLUSION BOUNDARIES.......................... 5-2 I
5.1-2      LOW POPULATION Z0NE.................................... 5-3 l
        ;          5.1-3      GASEOUS RELEASE P0INTS................................. 5-4            !                l 6.2-1      0FFSITE ORGANIZATION................................... 6-3
              \    6.2-2      ONSITE UNIT ORGANIZATION............................... 6-4      .
I
                                                                                                  /
                \' 'PALO VERDE - UNIT 1            - XIX ^                                v' CONTRORED PN 1mR
  ~
j
 
CONTROLLED BY USER                                                                                          :
                                                            .I_NDEX LIST OF TABLES
_?        ~
                                                                                                                            -PAGE 1.1            FREQUENCY    N0TATION......................................                                          1-8
: 1. 2 O P ERATIONA L M00 ES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-9 2.2-1          REACTOR PROTECTIVE INSTRUMENTATION TR LIMITS. . . . . . . . . . . . . . . . . . . . ...................
                                                                              . . . . . . . . . . . IP SETPOINT              2-3 2.2-2          CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS........ 2-7 R
  #                    LRE@ IRED)      MONITORING FREOUENCIES FOR BACKUP BORON                                          /
DILUTION DETECTI0km a buNCTION 0F~0PERATING CHARGINC'                                    . M{"'
                      @UMPDND7EANT OPERATIONAL MODES _ ................
3.1-1          FOR K,ff > 0.98... ..... .... ...... .................                                              3/4 1-16 y      3.1-2          FOR 0.98 > K,ff > 0.97..................................                                            3/4 1-17 3.1-3          FOR 0.97 > K,ff > 0.96..................................                                            3/4 1-18 3.1-4          FOR 0.96 > K,ff > 0.95..................................
3/4 1-19 3.1-5          FOR K,ff 3.3-1 10.95........................................./ 3/4 1-20 REACTOR' PROTECTIVE INSTRUMENTATION. . . . . . . . . . . . . . . . . . . . . 3                      . /4 3-3 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES.......                                            3/4 3-11 3.3-2a        INCREASES IN BERRO, BERR2, AND BERR4 VERSUS RTD DELAY  TIMES..................................                              ........              3/4 3-13 4.3-1          REACTOR PROTECTIVE INSTRUMENTATTOCSURVEILLANCE REQUIREMENTS............................................                                            3/4 3-14 3.3-3          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION.........................................                                            3/4 3-18 3.3-4          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP    VALUES.............................                                        3/4 3-25 3.3-5          ENGINEERED SAFETY FEATURES RESPONSE                TIMES...............                            3/4 3-28 4.3-2          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
              ._ ._. INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............
                                ~~
3/4 3-31
      . 3.3-6          RADIAfIONMONITORI'NGINSTRUMbTATION.$...............,...                                            3/4 3-38 4.3-3          RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.........................................i' ht    3.3-7
                                                                                                                ..          3/4 3-40 3/4 1-43 4.3-4          SEISMICMONITORINGINSTRUMENTATION................../...
SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS....................................../.....
                                                                                                        /                                -
3/4 3-44 3.3-8          METEOROLOGICAL MONITORING INSTRUMENTATION. . . . . . .                          ......              3/4 3-46 4
k.3-5            METEOROLOGICAL MONITORING INSTRUMENTATION                            /
S.URVEIL_ LANCE REQUIRE,_M, ENTS. . . . .~. . . .;. . . . . . .d . . . . . . . . . . .
                            -                                                                                              3/4 3-47 3.3-9          REMOTE _ SHUTDOWN 1,1NSTRUMENTATION LDISC0NNECL .3/4                                                . . . 3-49 (SWITCHESANDCONTROLCIRCUITSj PALO VERDE - UNIT 1                              XX CONTROLLED BY4 '"ER
 
_    . - _ . _              .-          -      ~
s i
i I
INDEX                                                      I
            ~ LIST OF TABLES PAGE 1.1                FREQUENCY NOTATION . . . . . . . . . . . . . . . . . ...            1-5 1.2                OPERATIONAL MODES . . . . . . . . . . . . . . . . . . . .            1-6 2.2-1              REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS .            2-3 2.2-2      . CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS . . . .                  2-7
^
3.1-1            PONITORING F EQUENCIES FOR BACXUP BORON DILUTION 7                          DETECTI    OR SYSTEM 80 EXTENDED FUEL CYCLE FOR dl-    3.1-2
[            K,ff >    8 FOR MODES 3, 4 AND 5 . . . . . . . . . . .          3/4 1-17 MONITORING FREQUENCIES FOR BACXUP BORON DILUTION
        /                        DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR                                1 l                        0.98 3 ,ff X > 0.97 FOR MODES 3, 4 AND 5 . . . . . . .
3/4 1-18 I3                  MONITORING FREQUENCIES FOR BACXUP BORON DILUTION l.1-3 3.1-4 DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.97 > X,ff > 0.96 FOR MODES 3, 4, AND 5 . . . . . . .            3/4 1-19 MONITORING FREQUENCIES FOR BACXUP BORON DILUTION                                      .
DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR                                  I 0.96 g X,ff > 0.95 FOR MODES 3, 4, AND 5 . . . . . . .          3/4 1-20
        ' 3.1-9            MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.95 > K g    FOR MODES 3, 4, AND 5 . . . . ._              . 3/4 1-21 j
                                                                                                                )
3.3-1 ~ REACTOR PROTECTIVE INSTRUMENTATION . . . . . . . . . . J .' . - 3/4-3 -3 '
3.3-2'          REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES . . . .            3/4 3-9 4.3-1            REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 3/4 3-12 3.3-3          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
,                                INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . 3/4 3-17 3.3-4          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM
  ,                              INSTRUMENTATION TRIP VALUES . . . . . . . . . . . . . .          3/4 3-24 3.3-5          ENGINEERED SAFETY FEATURES RESPONSE TIMES . . . . . . . .              3/4 3-27 4.3-2          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . .          3/4 3-30 3.3-6          REMOTE SHUTDOWN SYSTEM INSTRUMENTATION . . . . . . . . .                3/4 3-41 4.3-3          REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS . . . . . . . . . . . . . . . . . . . . . 3/4 3-42 3.3-7          POST-ACCIDENT MONITORING INSTRUMENTATION . . . . . . . .                3/4 3-44 XVII                        Amendment Number 9 February 27, 1984
 
CONTROLLED BY USER INDEX LIST OF TABLES PAGE 4.3-6      REMOTE SHUTDOWN INSTRUMENTATION SURVEILLANCE  REQUIREMENTS...............................                                        3/4 3-S6      .
I 3.3-10    POST-ACCIDENT MONITORING INSTRUMENTATION................                                          3/4,3-S8      -
i 4.3-7      POST-ACCIDENT MONITORING INSTRUMENTATION                                                                        I ,
SURVEILLANCE REQUIREMENTS...............................                                          3/4 3-59 3.3-11    FIRE DETECTION INSTRUMENT                            u..................
3/4 3-62 3.3-12    LOOSE PARTS SENSOR LOCATIONS.........                                                            3/4 3-70 c-      3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING
      !            INSTRUMENTATION.........................................                                        3/4 3-72
    )4.3-8        RADI0 ACTIVE GASEOUS EFFLUENT MONITORING l'              INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............                                        3/4 3-77 i
f4.4-1          MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED OURING INSERVICE  INSPECTION.............................                                        3/4 4-16 4.4-2      STEAM GENERATOR TUBE INSPECTION.........................                                          3/4 4-17 Q.4-1          REACTOR COOLANT SYCTEM PRESSURE ISOLATION' VALVES........                                        3/4 4-21
                          ~%_______..                          ---
3.4-2      REACTOR COOLANT SYSTEM  CHEMISTRY........................                                        3/4 4-23 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS............................................                                          3/4 4-24 4.4-4      PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM....................................                                        3/4 4-27 4.4-5      REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHD RAWA L SCHEDU LE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-31 4.6-1      T    N SURVEILLANCE NST7 EAR.T... . . . . . . . . . . . . . . . . . . . .
3/4 6-12 4.6-2      TENDON LIFT-OFF FORCE - FIRST YEAR......................                                          3/4 6-13 y      3.6-1      CONTAINMENT ISOLATION VALVE $xFeNP.lf!E..$*.t.65..........                                        3/4 6-21 3.7-1      STEAM LINE SAFETY VALVES PER L00P,s......................                                        3/4 7-2 PALO VERDE - UNIT 1                          XXI CONTROLLED BY USER
 
                                                                                                                                        .t CONTROLLED BY USER                                                                  ll INDEX LIST OF TABLES                                                      -
PAGE 3.7-2      MAXIMUM ALLOWABLE STEADY STATE POWER LEVEL AND MAXIMUM VARIABLE OVERPOWER TRIP SETPOINT WITH INOPERABLE STEAM g-                              (INE SAFETY VALVES. .D,.W,9, M, ,@,g ,9ff4M.4f.". f.%. . . .
yuH(s ct62Ahr%                                                            3/4 7-3
_4.7-1        SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY W            SAMPLE AND ANALYSIS PR0 GRAM.............................                    3/4 7-8 i3.7-3          SPRAY AND/0R SPRINKLER  SYSTEMS..........................                    3/4 7-34 i
                        ;3.7-4 FIREHOSESTATIONS...............................\          ..'.....          3/4 7-39
                                                                                                          \
i '3. 7-5          YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE H0 VS E S . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4..        . . 7-41
                          '4.8-1            DIESEL GENERATOR TEST  SCHEDULE.........................i                . 3/4 8-7
                                                                                                                  \
3.8-1        D. C. ELECTRICAL SOURCES. . . . . -                                          3/4 8-11 4.8-2        BATTERY SURVEILLANCE REQUIREMENTS.......................                    3/4 8-12 3.8-2        CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT OPS                                      PROTECTIVE  DEVICES.....................................;.
3 i
3/4 8-19 3.8-3 i
MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION AND/OR BYPASS  DEVICES..................................'i              . 3/4 8-41 I
4.11-1        RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM.,                    3/4 11-2 i
I                                                                                                i 4.11-2        RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS                      '
t PR0 GRAM.................................................                    3/4 11-8
            !                                                                                                i 3.12-1        RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM. . . . . . . . . . .          3/4 12-3 i                3.12-2        REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL  SAMPLES................................                  3/4 12-7
          .                4.12-1        DETECTION CAPABILITIES FOR' ENVIRONMENTAL SAMPLE
          }                              ANA LYS I S . . . . . . . . . . . . . . . . . . . . . . . .3/4                . . 12-8 8 3/4.4-1 REACTOR VESSEL TOUGHNESS..................                                        B 3/4 4-8
                                                                                              ,_  ~
5.7-1        COMPONENT CYCLIC OR TRANSIENT          LIMITS.....................            5-7 5.7-2          PRESSURIZER SPRAY N0ZZLE USAGE FACTOR...................                      5-9
      .[ [ (6.2-1                        MINIMUM SHIFT 3 REUTCOMPOSIT Q ......................... 6-5 PALO VERDE - UNIT 1                                XXII e ONTP.C'.N. S.Y ". *. *wP.
l
 
            .sa' S
CONTROLLED BY USER
                                      =. . ..
i e
O SECTION 1.0 DEFINITIONS 4
6 4
O
* 9 CONTROL. LED BY USSR                    i J
 
CONTROLLED BY USER 1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.
AXIAL SHAPE INDEX
: 1. 2 The AXIAL SHAPE INDEX shall be the power generated in the lower half of the core less the power generated in the upper half of the core divided by the sum of these powers.
AZIMUTHAL POWER TILT - Tq D
1.3 AZIMUTHAL POWER TILT shall be the power asymmetry between azimuthally symmetric fuel assemblies.
CHANNEL CALIBRATION                                          ,
1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channei output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL                    .
CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observatica. This determination shall include, where                        i possible, comparison of the c' anel indication and/or status with other indications and/or status der.ved from independent instrument channels measuring the same parameter.                                                                    j l
PALO VERDE - UNIT 1                        1-1 CONTROLLED BY USER
 
CONTROLLED BY USER                                                        i DEFINITIONS CHANNEL FUNCTIONAL TEST
: 1. 6 A CHANNEL FUNCTIONAL TEST shall be:
: a. Analog channels - the. injection    of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.
  .                  b. Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.                    s
: c. Digital computer channels - the exercising of the digital computer          .              !
hardware using diagnostic programs and the injection of simulated                      ,
                          -process data into the channel to verify OPERABILITY including alarm                    a and/or trip functions _                                          w                    \,'
( d.        Ridiological~ effluent process monitoring channels - the CHANNEL N                      '
      @p        (          FUNCTIONAL TEST may be performed by any series of sequential,        T                    j s
overlapping, or total channel steps such that the entire channel      i C.M}          Q Lfunctionally u        tested.                                            _j                  i The CHANNEL FUNCTIONAL TEST shall include adjustment, as necessary, of                      ;
the alarm, interlock and/or trip setpoints such that the setpoints are within the required range and accuracy.                            .
CONTAINMENT INTEGRITY
: 1. l $.e pp;g cg                                                    {
f 1.7 CONTAINMENT INTEGRITY shall exist when:
a.
li n                    All penetrations required to be closed during accident conditions g
are either:                                                --
i                          1. Capable of being closed by a.. OPERABLE containment automatic.                  ;
isolation valve system, or l
A                2. Closed by manual valves, blind flanges, or deactivated automatic
          \                      valves secured in their closed positions, except as provided in                [
t Table 3.6-1 of Specification 3.6.3.
            \        b. All equipment hatches are closed and sealed,                                      i  i i      c. Each air lock is in compliance with the requirements of                          f Specification 3.6.1.3,
: d. The containment leakage rates are within the limits of Specification 3.6.1.2, and
                                                                                                              )
(q bellows
: e. The sealing mechanism associated with each penetration (e.g.', welds, or 0-rings) is OPERABLE.                                                  '
j CONTROLLED LEAKAGE 1.8 Not Applicable.
CORE ALTERATION                                                                        -
1.9 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
PALO VERDE - UNIT 1                      1                                    C O N ' R O !,! 5 D B V U S S
 
CONTROLLED BY USER DEFINITIONS DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131"(microcuries/
gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
Y - AVERAGE DISINTEGRATION ENERGY 1.11 I shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than fodines, with half-lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from.when the monitored parameter exceeds its ESF actuacion setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel                    ,
generator starting and sequence loading delays where appitcable,                          j i
FREQUENCY NOTATION
                                                                                                    ^
1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.
f GASEOUS RADWASTE SY3 EM '                                                                  -
gof -r    1.14 A GASEOUS RADWASTE SYSTEM shall be any system designed and installed to N            '
reduce radioactive gaseous ef fluents by collecting primary coolant system QM        offgases from the primary system and providing for delay or holdup for the gose of reducing the total radioactivity prior to release to the environment.
l IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:
: a. Leakage into closed systems, other than reactor coolant pump                      I controlled bleed-off flow, such as pump seal or valve packing leaks that are captured and conducted to a sump or collecting. tank, or
: b. Leakage into the containment atmosphere frem sources that are both                i specifically located and known either not to interfere with the                    I operation of leakage detection systems or not to be PRESSURE                      l BOUNDARY LEAKAGE, or                                                              !
1
: c. Reactor Coolant' System leakage through a steam generator to the                  '
secondary system.
l PALO VERDE - UNIT 1                    1-3 CONTROLLED BY USER                                                          l 1
l
 
                                                                                                ~
CONTROLLED BY USER                                                      L DEFINITIONS l
i1 MEMBER (S) 0F THE PUBLIC 1.16 MEMBER (S) 0F THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from            i this category are persons who enter the site to service equipment or to make            ;
deliveries. This category does include persons who use portions of the site              i for recreational, cccupational, or other purposes not associated with the                I gof  plant.
(pE    OFFSITE 00SE CALCULATION MANUAL (00CM)                                                !
1.17 The OFFSITF DOSE CALCULATION MANUAL shall contain the current methodology .'
and parameters usad in the calculation of offsite doses due to radioactive        -
gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm / trip setpoints, and in the conduct of the environmental      /
qdiologicalmonitoringprogram.                                          .. /
OPERABLE - OPERABILITY 1.18 A systain, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s),                  i and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).
OPERATIONAL MODE - MODE 1.19 An OPERATIONAL MODE (i.e. MODE) shall correspond to any one inclusive combination of c~ ore reactivity condition, power level, and cold leg reactor coolant temperature specified in Table 1.2.
PHYSICS TESTS 1.20 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved oy the Commission.
PLANAR RADIAL PEAKING FACTOR - F xy                                                            I 1.21 The PLANAR RADIAL PEAKING FACTOR is the ratio of the peak to plane average power density of the individual fuel rods in a given horizontal plane, excluding the effects of azimuthal tilt.                                                        ;
PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a nonisolable fault in a Reactor Coolant System component body, pipe wall, or vessel wall.
PALO VERDE - UNIT 1                      1-4                  .
CO.NTROLLED BY USER 1
j
 
l CONTROLLED BY USER DEFINITIONS N
                  / PROCESS CONTROL PROGRAM (PCP)
              /      1.23 The PROCESS CONTROL PROGRAM shall contain the provisions to assure that
            ;        the SOLIDIFICATION of wet radioactive wastes results in a waste form with                                t
{        properties that meet the requirements of 10 CFR Part 61 and of low level I        radioactive waste disposal sites. The PCP shall identify process parameters
            !        influencing SOLIDIFICATION such as pH, oil content, H2 O content, solids content,                            ,
l        ratio of solidification agent to waste and/or necessary additives for each                                  4
:        type of anticipated waste, and the acceptable boundary conditions for the
    @f                process parameters shall be identified for each waste type, based on laboratory
( Refd            scale and full-scale testing or experience. The PCP shall also include an                              ;
identification of conditions that must be satisfied, based on full-scale                              i
          ,          testing, to assure that dewatering of bead resins, powdered resins, and filter                      i sludges will result in volumes of free water, at the time of disposal, within                    l the limits of 10 CFR Part 61 and of low level radioactive waste disposal                          i sites.
l PURGE - PURGING i          1.24 PURGE or PURGING shall be the controlled process of discharging air or
(\        gas from a confinement to maintain temperature, pressure, humidity, concentra-tion, or other operating condition, in such a manr.ar that replacement air or                        ,
                \ gas is required to purify the confinement.                                                                }
RATED THERMAL POWER 1.25 RATED THERMAL POWER ,shall be a total reactor core heat transfer rate to
* the reactor coolant of 3800 MWt.
REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA drive mechanism.
REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in
    , % x > Section  pHIE C  50.73 hu:Lto;th 10tnt-GC&l CFR Part- 50.
See Ct(fl,
                                                                .t con % S6R
                                                                                                      ~
                                                                                                                                              ' ~~
    ^'
SHUTDOWN MARGIN 1.28 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be.suberitical from its present condition i                    assuming:
: a. No change in part-length control element assembly position, and
: b. All full-length control element assemblies (shutdown and regulating)
I                                are fully inserted except for the single assembly of highest reactivity worth which is assumed to be fully withdrawn.
l l                    PALO VERDE - UNIT 1                        1-5 1
CONTRO'.tE SY USER L.
 
CONTROLLED BY USER DEFINITIONS SITE BOUNDARY                                                          '' N h            1.29 The SITE BOUNDARY shall be that line beyond which the land is neither 6                    nor leased, nor otherwise controlled by the 1icensee.
SOFTWARE 1.30 The digital computer SOFTWARE for the reactor protection system shall be the program codes including their associated data, documentation, and procedures.
SOLIDIFIC w
1.31 SOLIDIFICATION shall be the convers' ion of radioactive wastes from liquid h*p_>fi systems solid withtodefinite a homogeneous volume and(uniformly    distributed),
shape, bounded      by monolithic,  immobilized a stable surface    of distinct (7@yl        outline on all sides (free-standing).
          ! SOURCE CHECK l
h 1.32 A SOURCE CHECK shall be the qualitative assessment of channel response (when the channel sensor is exposed to a source of increased radioactivity.
                                                              ' ' ~ '              ~
STAGGERED TEST BASIS                                        ~~        -
1.33 A STAGGERED TEST BASIS shall consist of:
                                                                                              ~
: a. A test schedule for n systems, subsystems l, trains, or other designated components obtained by dividing the specified test interval                        ,
into n equal subintervals, and
: b. The testing of one system, subsystem, train, or .other designated              . . . _ _
component at the beginning of each subinterval.
THERMAL POWER 1.34 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
UNIDENTIFIED LEAKAGE 1.35 UNIDENTIFIED LEAKAGE shall be all leakage which does not constitute eith_e        TIFIED LEAKAGE or reactor coolant pump controlled.. bleed-off flow.
                                                                              ~
UNRESTRICTED AREA ~
ji(      1.36 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY s-      access to which is not controlled by the licensee for purposes of' protection Rfi5); 'of individuals from exposure to radiation and radioactive materials, or any
          \areawithintheSITEBOUNDARYusedforresidentialquartersorforindustrial, commercial, institutional, and/or recreational purposes.
PALO VERDE - UNIT 1                      1-6 CONTROLLED BY USER
 
CONTROLLED BY USER                                                                              !!
l DEFINITIONS
{i
        /(-
                                                                                                                                      +
VENTILATION E'!HAUST TREATMENT SYSTEM                                                            ;
l.37 A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and                                -      '
i            installed to reduce gaseous radioicoire or radioactive material in particulate                                  i I          form ir, effluents by passing ventilation or vent exhaust gases through charcoal                                {
Mp 4
          !        adsorbers and/or HEPA filters for the pumose of removing fodines or partic-                                      '
ulates from the gaseous exhaust stream prior to the release to the environment.
In@g/            Such a system is not considered to have any effect on noble gas effluents.
T                Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered                                  I' to be VENTILATION EXHAUST TREATMENT SYSTEM components.                                                          I VENTING
{      1.38 VENTING shall be the controlled process of discharging air or gas from a (g      confinement to maintain temperature, pressure, humidity, concentration, or other operating condition, in such a manner that replacement air or gas is not x    provided or required during VENTING. Vent, used in system names, does not N imply a VENTING process.
N    ~
e 6
PALO VERDE - UNIT 1                      1-7 CONTROLLED BY USER
 
                                                                ~
CONTROLLED BY USER                              l DEFINITIONS
                                                                        )
TABLE 1.1 FREQUENCY NOTATION NOTATION                        FREQUENCY S              At least once per 12 hours.
D              At least once per 24 hours.
W                At least once per 7 days.
B o 9 ->      4              (Atleast4timespermonth' (Nj'gg}                        at intervals no greater than 9 days-and a minimum of 48 times per year.
M                At least once per 31 days.
Q              At least once per 92 days.
SA        ,
At least once per 184 days.
R              At least once per 18 months.
  /3,ep 4    [                  Completed prior t M h rETease n (P4I            S/U            Prior to each reactor startup.
N.A.            Not applicable.
S I
l PALO VERDE - UNIT 1              1-8 CONTROLLED BY USER
 
                                                                                        ~
CONTROLLED BY USER OEFINITIONS
                                                                                                          ~ ~
TABLE 1.2 OPERATIONAL MODES REACTIVITY          % OF' RATED    COLD LEG ~                        ,
OPERATIONAL MODE          CONDITION. K,ff      THERMAL POWER *                                      !
TEMPERATURE gecap      (Tcold) p,S._,  1.                    W                            > 5%          > d50*F)
POWER OPERATION          t 0.99
: 2. STARTUP "                g 0.99                < 5%          1 [.iQ'F
: 3. HOT STANDBY              < 0.99                    0          g J50'F
: 4. HOT SHUTDOWN            < 0.99                    0 350' > Tcold > 210*F                    ,
: 5. COLD SHUTDOWN            < 0.99                    0          < 210*F 4 +-y.                                                                                    l ph      6. REFUELING *3            1 0.95                    0          1 135'F G                                                                                                            '
i f
f
            *
* Excluding decay heat.                                                                              !
b M 4 O uel in the reactor vessel with the vessel head closure bolts less than                          ,
1 g                fully tensioned or with the head removed.                                                          f f
y Je 4 j,,: d m X Lffl & M      W    W 3*I*I*N.
Lg g c3 e
PALO VERDE - UNIT 1                      1-9 CONTROLLED BY USER
 
                                                                ~
CONTROLLED BY USER                                                      .
I l        l f
1 SECTION 2.0                                    .
SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 1
f l
                                                                                          )
l COF7 ROLLED BV U.9:R r                                                                                      l
 
CONTROLLED BY USER 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS 2.1.1 REACTOR CORE DNBR 2.1.1.1 The calculated DNBR of the reactor core shall be maintained greater than or equal to 1.231.
. APPLICABILITY: MODES 1 and 2.
ACTION:
Whenever the calculated DNBR of the reactor has decreased to less than 1.231, be in HOT STANDBY within 1 hour, and comply with the requirements of Specifi-cation 6.7.1.
PEAK LINEAR HEAT RATE 2.1.1.2 The peak linear heat rate (adjusted for fuel rod dynamics) of the            ,
fuel shall be maintained less than or equal to 21 kW/ft.
APPLICABILITY: MODES 1 and 2.                                                      '
ACTION:
Whenever the peak linear heat rate (adjusted for fuel rod dynamics) of the fuel has exceeded 21 kW/ft, be in HOT STANDBY within 1 hour, and comply with the requirements of Specification 6.7.1.                                          *'
i REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.                i APPLICABILITY: MODES 1, 2, 3, 4, and 5.
ACTION:
MODES 1 and 2:
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour, and comply with the requirements of Specification 6.7.1.
MODES 3, 4, and 5:
Whenever the Reactor Coolant System pressure has exceeded 2750 psia, reduce the Reactor Coolant System pressure to within its limit within 5 minutes, and comply with the requirements cf Specification 6.7.1.
PALO VERDE - UNIT 1                    2-1 CONTROLLED BY USER
 
                                                                                                        ~
CONTROLLED BY USER SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETV SYSTEM SETTINGS l
REACTOR TRIP SETPOINTS 2.2.1 The reactor protective instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.                                                                      '
APPLICABILITY: As shown for each channel in' Table 3.3-1.
ACTION:
With a reactor protective instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS 2.2.2 Core Protection Calculator Addressable Constants shall be in accordance with Table 2.2-2.
APPLICABILITY: As shown for' Core Protection Calculators in Table 3.3-1.
ACTION:
With a Core Protection Calculator Addressable Constant less conservative than the value shown in the Allowable Value column of Table 2.2-2, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status.
2-2 PALO VERDE - UNIT 1 CONTROLLED BY USER i
                                                            -                            _              -    ~
 
s                                                                        TABLE 2.2-1 5
      <                                            REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E                    '
E
* e            FUNCTIONAL UNIT                                          TRIP SETPOINT          ALLOWABLE VALUES E
(Q')                        -
I. TRIP GENERATION                                                          d
                                                                                  /
                                                                                      - - ~ - ~ .. ~.                                                                                    .
: 1. Pressurizer Pressure - High I                $ 2383 psia              1 2388 psia                                                              spg
: 2. Pressurizar Pressure - Low (2)              3 1837 psia (2)          -> 1822 psia (2) 2 q                          3. Steam Generator Level - Low cy)
                                                                              > 44.2% (4)              > 43.7% (4)                                                                    Z g                          4. Steam Generator Level - liigh c n)          1 91.0% (9)              i 91.5% (9)                                                                    y g                        5.
6.
Steam GeneraLor Pressure - Low Gi            > 919 psia (3)
_                      > 912 psia (3)
,s                              Containment Pressure - liigh                  5 3.0 psig              1 3.2 psig                                                                      y e"  '?                    7. Reactor Coolant Flow - Low (z)                                                                                                                i        f""
N Rate 4)                                        i 1.05%/s (6)(7)        5 1.10%/s (6)(7)                                                        \
: b. Floor (a)                                > 52.2% (6)(7)          > 47.2% (6)(7) 4                              c. Band C d                                  i 40.0% (6)(7)          < 42.1% (6)(7)                                                                  E
: 8. Local Power Density - High(S)                $ 21.0 kW/ft (5) i 21.0 kW/ft. (5)                                                              <
C                          9. DNBR - Low (5)                                > 1.231 (5)              > 1.231 (5)                                                                    C d                    B. Excore Neutron Flux                                                                                                                                        M Inf1                                                                                                                                                                              I I
.g                        1. Variable Overpower Trip d )                                                                                                                            FT1 T. Rate (to) y
                                                                              < 10.6%/ min of RATED    < 11.0%/ min of RATED
(                                                    TilERMAL POWER (8)      TilERHAL POWER (8)
                          '- b. Ceiling (tG                              < 110.0% of RATED        < 111.0% of RATED                                                            l TilERMAL POWER (8)      THERMAL POWER (8)                                                            l
: c. Band Q ed                                < 9.8% of RATED          < 10.0% of RATED                                                          I A
TilERMAL POWER (8)      TilERHAL POWER (8)
Y
 
Fb          l
                                                                                              ?I TABLE 2.2-1 (Continued)          /
                          ,.              REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LINITS
        =
y        FUNCTIONAL UNIT                                      TRIP SETPOINT        ALLOWABLE VALUES h a({ & a d 50 R c            2. Logaritimic Power Level - High (1) 5                a. Startup and Operating                    < 0.798% of RATED    < 0.895% of RATED O  [                                                    ,
THERMAL POWER        THERMAL POWER Q
: b. Shutdown                    k        < 0.798% of RATED    < 0.895% of RATED                            O
    ]                                                                        RMAL POWER    THERMAL POWER d            C. Core Protection Calculator System
                                                                                  ~~
H
: 1. CEA Calculators                              Not Applicable      Not Applicable                              N f""              2. Core Protection Calculators                  Not Applicable      Not Applicable                              y==
n  f""
T**
M  {        D. Supplementary Protection System
                                                                                        ~
g      Lg                    m Pressurizer Pressure - High              C.      1 2409 psia          12414      psia)      N g                                                              ~_                              -- %
q      II. RPS LOGIC                                          -
A. Hatrix Logic                                    Not Applicable      Not Applicable
{
M            B. Initiation Logic                                Not Applicable      Not Applicable                              M M                                                                -
m
    %      III. RPS ACTUATION DEVICES y
A. Reactor Trip Breakers                            Not Applicable      Not Applicable B. Manual Trip                                      Not Appilcable      Not Applicable 9
k                                                                                                                - . .__ _ _ __ _ _ _
 
CONTROLLED BY USER                                                              l TABLE 2.2-1    (Continued)                                      -
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (1) Trip may be manually bypassed above 10 4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10 +% of RATED THERMAL POWER.
(2) In MODES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pres-surizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(3) In MODES 3-6, value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(4) % of the distance between steam generator upper and lower level wide range instrument nozzles.
(5) As stored within the Core Protection Calculator (CPC). Calculation of the trip setpoint includes measurement, calculational and processor uncer-tainties, and dynamic allowances'. Trip may be manually bypassed below 1%
of RATED THERMAL POWER; bypass shall be automztically removed when THERMAL POWER is greater than or equal to 1% of RATED THERMAL POWER._
The approved DNBR limit is 1.231 which includes a partial rod bow N penalty x J^        compensation. If the fuel burnup exceeds that for which an increased rod bow penalty is required, the ONBR limit shall be adjusted. In this case a ONBR trip setpoint of 1.231 is allowed provided that the difference is com-pensated by an increase in the CPC addressable constant BERR1 as follows:
RB - R8 o d (% POL)
BERRi new
                      = BERRi old U+      100 d (% DNBR) where BERRi old is the uncompensated value of BERR1; R8 is the fuel rod bow penalty in % DNBR; R8, is the fuel rod bow penalty in % DNBR already accounted for in the DNBR limit; POL is the power operating limit; and d (% POL)/d (% DNBR) is the absolute value of the most adverse derivative                      ;
of POL with respect to DNBR.
                                ~                                                  -
PALO VERDE - UNIT 1                          2-5 CONTROLLED BY USER-
 
CONTROLLED BY USER TABLE 2.2-1 (Continued)                                      -
REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS TABLE NOTATIONS (Continued)
(7)@ RATE is the maximum rate of decrease of the trip setpoint. There are no                                ;
restrictions on the rate at which the setpoint can increasa.                                        i FLOOR is the minimum value of the trip setpoint.                                                    j BAND is the amount by which the trip setpoint is below the input signal                            '
unless limited by Rate or Floor.                                                                    i Setpoints are % of 100% power flow conditions.
l (3) @ The setpoint may be altered to disable trip function 'during testing pursuant to Specification 3.10.3.
[ ( @ @ RATE is the maximum rate of increase of the trip setpoint. There are no restrictions on the rate at which the setpoint can decrease.
I
    ,              CEILING is the maximum value of the trip setpoint.
l              BAND is the amount by which the trip setpoint is above the input signal *
    ,            unless limited by the rate or the ceiling.
I f( ) h % of the distance between steam generator upper and lower level narrow
    ;            range instrument nozzles.
I I
E                                          W      -
ew (9) (JAU.sd:) & f$[ED'N WAU 9
PALO VERDE - UNIT 1                      2-6 F, ONTP,N. 1.m. . ".Y UE. F.Fs
 
CONTROLLED BY USER
                                                                                                                      ~
TABLE 2.2-2                                                          -
CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS I. TYPE I ADDRESSABLE CONSTANTS POINT ID      PROGRAM                                                          ALLOWABLE NUMBER          LABEL          DESCRIPTION                                            VALUE 60        FC1            Core coolant mass flow rate calibration constant
                                                                                              -              b.
4f4 & O
                                                                                            ,                        SAR 61        FC2            Core coolant mass flow rate calibration      >< 0.0 N                                        constant                                    l 62        CEANOP          CEAC/RSPT inoperable flag                        0,1, 2 or 3 63        TR            Azimuthal tilt allowance                          1 1.02                            !
64        TPC            Thermal power calibration constant l10.90 65        KCAL          Neutronfluxpowercalibrationconstantf10.85 i
66        DNBRPT        ON8R pretrip setpoint                        . Unrestricted 67        LPDPT          Local power density pretrip setpoint              Unrestricted            '
N                                              \
II. TYPE II ADDRESSABLE CONSTANTS ~ X _
POINT ID      PROGRAM                                                                                            ,
NUMBER          LABEL        DESCRIPTION                                                                        ,
                                                                                        \
r      068        BERR0          Thermal power uncertainty bias used in        N l                                  DNBR calculation                                  i l                                                                                      '
  .\
069        BERR1          Power uncertainty factor used in DNBR calculation j        070        BERR2          Neutron flux power uncertainty bias used in DNBR calculation
      !'        071        BERR3          Power uncertainty factor used in local l                                    power density calculation 072        BERR4          Power uncertainty bias used in local power density calculation                                                    -
l          073        EOL            End of life flag I
l            074        ARM 1 t            075        ARM 2 076                        Multipliers for planar radial peaking
    \            077 ARM 3 ARM 4          factors N                                  . . . . .
                                                            ~~'
                                                                        ^        ~'
PALO VER0E - UNIT 1                          2-7        .
.                                  CONTROLLED BY USER                                                                      -
 
I
                                                                                                    ~
CONTROLLED BY USER
                                                                                          /
l
                                ,/                                TABLE 2.2-2 (Continued)
CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS II. TYPE II ADDRESSA8LE CONSTANTS (Continued)
[ POINT ID          PROGRAM                                                        -
NUMBER          LABEL              DESCRIPTION l            078        ARMS Q:,1 /                      079        ARM 6 C/ !                      080        ARM 7
        /                  081        SC11 082        SC12 083        SC13 084        SC21 085        SC22                Shape annealing correction matrix 086        SC23 i                  087        SC31 088        SC32 089        SC33
          !                090        PFMLTD              ONBR penalty factor multiplier 091        PFMLTL              Local power density penalty factor multiplier 092        ASM2 093        ASM3 i,            094        ASM4                Multipliers for CEA shadowing factors
                <          095        ASMS 096        ASM6
                  .i        097        ASM7
                    !      098        CORR 1              Slope of the temperature shadowing ~    "
l                                        correction factor (Cgj) i 1          099        BPPCC1 l            100        BPPCC2              Boundary Point Power Correlation l            101        BPPCC3                Coefficients l
102        BPPCC4 l
103        RPCLIM              Reactor Power Cutback Time Limit j                      CEAC Addressable Constant Value
                \
s        295        TCBP                  RPC Max Time (only addressable via operator module)
                        's
                                  '~~-          , _ - . . .
PALO VERDE - UNIT 1                                2-8 CONTROLLED BY USER                                              '
 
                                              ~
CONTROLLED BY USER                            !
i 1
BASES FOR SECTION 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l                                                -
i e
CON ~ROU.F,D P3Y IJSER
 
CONTROLLED BY USER NOTE The BASES contained in the succeeding pages summarize the            ;
reasons for the specifications of Section 2.0 but in accord-ance with 10 CFR 50.36 are not a part of these Technical Specifications.
e CONTROLLED BY USER
                                                            '      ~
 
CONTROLLED BY USER 2.1 abs SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING BASES                                                                                -
                                                                                                      )
2.1.1 REACTOR CORE The restrictions of these safety limits prevent overheating of the fuel                '
cladding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by (1) restricting fuel operation to within the nucleate boiling            ,
regime where the heat transfer coefficient is large and the cladding surface,            l temperature is slightly above the coolant saturation temperature, and                    i (2) maintaining the dynamically _ adjusted, peak linear heat rate of the fuel            j M'    at or less than 21 kW/ftihich will not caush fuel centerline melting in any fuel rod.                      tr frwed                                                  ;
First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface i
temperature is only slightly greater than the coolant saturation temperature.
The upoer boundary of the nucleate boiling regime is termed " departure from nucleate boiling" (DNB). At this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperatures and the possibility of cladding failure.
Correlations predict DNB and the location of DNB for axially uniform and non uniform heat flux distributions. The local DNB ratio (DNBR), defined as the ratio of the predicted DNB heat flux at a particular core location to the        '
actual heat flux at that location, is indicative of the margin to DNB. The minimum value of DNBR during normal operation and design basis anticipated operational occurrences is limited to 1.231 based upon a statistical combination of CE-1 CHF correlation and engineering factor uncertainties and is established
                                                      ~            ~
e as a Safety _ Limit.JThe ONBR limit ~of'1.231 includes a rod bow compensat' ion if 0.8% on DNBR. For fuel burnups which exceed that for which an increased rod s h      bow penalty is required, the DNBR limit shall be adjusted. In this case the DNBR trip setpoint of 1.231 is allowed if the required DNBR increase is empensated_by_an increase _of_the addr_essable constant BERR1.
Second, operation with a pe,ak linear heat rate below that which would
  .      cause fuel centerline melting maintains fuel rod and cladding integrity.
Above this peak linear heat rate level (i.e., with some melting in the center),
      ' fuel rod integrity would be maintained only if the design and operating conditions are appropriate throughout the life of the fuel rods. Volume changes which accompany the solid to liquid phase change are significant and require accommodation. Another consideration involves the redistribution of the fuel which depends on the extent of the melting and the physical state of the fuel rod at the time of melting. Because of the above factors, the steady state value of the peak linear heat rate which would not cause fuel centerline melting is established as a Safety Limit. To account.for fuel rod dynamics (lags), the directly indicated linear heat rate is dynamically adjusted by the CPC program.
PALO VERDE - UNIT 1                  8 2-1 CONTROLLED BY USER                                                -
 
                                                                                                  ~
CONTROLLED BY USER SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS                                                  '
BASES Limiting Safety System Settings for the Low ONBR, High Local Power Density, High Logarithmic Power Level, Low Pressurizer Pressure and High Linear Power Level trips, and Limiting Conditions for Operation on DNBR and kW/ft margin are specified such that there is a high degree of confidence that the specified acceptable fuel design limits are not exceeded during normal operation and design basis anticipated operational occurrences.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the                                  l Reactor Coolant System from overpressurization and thereby prevents the                          _
release of radionuclides contained in the reactor coolant from reaching the                  t  A ,)
containment _ atmosphere.                      - - - -      -~-
7 The Reactor Coolant System components are designed to Section III,                    ?
N        1974 Edition, Summer 1975 Addendum, of the ASME Code for Nuclear Power Plant Components which permits a maximum transient pressure of 110%'(2750 psia) cf                  '
design pressure. The Safety Limit of 2750 psia is therefore consistent with (the_desi gn_ cr.i ta ti a_ and _as s oc i ated_ code _requi remen ts .              s The entire Reactor Coolant System is hydrotested at 3125 psia to e5    $'" ? {:# n$ k 'G r$ ) $ t h $ $ b N ) A $ eb*"*
2.2.1 REACTOR TRIP SETp0INTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their Safety Limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered 2d Safety Features Actuation System in mitigating the consequences of <acci_de~nts.k1G                        '
Operation with r. trip set less conservative than its Trip Setpoint but withTn $3 its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.
The DNBR - Low and Local Power Densith - High are di                                *
      .4 trip setpoints based on Safety Limits of 1.231 and 2I(kW/gitally                      generated ft, respectively.
Since these trips are digitally generated by the Core Protection Calculators, the trip values are not subject to drifts common to tript, generated by analog type equipment. The Allowable Values for these trips are therefore the same as the Trip Setpoints.                                          .
To maintain the margins of safety assumed in the safety analyses, the calculations of the trip variables for the DN8R - Low and Local Power Density -
High trips include the measurement, calculational and processor uncertainties cd and dynamic allowances as defined in<CESSA1 System 80 applicable system descriptions and safety analyses.
PALO VERDE - UNIT 1                            8 2-2
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2.1.2      REACTOR COOLANT SYSTEM PRESSURE The restriction ~of this safety limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
        ,,m f 'J / The reactor pressure vessel, piping, and pressurizer are designed to Section
        -/ III of the ASME Code for Nuclear Power Plant Components which pemits a l maximum transient pressure of 110% (2750 psia) of the design pressure. The
            !    Reactor Coolant System valves and fittings, are designed to either Section III of the ASME Code or ANSI B 31.7, Class I, which permits a maximum transient pressure of 110%(2750 psia) of component design pressure. See Applicant's FSAR for specific Code, Standard Editions, and Addenda. The safety limit of 2750 psia is therefore consistent with the design criteria and associated code g equirements.
The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.
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Amendment Number 9 I                                                                                February 27,1984
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CONTROLLED BY USER
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7a,J2.QAFETY          LIMITS ANDiLIMITING SAFETY SYSTEMS SETTINGS                                          .
BASES
[REACTORTRIPSETPOINTS(Continued)..
t The methodology for the calculation of the PVNGS trip setpoint values, plant protection system, is discussed in the CE Document No. CEN-286(V) dated (July 31,1984.
Manual Reactor Trip The Manual reactor trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.
Variable Overpower Trio A reactor trip on Variable Overpower is provided to protect the reactor core during rapid positive reactivity addition excursions. This trip function will trip the reactor when the indicated neutron flux power exceeds either a rate limited setpoint at a great enough rate or reaches a preset ceiling. The flux signal used is the average of three linear subchannel flux signals originating in each nuclear instrument safety channel. These trip setpoints are provided in Table 2.2-1.
Locarithmic Power level - Hich The Logarithmic Power Level - High trip is provided to protect the integrity of fuel cladding and the Reactor Coolant System pressure boundary in the event of an unplanned criticality from a shutdown condition. A reactor trip is initiated by the Logarithmic Power Level - High trip unless this trip is manually bypassed by the operator. The operator may manually bypass this trip when the THERMAL POWER level is above 10 4% of RATED THERMAL POWER; this bypass is automatically removed when the THERMAL POWER level decreases to
!            10 *% of RATED THERMAL POWER.
Pressurizer Pressure - High The Pressurizer' Pressure - High trip, in conjunction with the pressurizer safety valves and main steam safety valves, provides Reactor Coolant System protection against overpressurization in the event of Toss of load without reactor trip. This trip's setpoint is below the nominal lift setting of the pressurizer safety valves and its operation minimizes the undesirable opera-tion of the pressurizer safety valves.
Pressurizer Pressure - Low The Pressurizer Pressure - Low trip is provided to trip the reactor and to assist the Engineered Safety Features System in the event of a decrease in Reactor Coolant System inventory and in the event of an increase in heat PALO VERDE - UNIT 1                  B 2-3 CONTROLLED BY USER
 
CONTROLLED BY USER                                          ,
      .2 2          - x (SAFETYLIMITSAND'LIMITINGSAFETYSYSTEMSSETTINGS BASES                                                -
Pressurizer Pressure - Low (Continued)
          -removal by the secondary system. During normal cperation, this trip's set-point may be manually decreased, to a minimum value. of 100 psia, as pressurizer pressure is reduced during plant shutdowns, provided the margin    .
between the pressurizer pressure and this trip's setpoint is maintained at less than or equal to 400 psi; this setpoint increases automatically as pressurizer pressure increases until the trip setpoint is reached. The operator may manually bypass this trip when pressurizer pressure is below 400 psia. This bypass is automatically removed when the pressurizer pressure increases to 500 psia.
Containment Pressure - High The Containment Pressure - High trip provides assurance that a reactor trip is in'tiated in the event of containment building pressurization due to a pipe break inside the containment building. The setpoint for this trip is identical to the safety injection setpoint.
Steam Generator Pressure - Low The P.eam Generator Pressure - Low trip provides protection in the event of an increase in heat removal by the seconca y system and subsequent cooldown of the reactor coolant. The setpoint is sufficiently below the full load' operating point so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This trip's setpoint may be manually decreased as steam generator      '
pressure is reduced during plant shutdowns, provided the margin between the steam generator pressure and this trip's setpoint is maintained at less than or equal to 200 psi; this setpoint increases automatically as steam generator pressure increases until the normal pressure trip setpoint is reached.              ,
Steam Generator Level - Low The Steam Generator Level - Low trip provides protection against a loss of feedwater flow incident and assures that the design pressure of the Reactor Coolant System will not be exceeded due to a decrease in heat removal by the secondary system. This specified setpoint provides allowance that there will l    ,. be sufficient water inventory in the steam generator at the time of the trip l    't>    to provide a margin of at least 10 minutes beforesauxiliary feedwater is required to prevent degraded core cooling,          h w .a g u    e Local Power Density - High
* The Local Power Density - High trip is provided to prevent the Ifnear heat rate (kW/ft) in the limiting fuel rod in the core from exceeding the fuel design limit in the event of any design bases anticipated operational occur-rence. The local power density is calculated in the reactor protective system utilizing the following information:
PALO VERDE - UNIT 1                    8 2-4 CONTP,00.ED BY USfR i
 
CONTROLLED BY USER SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES Local Power Density - High (Continued)
: a. Nuclear flux power and axial power distribution from the excore flux monitoring system;
: b. Radial peaking factors from the position measurement for the CEAs;
: c. Delta T powet from reactor coolant temperatures and coolant flow measurements.
* s The local power density (LPD), the trip variable, calculated by the CPC incorporates uncertainties and dynamic compensation routines. These uncer-tainties and dynamic compensation routines ensure that a reactor trip occurs when the actual core peak LPD is sufficiently less than the fuel design limit such that the increase in actual core peak LPD after the trip will not result in a violation of the Peak Linear Heat Rate Safety Limit. CPC uncertainties related to peak LPD are the same types used for DNBR calculation. Dynamic compensation for peak LPD is provided for the effects of core fuel centerline temperature delays (relative to changes in power density), sensor time delays, and protection system equipment time delays.
DNBR - i.ow The DN8R - Low trip is provided to prevent the DN8R in the limiting-coolant channel in the core from exceeding the fuel design limit in the event of design bases anticipated operational occurrences.. The DN8R - Low trip 1@ incorporates a low pressurizer pressure floor of 0861 psia. At this pressure
    ~ a ON8R - Low trip will automatically occur. The ONBR is calculated in the CPC utilizing the following information:
: a. Nuclear flux power and axial power distribution from the excore neutron flux monitoring system;
: b. Reactor Coolant System pressure from pressurizer pressure measurement; c .'  Differential temperature (Delta T) power from reactor coolant temperature and coolant flow measurements;
: d. Radial peaking factors from the position measurement for the CEAs;
: e. Reactor coolant mass flow rate from reactor coolant pump speed;
: f. Core inlet temperature from reactor coolant cold leg temperature measurements.
PALO VER0E - UNIT 1                    B 2-5 C O ' F R 01,11 3 3Y U M R
 
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SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS f
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BASES ONBR - Low (Continued)
The ONBR, the tr4 variable, calculated by the CPC incorporates various uncer-tainties and dyranic compensation routines to assure a trip is initiated prior to violation of fuel design limits. These uncertainties and dynamic compensa-tion routines ensure that a reactor trip occurs when the calculated core ONBR is sufficiently greater than 1.231 such that the decrease in calculated core                  .
DNBR after the trip will not result in a violation of the DNBR Safety Limit.
CPC uncertainties related to DNBR cover CPC input measurement uncertainties, algorithm modelling uncertainties, and computer equipment processing uncertainties. Dynamic compensation is provided in the CPC calculations for the effects of coolant transport delays, core heat flux delays (relative to changes in core power), sensor time delays, and protection system equipment time delays.
The ONBR algorithm used in the CPC is valid only within the limits indicated below and operation outside of these limits will result in a CPC initiated trip.
Parameter                      Limitino Value a.
b.
RCS Cold Leg Temperature-Low RCS Cold Leg Temperature-High
                                                              $ 70*F f I 610*F                .
            .      c. Axial Shape Index-Positive                Not more positive than + 0.5
: d. Axial Shape Index-Negative                Not more negative than - 0.5
: e. Pressurizer Pressure-Low                  > 1861 psia g
: f. Pressurizer Pressure-High                7 2388 psia
                                                                  ~
                                                                                                      /
: g. Integrated Radial Peaking                                                    /
Factor-Low                            > 1.28
                                                                                                  ,/
: h. Integrated Radial Peaking                                              /
i.
Factor-High Quality Margin-Low
                                                              - 1 4.28-                  /$:aCpicawt2M 0            y Steam Generator Level - Hich yeg,gg, g 4 m_ m g ,[4/tt w u _ ~ - -
The.SteamGeneratorLevel-Hightripfisprovidedtoprotecttheturbine Ffrom excessive moisture carry ever.~ 5fEce the turbine is automatically 9[ .*,h 1
tripped when the reactor is tripped, this trip provides a reliable means for e      providing protection to the turbine from excesssive moisture carryover. This trip'ssetpointdoesnotcorrespondtoasafety_Jimit.,andprovidesprotection;)
Qn the event of excesLfte ater flowdThe setpoint is identical to the main steafn'Is51stfon setpoint. Iti finctional'eapabilit/~at~the specified~ trip g Qe tt fjniEinhanc e s _th e cier_a l Lte l i ab i li_ty_o lth e_ro ac to r_ p ro tec t i o n .sy s t.em.
PALO VERDE - UNIT 1                        826 f"'()N s        TW  . . . l3.:B. QY 1 Typ
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i CONTROLLED BY USER                                                      t i
SAFETY LIMITS AND LIMITING SAFETY SYSTEMS SETTINGS BASES Reactor Coolant Flow - Low                  tteepwp6(poCAI*[>fh-C d '^ M 4
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The Reactor Coolant Flow - Low t p provides protection against,a reactor coolant pump shear'ed shaft event and        fouFpt:mp flow coasTd6wn duringTsteam
    @ Gint briakTith~To'ss~of offsite powr. Atrip is initiated when the pressure ~
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                                                                                ~  ~
iliffe~riTitfil acrois the primary ~ side 6f either steam generator decreases below a variable setpoint. This variable setpoint stays a set amount below the pres-sure differential unless limited by a set maximum decrease rate or a set minimum value. The specified setpoint_ ensures that a reactor trip occurs 't'o'Trevent .
Vi614 tion ~of~ Peak LfneTr Heat Rate or DNBR~$ifetiTimits under the stated /
04waisaf V di N :n [< N          di Pressurizer Pressure - Hiah (SPS)
The Supplementary Protection System (SPS) augments reactor protection against overpressurization by utilizing a separate and diverse trip logic from the Reactor Protection System for initiation of reactor trip. The SPS will initiate a reactor trip when pressurizer pressure exceeds a predetermined value.
2.2.2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS The Core Protection Calculator (CPC) addressable constants are provided to allow calibration of the CPC system to more accurate indications of power                ,
level, RCS flow rate, axial flux shape, radial peaking factors and CEA devia-tion penalties. Other CPC addressable constants allow penalization of the cal-
;      culated DNBR and LPD values based on measurement uncertainties or inoperable equipment. Administrative controls on changes and periodic checking of
,      addressable constant values (see also Technical Specifications 3.3.1 and l      6.8.1) ensure that inadvertent misicading of addressable constants into the CPC's is unlikely.
4 PALO VERDE - UNIT 1                    8 2-7 CON ~ ROV.'M BY 'JSER
 
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CONTROLLED BY USER 1
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I SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS I
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CONTROLLED BY USER 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 APPLICABILITY LIMITING CONDITION FOR OPERATION l
4 3.0.1 Compliance with the Limiting Conditions for Operation contained in the l
succeeding specifications is required during the OPERATIONAL MODES or other                ,
conditions specified therein; except that upon failure to meet the Limiting                ;
  , Conditions for Operation, the associated ACTION requirements shall be met.                '
3.0.2 Noncompliance with a specification shall exist when the requirements of              ,
the Limiting Condition for Operation and/or associated ACTION requirements are not met within the specified time intervals. If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.                      ,
3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within 1 hour.' action shall be initiated to place the unit in a MODE in which the specification does not apply by placing it, as applicable, in:
: 1. At least POT STAN08Y within the next 6 hours,
: 2. At least HOT SHUT 00WN within the following 6 hours, and
: 3. At least COLD SHUTOOWN within the subsequent 24 hours.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications.
This specification is not applicable in H00E 5 or 6.
3.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the conditions of the Limiting Condition for Operation are met without reliance on provisions contained in the ACTION requirements. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION statements. Exceptions to these requirements are stated in the individual specifications.            .
PALO VER0E - UNIT 1                    3/4 0-1 CONTROLLED BY USER                                                        ;
 
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CONTROLLED BY USER l
APPLICABILITY                                                                      .,
SURVEILLANCE REQUIREMENTS l
4.0.1    Surveillance Requirements shall be applicable during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Suiveillance Requirement shall be performed within the specified time interval with:
: a. A maximum allowable extension not to exceed 25% of the surveillance interval, and                                                        .
: b. The combined time interval for any three consecutive surveillance intervals not to exceed 3.25 times the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the specified time interval shall constitute a failure to meet the OPERABILITY requirements for a Limiting Condition for Operation. Exceptions to these requirements are stated in the individual specifications. Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an CPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation have been performed within the stated surveillance interval or as otherwise spec'ified.                                    -
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class I, 2, and 3 components shall be applicable as follows:
: a. Inservice inspection of ASME Code Class 1, 2, and 3 components and inservice testing ASME Code Class I, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50 Section 50.55a(g)(6)(1).
: b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressuro Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
PALO VERDE - UNIT 1                    3/4 0 2 CONTROLLED BY' USER
 
CONTROLLED BY USER
  . APPLICABILITY                                                                    . .
SURVEILLANCE REQUIREMENTS (Continued) 4.0.5 (Continued)
ASME Boiler and Pressure Vessel Code and applicable            Required frequencies Addenda terminology for              for performing inservice inservice inspection and            inspection and testing testino activities                      activities Weekly                      At least once per 7 days Monthly                      At least once per 31 days Quarterly or every 3 months          At least once per 92 days Semiannually or every 6 months      At least once per 184 days Yearly or annually            At least once per 366 days
: c. The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
: d. Perfor1ance of the above inservice inspection and testing activities            i shall be in addition to other specified Surveillance Requirements.
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: e. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed        f to supersede the requirements of any Technical Specification.
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PALO VERDE    UNIT 1                3/4 0 3 CONTROLLED BY USER
 
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CONTROLLED BY USER 3/4.1 REACTIVITY CONTROL SYSTEMS                                                      !
3/4.1.1 B0 RATION CONTROL
* SHUTOOWN MARGIN - T;,g GREATER THAN 210'F                                            !
LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTOOWN MARGIN shall be greater than or equal to 6.0%
delta k/k.
APPLICABILITY: H0 DES 1, 28, 3, and 4.
ACTION:
With the SHUTOOWN MARGIN less than 6.0% delta k/k, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution con-taining greater than or equal to 4000 ppm baron or equivalent until the required SHUT 00WN MARGIN is restored.
SURVE!LLANCE REQUIREMENTS 4.1.1.1.1 The $ HUT 00WN MARGIN shall be determin d to be greater than or equal to 6.0% delta k/k:
: a. Within 1 hour after detection of an inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable. If the inoperable CEA is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable CEA(s).
: b. When in H00E 1 or H00E 2 with K    greater than or equal to 1.0 at least once per 12 hours by veridIng that CEA group withdrawal Is within the Transtant Insertion t.imits of Specification 3.1.3.6.
: c. When in H00E 2 with K    less than 1.0, within 4 hours prior to achievingreactorcrilI[alitybyverifyingthatthepredicted critical CEA position is within the limits of Specification 3.1.3.6.
See Special Test Exception 3.10.1.
PALO VER0E - UNIT 1                    3/4 1 1
                                      ..              f)        .
 
CONTROLLED BY USER                                                l REACTIVITY CONTROL SYSTEMS                                                          I SURVEILLANCE REQUIREMENTS (Continued)
: d. Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e. below, with the CEA groups at the Transient Insertion Limits of Specification 3.1.3.6.
: e.    " hen in MODE 3 or 4, at least once per 24 hours by consideration of
  .        et least the following factors:                                    ,
: 1. Reactor Coolant System boron concentration,
: 2. CEA position,
: 3. Reactor Coolant System average temperature,
: 4. Fuel burnup based on gross thermal energy generation,
: 5. Xenon concentration an
: 6. SamariumconcentratIon.d 4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1.0% delta k/k at least once per 31 Effective Full Power Days (EFPD). Tiits comparison shall consider at least those factors stated in Specification 4.1.1.1.le., above. The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 EFPD af ter each fuel loading.
PALO VERDE - UNIT 1                        3/4 1 2 b            ...)        kb
 
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  ,A REACTIVITYCONTROLSYSTEMSj                                                            '
SHUT 00WN MARGIN - T g    LESS THAN OR EQUAL TO 210'F LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTOOWN MARGIN shall be greater than or equal to 4.0%
delta k/k.
APPLICABILITY: H00E 5.
ACTION:
With the SHUTOOWN MARGIN less than 4.0% delta k/k, immediately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 4000 ppm baron or equivalent until the required SHUT 00WN MARGIN is restored.                                                    -
SURVEILLANCE REQUIREMENTS 4.1.1.2 The SHUTOOWN MARGIN shall be determined to be greater than or equal to 4.0% delta k/k:
: a. Within 1 hour after detection of an inoperable CEA(s) and at least once per 12 hours thereafter while the CEA(s) is inoperable.
If the inoperable CEA is immovable or untrippable, the above required SHUTDCWN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immovable or untrippable CEA(s).
: b. At least once per 24 hours by consideration of the following factors:
: 1. Reactor Coolant System boron concentration,
: 2. CEA position,
: 3. Reactor Coolant System average temperature, 4 .'  Fuel burnup based on gross thermal energy generation,
: 5. Xenon concentration, and
: 6. Samarium concentration.
4 PALJ VERDE = UNIT 1                    3/4 1-3
                          ,    . a,,.,.,..      .1 9
 
r CONTROLLED BY USER W  q~ REACTIVITY CONTROL SYSTEMS                                                      .
  ,      MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (HTC) shall be within the area of Acceptable Operation shown on Figure 3.1-1.
APPLICABILITY: MODES 1 and 2*#
ACTION:
With the moderator temperature coefficient outside the area of Acceptable Operation shown on Figure 3.1-1, be in at least HOT STANDBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.1.3.1 The MTC shall be determined to be within its Ilmits by confirmatory measurements. MTC measured values shall be extrapolated and/or compensated to permit direct comparison with the above limits.
4.1.1.3.2 The MTC shall be determined at the following frequencies and THERMAL POWER conditions during each fuel cycle:            .'
: a. Prior to initial operation above 5% of RATED THERMAL POWER, after each fuel loading,
: b. At any THERMAL POWER, within 7 EFP0 after reaching a core average
!                    exposure of 40 EFP0 burnup into the current cycle.
: c. At any THERMAL POWER, within 7 EFPD after reaching a core average exposure equivalent to two thirds of the expected current cycle end-of-cycle core average burnup.
          *With Keff greater than or equal to 1.0.
        #5ee $pecial Test Exception 3.10.2.
PALO VERDE - UNIT 1                          3/4 1 4
                                                    .            k
 
1 w
T                                -
M                                                                                                                        FIGURE 3.1 1                                                                                      ,i
                                                  ~.
                                                          'x          '
ALLOWABLE MTC MODES 1 AND 2                                                                                  c ~ P#
                                                                                                                                                                                                              ' I' N              '
[                            ~
                                                                            ~~-~~~      ~
PALO VERDE UNIT 1 CYCLE 1'
                                                                                                                                                                                                                      ~~~
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                                                                    +0.22 x 10          4 Ap/*F                                                                                      Ijl! !. '                              I Q-                o            ij                                                                                                        !          I l l (s'I    'es *F,0.0apff) l
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h
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II  ,                        1 Z
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* T,,          l          I
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                                    !I      !
j                          !          l                                        l                      !l-! '5 l y,!!i!!ii!I1.c                        !    .
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i      ,      .. lll l!,I; i      llllllllllllllllllilllllllllllll!lli e                      sae                                                                                sse                                                                                    soe AVERAGE MODERATOR TEMPERATURE,*F                                                                                                                                                        ,
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                                                                                    ~*                                *'                                                                          e                      wm,
                                                                                                                                                                                                      -a.O    mm M &
 
T CONTROLLED BY USER 4/    REACTIVITY CONTROL SYSTEMS T                                                    -
s_-                          ~
MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4* Thi Reactor Coolant System lowest operating loop temperature (Teold) shall be greater than or equal to 552*F.
APPLICABILITY: MODES 1 and 2#*.
ACTION:
With a Reactor Coolant System operating loop temperature (Tcold) less than 552*F, restore Tg ,jg to within its limit within 15 minutes or be in HOT STANDBY within the next 15 minutes.
SURVEILLANCE REQUIREMENTS 4.1.1. 4 The Reactor Coolant System temperature (Tcold) shall be determined to be greater than or equal to 552*F:
    ,              a. Within 15 minutes prior to achieving reactor criticality, and b* At least once per 30 minutes when the reactor is critical and the Reactor Coolant System T cold is less than $57'F.
l
            #With K,g greater than or equal to 1.0.
            *$ee special Test Exceptien 3.10.5.
PALO VERDE - UNIT 1                    3/4 1 6 CONTROLLED BY USER
 
CONTROLLED BY USER L N REACTIVITY CONTROL SYSTEMS.
3/4.1*.2 BORATION SYSTEMS FLOW PATHS - SHUTOOWN LIMITING CONDITION F09 OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:                                                                            ,
: a. If only the spent fuel pool in Specification 3.1.2.Sa. is OPERABLE, a flow path from the spent fuel pool via a gravity feed connection and a charging pump to the Reactor Coolant System.
: b. If only the refueling water tank in Specification 3.1.2.5b. is OPERABLE, a flow path from the refueling water tank via either a charging pump, a high pressure safety injection pump, or a low pres-sure safety injection pump to the Reactor Coolant System.
APPLICABILITY: MODES 5 and 6.
ACTION:
With none of the above flow paths OPERA 8LE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.
SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required flow paths shall be demonstrated OPERA 8LE at least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
PALO VERDE - UNIT 1                      3/4 1-7 s**Q Y"D
                                              ^]1q%,
                                              ....4
                                                          '''} V 1 i f f',T)
                                                                  , . 4,
 
CONTROLLED BY USER
    ,  _ REACTIVITY CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three baron injection flow paths shall be OPERABLE:                                                  .
: a. A gravity feed flow path from either the ' refueling hater tank or the ed.        ' spent tuel " pool through CH-536 (RWT Gravity Feed Isolation Valve) and a charging pump to the Reactor Coolant System, M.,    b. Agravityfeedflowpathfromthe'r,efuelinghaterIankthrough
* CH-327 (RWT Gravity Feed / Safety Injection System Isolation Valve) and a charging pump to the Reactor Coolant System,
: c. A flow path from either the'r,efueling ' water' tank or the spent fuel pool throughtCH-164 (Boric Acid Filter Bypass Valve), utilizing h            gravity feed'and a charging pump to the . Reactor Coolant System.
L cti%f Wdc. ocW Filtu !.wuNm 'huc)
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two baron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours or be in at least HOT STAN08Y and borated to a SHUTOOWN MARGIN equivalent to at least 6% delta k/k at 210*F within the next 6 hours; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within the next 30 hours.        _
SURVEILLANCE REQUIREMENTS 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:
: a. At least once per 31 days by verifying that each valve (manual,
      .              power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.
: b. At least once per 18 months when the Reactor Coolant System is at normal operating pressure by verifying that the flow path required by Specification 3.1.2.2 delivers at least 26 gpm for 1 charging pump and 68 gpm for two charging pumps to the Reactor Coolant System.
PALO VERDE - UNIT 1                    3/4 1 0 Y
 
CONTROLLED BY USER y b* ' REACTIVITY x
CONTROL SYSTEMS ^
CHARGING PUMPS - SHUTDOWN                                      ;
LIMITING CONDITION FOR OPERATION 3.1.2.3 At least one charging pump
* or one high pressure safety injection pump or one low pressure safety injection pump in the boron injection flow path required OPERABLE pursuant to Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.
APPLICABILITY: MODES 5 and 6.
ACTION:
With no charging pump or high pressure safety injection pump or low pressure safety injection pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or            ;
positive reactivity changes.                                                            ;
SURVEILLANCE REQUIREMENTS 4.1.2.3 No additional Surveillance Requirements other than those required by            '
Specification 4.0.5.                                                                .
i 1
Whenever the reactor coolant level is below the bottom of the pressurizer in MODE 5, one and only one charging pump shall be OPERABLE, by verifying at least once per every 7 days that power is removed from the remaining charging pumps.
s PALO VER0E - UNIT 1                      3/4 1-9 CONTROLLED BY USER
 
                                                                                              ~
                                                              ~
CONTROLLED BY USER sb-
:    REACTIVITY CONTROL SYSTEMS''.
        <                                l CHARGING PUMPS - OPERATING el LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours or be in at least HOT STANOBY and borated to a SHUTDOWN MARGIN equivalent to at least 6% delta k/k at 210*F within the next              l 6 hours; restore at least two charging pumps to OPERABLE status within the next            l 7 days or be in COLD SHUTDOWN within the next 30 hours.                                    !
I i
I t
SURVEI'LLANCE REQUIREMENTS                                                                I l
i 4.1.2.4 No additional Surveillance Requirements other than those required                f by Specification 4.0.5.                                                                  i l
O t
PAL 0 VERDE - UNIT 1              3/4 1-10 CONTROLLED BY USER
 
l
                                                                                                ~
CONTROLLED BY USER                                                    !
                                                                                                      }
2 ''' . REACTIVITY CONTROL SYSTEMS'                                                                  j i l
80 RATED WATER SOURCES - SHUTOOWN l
LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERA 8LE:
: a. The spent fuel pool with:              M sy dh si /.'c' au.      3.!-2 . M
              @      1. A minimum borated water volume <jf33,500 gallonsAdh si 2. A boron concentration of between 4000 Tp3, and 4400 ppm boron, and
: 3. A solution temperature between 60*F and 180*F.
: b. The refueling water tank with:
u s p e_ W E. @ g a g 3,( ,2,
                @    1. A minimum contained borated water volume [of 33,500 gallon,s 1 and g d 2. A baron concentration of between 4000(ppm and 4400 ppm baron, and
: 3. A solution temperature between 60*F and 120*F.
h APPLICABILITY: MODES SI and 6S ACTION:
With no borated water sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until, at least one borated water source is restored to OPERABLE status.
SURVEILLANCE REQUIREMENTS                                                        ,
4.1.2.5 The above required borated water sources shall be demonstrated OPERABLE:
: a. At least once per 7 days by:
: 1. Verifying the boron concentration of the water, and
: 2. Verifying the contained borated water volume of the refueling water tank or the spent fuel pool.
: b. At least once per 24 hours by verifying the refueling water tank temperature when it is the source of borated water and the outside air temperature is outside the 60*F to 120*F range.
: c. At least once per 24 hours by verifying the spent fuel pool temperature when it is the source of borated water and irradiated fuel is present in the pool.
                                            ~
~{ ( "See Special Test Exception 3.10.7.
PALO VERDE - UNIT 1                    3/4 1-11    .
: e. n. .N.TP O D..-      .m T( 'J N. . ...
 
CONTROLLED BY USER                                                                                                                                                          ~
                                                                                  -U-  ,
N
__ i _                                                  _ . _ . . _                __                                  l 7  136* 6" (40K)    . .. .. .
                                                                                                            ._ . . r. .              _ _ _ _
                                                                  ,                      ::_ [                  .h ,
                                                                                                                                          ,-'~135' 1'OT(333iK)h                                                                        i 135'-6" (30K)                                                                                                                        4
                                                                ;                          _~.--~~---'_
                                                          .j                                    ~ _- :: -                                                                                      l                                                            '
                                                              ,        134' 6" (20K)                                                                                                      _-_'                                          ..
j                          _ J.:133* 3" (7.25K)- '
i                            ;- % _                        _ - -y 7                        l          133'-6" (10K)    22_*
c._          '
COLD S/D VOLUME                                                                                                                            ,
                                                            ;            t          0
                                                                                                            -                                                                                        ~ ~ - - - - -
0                        200                                                                    400                          600 AVERAGE REACTOR COOLANT SYSTEM TEMP., OF
                                                                                                                                                                                                                                                  \
                                                                                                                                                                                                                            .-                    \
I                                                                                                                                                                              _
80% . J.          600,000 GAL. (565 0F)                                                                                    -          -  690K
                                                        /
                                                        /
                                                                                                                                                                                                                        -  57MK 75%  -
573,744 G AL. (1200 F)                                                                                                .
                                                                                                                                                                                                                        . sank MINIMUM USEFijL-COLD S/D VOL PLUS                                                                                                VOLUME (1) 70%  .                      MARGIN                                                                                                    . 525K 9EQUIR ED IN THE RWT
                                          /                                                -
                                      ,'    RWT LEVEL                                                                                                                                                                  - 500K i      INSTRUMENT 65%
READING (1)                      ,
475K                          1 ESF VOL PLUS MARGIN
                                          \                                        0*.          (2)                ,                                                                      ,                              og                          f
(                                            ,
200                                                                    400                          fiOO AVERAGE RCS TEMPERATURE,0F s                        (1) THE TANK LEVEL AND VOLUME SHOWN ARE THE USEFUL
                                                \                          LEVEL AND VOLUME ASOVE THAT IN THE TANK 'alHICH
                                                  '\ 'N                      IS REQUIRED FOR VORTEX CONSIDERATIONS (2) DURING MODE 5 AND 6 ONE OF THESE BOR ATED SOURCES SHALL CONTAIN A MINIMUM OF 33,500 GALLONS (3) THIS VOLUME IS NOT REQUIRED OURING MODE 6
                                                                                                                  -                                                                          ~
FIGURE 3.12                                                                                                $ agc2c.,0f;, 6(4R, .
MINIMUM BORATED WATER VOLUMES                                                                                                                18 PALO VERDE - UNIT 1          .                              3/4 1-12 CONTROLLED BY USER
 
CONTROLLED BY USER
                              -~
l l
cd t. REACTIVITY CONTROL SYSTEMSN BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 Each of the following borated water sources shall be OPERABLE:
: a. The spent fuel pool with:
: 1. A minimum borated water volume as specified in Figure 3.1-2, and
                  *l 2. A baron doncentration of between 4000 Jp'pgand 4400 ppm boron, and
: 3. A solution temperature between 60*F and 180*F.
: b. The refueling water tank with:
: 1. A minimum contained borated water volume as specified in Figure 3.1-2, and
: 2. A boron concentration of between 4000 and 4400 ppm of boron, and
: 3. A solution temperature between 60*F and 120*F.
    @ APPLICABILITY: MODES 1, 2,93,@ and 4.2        1 ACTION:
: a. With the above required spent fuel pool inope'rable, restore the pool to OPERABLE status within 72 hours or be in at least NOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to at least 6% delta k/k at 210*F, restore the above required spent fuel pool to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours.
                                                                                                ~
: b. With the refueling water tank inoperable, restore the tank to OPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
SURVEILLANCE REOUIREMENTS 4.1.2.6 Each of the above required borated water sources shall be demonstrated OPERABLE:
: a. At least once per 7 days by:
: 1. Verifying the boron concentration in the water, and
: 2. Verifying the contained borated water volume of the water source.
: b. At least once per 24 hours by verifying the refueling water tank temperature when the outside ' air temperature is outside the 60*F to 120*F range.
: c. At least once per 24 hours by verifying the spent fuel pool temperature when irradiated fuel is present in the pool.
            ~
t
    @ See Special Test Exception 3.16~. {
PALO VERDE - UNIT 1                  3/4 1-13 f M .h. lT P. f ..t @. 9 -Y l M W.- .
 
I CONTROLLED BY USER g ,y REACTIVITY CONTROL SYSTEMS;
            ~
BORON DILUTION ALARMS LIMITING CONDITION FOR OPERATION 3.1.2.7 Both startup channel high neutron flux alarms shall be OPERABLE.
APPLICABILITY: MODES 3*, 4, 5, and 6.
ACTION:
: a. With one startup channel high neutron flux alarm inoperable:
: 1. Determine the RCS baron concentration when entering MODE 3, 4, 5, or 6 or at the time the alarm is determined to be inoperable.
From that time, the RCS boron concentration shall be determined
              , 2,          at the applicable monitoring frequency _in Tab.le 3.1-Irby either        -
              ;2 b.
boronometer or RCS sampling.** ~~            i,    ,
{jQ4 "" " '
                                                                                                  ' ~'
With both startup channel high neutron flux alarms , inoperable:
: 1. Determine the RCS boron concentration by either boronmeter and RCS sampling ** or by independent collection and analysis of two RCS samples when entering Mode 3, 4, or 5 or at the time both alarms are determined te be inoperable. From that time, the RCS boron concentration shall be determined at the applicable monitoring frequency in Tables 3.1-1 through 3.1-5, as applicable,
(                            by either baronmeter and RCS sampling ** or by collection and l                            analysis of two independent RCS samples. If redundant determina-l                            tion of RCS boron concentration cannot be accomplished immediately, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until the method for determining and confirming RCS boron concentration is restored.
: 2. When in MODE 5 with the RCS level below the centerline of the hotleg or MODE 6, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until at least one startup channel high neutron flux alarm is restored to OPERABLE status.
: c. The provisions of Specification 3.0.3 are not applicable.
        . SURVEILLANCE REQUIREMENTS 4.1.2.7 Each startup channel high neutron flux alarm shall be demonstrated OPERABLE by performance of:
              *Within 1 hour after the neutron flux is within the startup range following a reactor shutdown.
  #' S k**Withoneormorereactorcoolantpumps(RCP)operatingthesampleshouldbe 4 obtained from the hot leg. With no RCP operating, the sample should be obtained from the discharge line of the low pressure safety injection (LPSI) pump operating in the shutdown cooling mode.
l PALO VERDE - UNIT 1                  3/4 1-14 l                              CONTROII En RY IICFD
 
CONTROLLED BY USER s
y., REACTIVITY CONTROL SYSTEMS                                                    * '
SURVEILLANCE REQUIREMENTS (Continued) i l
I          a. A CHANNEL CHECK:
l                  1. At least once per 12 hours.
i                  2. When initially setting setpoints at the following times:
a)    One hour after a reactor trip.
b)    After a controlled reactor shutdown: Within 1 hour after the neutron flux is within the startup range in MODE 3.
: b. A CHANNEL FUNCTIONAL TEST every 31 days of cumulative operation during shutdown.
l l
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i l
I i
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PALO VERDE - UNIT 1                3/4 1-15 CONTROLLED BY USER
 
                                                                                                                                      ~
I CONTROLLED BY USER                                                                            !
TABLE 3.1-1 A
ll REQUIRED MONITORING FREQUENCIES _FOR' BACKUP-BORON                                              !
                            /-            DILUTION DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT ,0PERATIONAL MODES FOR K,ff > 0.98 i
Number of Operatino Charoing Pumps 4 ef                    OPERATIONAL MODE                      O            1                    2          3        \
7 3                    12 hours      I hour            Operation not allowed l                        4                    12 hours      I hour            Operation not allowed                        .
t                        5 RCS filled            8 hours  1 hour            Operation not allowed I                                                                                                                              I i                      5 RCS partially                                                                                      '
drained                    Operation not allowed i                                                                                                                        i 6                    24 hours    8 hours            4 hours          2 hours
                          'N
                                                                                                              /
                                ~
                                                                                                          /
v\ '
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PALO VERDE - UNIT 1                          3/4 1-16 l
l
  ~
CONTROLLED BY USER
 
P' f"
s i
TABLE 3.1-1:.' MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DETECTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR                        '
Kg > 0.98.FOR MODES 3, 4, AND 5                                            ,
          ,                                                                                                                                              \
                                                                                                      ~
l TIME PERIOD IN HOURS AT WHICH M0NITORING IS REQUIRE 0 j OPERATIONAL
        '                                                            WITH THE FOLLOWING-~RUMBER OF CHARGING PUMPS OPERATING          -
                                                                                                                                                          \~  g MODE 0                      1                      2                        3 1
3                                              12.0                    1.0 f(HotStandby)                                                                                                          OPERATION NOT ALLOWED            .
I i
4
                                                                                                                                                            /
12.0
      !                                                                                          1.0                        OPERATION NOT ALLOWED w i        ( Hot Shutdown )
S'                    5                                                8.0 1.0 T' '      ( Cold Shutdown )                                                                                                OPERATION NOT ALLOWED        I C                                                                                                                                                        {
'                        5                                                                                                                                    !
OPERATION NOT ALLOWED                                  j
( RCS PARTIALLY DRAINED FOR SYSTEM REPAIRS )                                                                                                                        l i
6*
24.0                    8.0                    4.0 i
:( Refueling )                                                                                                                        2.0 /
o' l
A conservative value of an initial baron concentration is assumed which is bounded by the Technical                                I Specification 3.9.1.
Furthermore, during refueling the LPSI pumps should be used for any makeup operation!
If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used. ,
g.
[a ,I (fi sa N                                                                                                                                                                .
 
l CONTROLLED BY USER                                                                    z              l 1!
g^--- -                  _ TABLE 3.1-2
                                                                                                                  ^
                                                                                                                      ' I( w'
                                                                              ~ _ _ _ -                                  '
REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION OETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS AND PLANT                                i OPERATIONAL MODES FOR 0.98 > K,ff > 0.97
                /
                                                                                                                                \
j                                                                                                                  -;
OPERATIONAL Number of Operating Charging Pumos l                                                                                                                        \
s ho
<          !              MODE                  O            1                        2                        3                  \
3                  12 hours      2.5 hours              I hour                  0.5 hours                  !
4                  12 hours      2.5 hours              I hour                  0.5 hours i
5 RCS filled        8 hours      2.5 hours              I hour                  0.5 hours                l 5 RCS partially                                                                                            l l
drained ,        8 hours    0.5 hours                  Operation not allowed 6                  24 hours    8 hours                4 hours                  2 hours 1
                      \                                                                                                    /
                                                                  ~ . - ~ - . . . . _ . , _ , _ , _ _ _ _    ...e*#"
l l
9 PALO VERDE - UNIT 1                    3/4 1-17 CONTROLLED BY USER
 
e/0 TABLE 3.1-2: ' MONITORING FREQUENCIES FOR BACKUP BORON DILUTION DtIECTION FOR SYSTEM 80 EArtNDED FUEL CYCLE FOR                      '
                                                                                        / 0.98 2 K              > 0.97 FOR riuuES 3, 4, AND 5                        's i
i l                                                              TIME PERIOD IN HOURS AT WHICH MONITORING IS REQUIRED i OPERATIONAL                                                  WITH THE FOLL0llH F RUPBER OF CHARGING PLMPS OPERATING i
                                                                                                                                                                                            \
{ MODE 0                              1                2                                        3 3                                                  12.0                            2.5              1.0 0.5 l(HotStandby)                                                                                                                                                              ,!
4                                                  12.0                            2.5              1.0                                    0.5
( Hot Shutdown )                                                                                                                                                j w              !
(
,  5                                      5                                                    8.0                          2.5              1.0                                    0.'S            i l
y                ( ( Cold Shutdown )                                                                                                                                                      ;
  . ;;;              \                                                                                                                                                                      '
;                                          5                                                    8.0                          0.5                    OPERATION NOT'. ALLOWED i
                      ;(RCSPARTIALLYDRAIN(D                                                                                                                                                  -
g      FOR SYSTEM REPAIRS )                                                                                                                                            j i
I                                                                                                                                                                                                i i              6*                                                  24.0                            8.0                4.0
( Refueling )                                                                                                                                              2.0 j t
t i ',
The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and
,                                  , plant condition.                                                                                                                                                    ,
i                                                                                                                              -
                                                                                                                                                                                                  }
j        y;y
* i      A conservative value of an initial boron concentration is assumed which is bounded by the Technical
,        g5                        Specification 3.9.1.      Furthennore, during refueling the LPSI pumps should be used for any makeup oper g 8.                      If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used.                                            j E!
,      ?=
E' 4
                                                                                                                                                                  . ~ . , . . ~ ~ . . ~
 
CONTROLLED BY USER                                                , e. .
I TABLE 3.1-3
                                                                                                .y n'        ,
                                /    REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION f
                            /            DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS
                          ,/              AND PLANT OPERATIONAL MODES FOR 0.97 > K,ff > 0.96                                s j'                                                                                                      \
1-              I
                      /                                                                                                          \  \
Number of Oceratina Charging Pumps                                      *
[      OPERATIONAL MODE                    O          1                  2                    3                          1 l
                !                  3                  12 hours    3.5 hours        1.5 hours                I hour                        {
3.5 hours i
            /                      4                  12 hours                      1.5 hours                I hour l
l
[                        5 RCS filled          8 hours  3.5 hours        1.5 hours                I hour                      /
                                                                                                                                          /
          ;                    _ 5 RCS partially                                                                                      i
          \                        drained            8 hours  I hour              Operation not allowed                        ,
                '                                                                                                                  /
s              6                  24 hours    8 hours          4 hours                  2 hours          ,
                                                                                                                              /
                                                                                                  ,/
l                                                                                                -
PALO VERDE - UNIT 1
* 3/4 1-18
: l.  .
CONTROLLED BY USER
 
                                                                                                                                                    ~ '            '
TABLE 3.1-3: MbNITORING FREQUENCIES FOR BACKUP BORON DILUTION 'N DtitCTION FOR SYSTEM 80 EXTENDED FUEL CYCLE FOR 0.97 2                K_,, > 0.% FOR N0 DES 3, 4, AND 5
                                                                                                                                                                                              .s
                                                                                                                                                                                                  'N 4
;                                                      OPERATIONAL                                            TIME PERIOD IN HOURS AT WHICH MONITORING IS REQUIRE 0                                        i
* MODE                                                  WITH THE FOLL0llTRG RUMBER OF CHARGING PUMPS OPERATING                    ,
I O                                    1                  2                                  I 3
i 3                                                12.0                                3.5                  1.5                                1*0
( Hot Standby )
4                                                12.0                                3.5                  1.5                                1*0
( Hot Shutdown )
( Cold Shutdown )
                                                                                                                                                            -                    .                                1.0 1                      5                                                                                                                                            l e
8.0                                1.0                      OPERATION NOT ALLOWED (RCSPARTIALLYDRAINgD FOR SYSTEM REPAIRS )
6*                                                24.0                                8.0                                            '
( Refueling )                                                                                                          4.0                        2~0 l
t The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and plant condition.
yg
* A conservative value of an initial boron concentration is assumed which is bounded by the Technical gg        Specification 3.9.1. Furthermore, during refueling the LPSI pumps should be used for any makeup operation.
g o.      If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used.
Q  -
m                                                                                    .
                                                  .~ E GE
: n.                                                                                                                                          -b i W
 
                                                                                                                                      \
CONTROLLED BY USER                                                  < /;
r; I
l
                            ,7 TABLE 3.1-4                      [
                                                        " ~ . , . _              .
                                                                                      ./_
                        / REQUIRED MONITORING FREQUENCIES FOR" BACKUP' BORON DILUTION
                      /          DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS
                    /            AND PLANT OPERATIONAL MODES FOR 0.96 > K,ff > 0.95                                                ,,
                  /
8 b*
l l                                                                                                    '
Number of Ooerating Chargina Pumos                            i
            /      OPERATIONAL                                                                                            ,
          /            MODE                          O        1                      2          3                          i i
3                  12 hours        5 hours            2 hours.        I hour                        .
4                  12 hours        5 hours            2 h'ours        I hour 5 RCS filled              8 hours  5 hours            2 hours          1 hour                          l 5 RCS partially                                                                                      i drained                  8 hours  1.5 hours            Operation not allowed 6                  24 hours        8 hours            4 hours          2 hours          ,
                                                                                                                /
                                                                                                              /
  \
6
                                                                      ,  _,,___s-"
l PALO VERDE - UNIT 1                        3/4 1-19 l
1 CONTROLLED BY USER
: n. - _ . _. . ._                    _ =n    w -    - - - - - - = = - - - -
                                                                                                              , _ , , . ~ . , -- --* ~ ,
p-                                          ~x, '-
[y .
s 's.      %~
TABLE 3.1-4        MONITORING FREQUENCIES FOR BACKUP BORON DILUTION                                A
                                                                              /  DtitCTION FUR 5Y5ILM 80 LXItNDED FUEL CYCLE PUR
                                                                          /      0.96 > K ,, > D.95 FOR swuES 3, 4, AND 5
                                                                                                                                                                            \  \
                                                                                                                                                                                  \
                                                                                                                                                                                    \
l
                        /                                                                                                                                                              \
t OPERATIONAL                                          TIME PERIOD IN HOURS AT WHICH MONITORING IS REQUIRED                                .
WITH THE FOLLO6fDRl~NUpBER OF CHARGING PUMPS OPERATING MODE O                                  1                          2                          3 3                                          12.0
( Hot Standby )                                                                            5.0                -
2.0                        1.0 4                                        12.0                                  5.0                        2.0                        1.0 m                        ( Hot Shutdown )
_,                                    5                  -
8.0                              5.0                        2.0                        1.0 o
4                        ( Cold Shutdown )
5                                            8.0                              1.5
{(RCSPARTIALLYDRAIN(0 FOR SYSTEM REPAIRS )
OPERATION NOT ALLOWED 6*                                          24.0                                  8.0
( Refueling )
4.0                      2.0 I*
The Technical plant    condition.Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and                                          !
                                                                                                                                                                                      /
7 3 k g*x          A  conservative value of an initial boron concentration is assumed which is bounded by the Technical Specification 3.9.1.                                                                                                                            !
1                                              Furthermore,    during  refueling    the LPSI    pumps      should be used for any  makeup  operation's
: 2. g 'Ns If it is necessary to use the charging pumps, the appropriate monitoring frequency above should be used.. ~
e a=  t+                                                                                                                                                .
2                                      ' ' '          *
                                                                                  - -'- ~ -~- *"                --
k+  'c
                                                                                                                                                                                          ~
 
CONTROLLED BY USER                                              @
_ TABLE 3.1-5                        CP              -
                            /,- -
                                                                  ~ ____..
y                N
                  , f REQUIRED MONITORING FREQUENCIES FOR BACKUP BORON DILUTION                        '
                /              DETECTION AS A FUNCTION OF OPERATING CHARGING PUMPS
              /                    AND PLANT CPERATIONAL MODES FOR K,ff < 0.95                              \
            /                                                                                                  'N            .
        /                                                                                                      \          !
                                                                                                                    \
he /          OPERATIONAL                          Number of Ooeratino Charcino Pumps
      /            MODE                        O          1                  2          3 I                                                                                                                ,  i
                                                                                                                        \  ;
l 3                    12 hours    6 hours          3 hours          1.5 hours                    ' '
4                    12 hours    6 hours          3 hours          1.5 hours                      .
    \
      \              5 RCS filled          8 hours    6 hours          3 hours          1.5 hours 5 RCS partially drained            8 hours    2 hours
* Operation not allowed                  /
                                                                                                                  /
                                                                                                                    /
6                    24 hours    8 hours          4 hours          2 hcurs s
1                                                                        .
i                                          .
PALO VERDE - UNIT 1                        3/4 1-20 CONTROLLED BY USER
 
                                                                                                        ~
                                                                    ..-                                        y
                                                                ,/                                                    %. _                      .
TABLE 3.1-5: MDNITORING FREQUENCIES FOR BACKUP BORON DILUTION Dtit0 TION rou SY5f tn 80 LXItnuto FUEL LyuE tuu                    N    '
                                                  / K_,, .<  0.95 FOR nuuES 3, 4. Alm 5 f..
                  /                                  TIME PERIOD IN HOURS AT lef!CH MONITORING IS REQUIRED OPERATIONAL                            WITH THE FOLLO G lD hstER OF CHARGING PUMPS OPERATING N00E O                        1                      2                                    3 3                              12.0                      6.0                      3.0                                  1.5
( Hot Standby )
4                              12.0                      6.0                      3.0                                  1.5
            .( Hot Shutdown )
R
* 5                                8.0                      6.0                      3.0                                  1.5 7            ( Cold Shutdown )
no 5                                8.0                      2.0                          OPERATION NOT ALLOWED (RCSPARTIALLYDRAly)D FOR SYSTEM REPAIRS 6*                                24.0                      8.0                      4.0
( Refueling )                                                                                                                      2.0 i
          ,      The Technical Specification 3.1.2.3 will allow operation of only ONE charging pump during this MODE and plant condition.
    ,, m.
* A conservative value of an initial boron concentration is assumed which is bounded by the Technical fr E          Specification 3.9.1. Furthermore, during refueling the LPSI pumps should be used for any makeup operation.
g              If it is necessary to use the charging pumps.. the appropriate monitoring frequency above should be used.
S Oz                                                -
  '_. 5 E%
 
CONTROLLED BY USER gd,.      TIVITY CONTROL SYSTEM                                                                ,
3/4.1.3 MOVABLE CONTROL ASSEMBLIES CEA POSITION LIMITING CONDITION FOR OPERATION 3.1.3.1 All full-length (shutdown and regulating) CEAs, and all part-length CEAs which are inserted in the core, shall be OPERA 3LE with each CEA of a given group positioned within 6.6 inches (indicatar._cosition) af all othme CEAs in its oroup. JIn addition, the position of tae part length CEAs Groups
    @ Ghall be limited to the intertina 14=4+= <hnwn in Figure 3.1-2A.
APPLICABILITY: MODES la and 2*.
ACTION:
: a. With one or more full-length CEAs inoperable due to being immovable
                      .as a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTOOWN MARGIN require-ment of Specification 3.1.1.1 is satisfied within 1 hour and be in at least HOT STANDBY within 6 hours.
: b. With more than one full-length or part-length CEA inoperable or misaligned from any other CEA in its group by more than 19 inches (indicated position), be in at least HOT STANDBY within 6 hours.
  .            c. With one or more full-length or part-length CEAs taisaligned from any other CEAs in its group by more than 6.6 inches, operation in MODES 1 and Z_mayJQD11.n_ue,_puovided n                  that C5ta:powee-is-reducyd 1FTc6d
              @ C%Tth Ficure 11-28 and thaT,within 1 hour the misaligned scEA@fng either:                        ~
: 1. Restored to OPERABLE status within its above.specified alignment requirements, or
: 2. Declared inoperable and the SHUTOOWN MARGIN r2quirement of Specification 3.1.1.1 is satisfied. After declaring the CEA
                          ' inoperable, operation in MODES 1 and 2 may continue pursuant to the reqJirements of Specification 3.1.3.6 provided:
a)  Within 1 hour the remainder of the CEAs in the group with l                                the inoperable CEA shall be aligned to within 6.6 inches of the inoperable CEA while maintaining the allowable CEA sequence and insertion limits shown on Figure 3.1-3; the THERMAL POWER level shall be restricted pursuant to            ,
i                                Specification 3.1.3.6 during subsequent operation.
!        "See Special Test Exceptions 3.10.2 and 3.10.4.
(
PALO VERDE - UNIT 1                        3/4 1-21 I
l l                              CCFTROldED P' USER L
 
CONTROLLED BY USER                                                                                    :
g 4 Q TIVITY CONTROL SYSTEMS, 7                                                                                                  -
LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued) b)  The SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours.
l                  Otherwise, be in at least HOT STANDBY within 6 hours.
i i
: d. With one full-length CEA inoperable due to causes other than addressed by ACTION a., above, and inserted beyond_the Long Term
              @ Steady State Insertion Limits (Figures 3.1-3 cand 3.1-4) but within its above specified alignment requirements, operation in MODES 1 and 2 may continue pursuant to the requirements of Specification 3.1.3.6.
: e. With one part-length CEA inoperable and inserted in the core, operation may continue provided the alignment of the inoperable part length CEA is maintained within 6.6 inches (indicated position) of all other part-length CEAs in its group.
SURVEILLANCE REQUIREMENTS
(        4.1.3.1.1 The position of each full-length and par +,-len determined to be within 6.6 inches (indicated position) gth    CEA of all    shall other      CEAs                    be in its group at least once per 12 hours except during time intervals when one CEAC is inoperable or when both CEACs are inoperable, then verify the individual CEA positions at least once per 4 hours.
4.1.3.1.2 L -5 full-length CEA not fully inserted and each part-length CEA which is int *ed in the core shall be determined to be OPERABLE by movement of at least 5 inches in any one direction at least once per 31 days.
l l
i I
l      PALO VERDE - UNIT 1                  3/4 1-22 l
i s                      .  .
u
 
                                                                                                                                                                                            ~
a f
i l                                                                                              FIGURE 3.12A p
I        g                                                                                                                                              ('-g,>)
PART LENGTH CEA INSERTION LIMIT vs THERMAL POWER o
S g                  ,/                    .
s
                              ,-                IM        i    i    s        a      a                i      s      i        s      :    i    i    :
E            /        .                                      112.5"          ,
4          /                    0.90 O
g                                                                                                                                                            O
  - h                    N                  y f                  /h                  2 0.80 -
ACCEPTABLE
                                                                                                    'e                                            .                  -
O fy            ,      Y 0.70 -
OPERATION
* UNACCEPTABLE
                                                                                                                                                                    ~
{
q y                    n                  .g                                                                                      OPERATION
                                                                                                                                                                                  .y 5
0 W          $        $ m 5
O 0"  -
o
* OM                                                                  50% POWER LINE M                ,
Q E                g        -
h
                                                                                                                                                                      .                F, y-T        ;! E              <
r:1 "w          m l
                  !    o m                E                              \ INSERTION LIMIT                          .
                                                                                                                                                                                  <m J              !
2 o            g    0.40 -
                                                                                                                                                                    -          Im    %#
14                    ET                o                                                                                                                                          W q                    4 Bt              g    030  -
4 5                  <
WI                    w                  [    0.20                '
C m              i (n
d                                            0 10 -
o 22.5" 0.00      '    '    '      '        '        '      '      '      '        '      '    '    '    '
150    140  130  120  110        100      90      80    70      60      50    40    30    20    10
                      \                                                                                                                                                0 PART LENGTH CEA POSITION, INCHES WITHDRAWN
                                          ~
                                                                                                                            ,Y
                                                                                    *"' . - .        ~
O
                                                                                      ^
                                                                                                                                                                .m          eeem + < mer
 
1 CONTROLLED BY USER i
I FIGUR E 3.128
      #.                    CORE POWER LIMIT AFTER CEA DEVIATION
* 2                                                                                              -
9
                  -a bI E2
                                          .'                I 5i      20    - - - - -
i i
(60 MIN,20%)
1
                  -w a    10    --.*-                .-        -
t'*~
: i.                        .          !                                t bE
* y[
o 0
t
[l - - - -
e t
ll- - -- -f - -
t '    e 0        10            20        30          40      50    60 g 3R Z.                  TIME AFTER DEVIATION, MINUTES 2                                          -
I
              *WHEN CORE POWER IS REDUCED TO 55% OF RATED THERMAL                                                1 POWER PER THIS l.lMIT CURVE, FURTHER REDUCTION IS NOT REQUIRED t
        \
          \,
N FIGURE 3.1-28 CORE POWER      , ,
LIMIT AFTER CEA DEVIATION *
                                                                    - - _ . _ _ _      y PALO VERGE - UNIT 1                                        3/4 1-24 e()N.TPQ, 1.IB. P, f l,icER  .                  .          .        ,
 
                                                                                                                  ~
CONTROLLED BY USER ye w  REACTIVITYCONTROLSYSfEMS a
POSITION INDICATOR CHANNELS - OPERATING LIMITING CONDITION FOR OPERATION l
3.1.3.2 At least two of the following three CEA position indicator channels            !
shall De OPERABLE for each CEA:                                ,      .
tuelecteJ                l
                                #g'          a. CEA Reed Switch Position Transmitter (RSPT 1),with the capability of        I determining the absolute CEA positions within 5.2 inches,                  !
b.
Lth sh'es CEA Reed Switch Position Transmitter (RSPT 2) with the capability of Ob'                determiningtheabsoluteCEApositionswithinI.2 inches,and
: c. The CEA pulse counting position indicator channel.
* APPLICABILITY: MODES 1 and 2.
ACTION:
With a maximum of one CEA per CEA group having only one of the above required CEA position indicator channels GPERABLE, within 6 hours either:
: a. Restore the inoperable position indicator channel to OPERABLE status, or
: b. Be in at least HOT STAND 8Y, or
: c. Position the CEA group (s) with the inoperable position indicator (s) at its fully withdrawn position while maintaining the requirements of Specifications 3.1.3.1 and 3.1.3.6. Operation may then continue provided the CEA group (s) with the inoperable position indicator (s) is maintained fully withdrawn, except during surveillance testing pursuant to the requirements of Specification 4.1.3.1.2, and each CEA in the group (s) is verified fully withdrawn at least once per l                                                12 hours thereafter by its " Full Out" limit."
                                                                                ~
SURVEILLANCE REQUIREMENTS 4.1.3.2 Each of the above required position indicator channels shall be determined to be OPERABLE by verifying that for the same CEA, the position indicator channels agree within 5.2 inches of each other at least once per 12 hours.                    -
                                      "CEAs are fully withdrawn (Full Out) when withdrawn to at least 144.75 inches.
A                            (n3c.ec.
i l
PALO VERDE - UNIT 1                3/4 1-25 i
CONTROLLED BY USER e                                                                                                - - -
 
CONTROLLED BY USER m
g h' . ' , REACTIVITY CONTROL .s      SYSTEMS s                                              '
POSITION INDICATOR CHANNELS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 At least one CEA Reed Switch Position Transmitter indicator channel shall be OPERABLE for each shutdown, regulating or part-length CEA not fully inserted.                                .
APPLICABILITY: MODES 3*, 4*, and 5*.
ACTION:
With less than the above required position indicator channel (s) OPERABLE, immediately open the reactor trip breakers.
SURVEILLANCE REQUIREMENTS 4.1.3.3 The above required CEA Reed Switch Position Transmitter indicator channel (s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.
* l l
i "With the reactor trip breakers in the closed position.
PALO VERDE - UNIT 1                    3/4 1-26 CONTROLLED BY USER t
 
                                                                                            \
                                                                                          ! i CONTROLLED BY USER
                              -~
REACTIVITY CONTROL SYSTEMS; eL    ..-                            -
CEA OROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full-length (shut'down and regulating) CEA , drop time, 06- from a fully withdrawn position, shall be less than or equal to 42 seconds from when the electrical power is interrupted to the CEA drive mechanism until the CEA reaches its 90% insertion position with:
: a. Tcold greater than or equal to 552*F, and
: b. All reactor coolant pumps operating.
APPLICABILITY: MODES 1 and 2.
ACTION:                                        -
: a. With the drop time of any full-length CEA determined to exceed the above limit, restore the CEA drop time to within the above Ifmit prior to proceeding to MODE 1 or 2.
SURVEILLANCE REQUIREMENTS 4.1.3.4 The CEA drop time of full-length CEAs shall be demonstrated through measurement prior to reactor criticality:
: a. For all CEAs following each removal and reinsta11ation of the reactor vessel head,
: b. For specifically affected individual CEAs following any maintenance on or modification to the CEA drive system which could affect the drop time of those specific CEAs, and
: c. At least once per 18 months.
PALO VERDE - UNIT 1                  3/4 1-27 COWRO'J,ED YPf 11SR
 
CONTROLLED BY USER (REACTIVITY CONTROL SYSTEMS w                        ,')
SHUT 00WN CEA INSERTION LIMIT LIMITING CONDITION FOR OPERATION
  'g,        3.1.3.5 All shutdown CEAs shall be withdrawn to at least 144.75 inches  g { M3 !5xed APPLICABILITY: MODES 1 and 2*#.
ACTION:
I
      #}'    With a maximum of one shutdown CEA withdrawn to less than 144.75 inches except for surveillance testing pursuant to Specification 4.1.3.1.2,    ^
within 1 hour either:
y,          a. Withdraw the CEA to at least 144.75s inches, or
: b. Declare the CEA inoperable and apoly Specification 3.1.3.1.
SURVEILLANCE REQUIREMENTS Og'    4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to at least 144.75 inctes:U'13 sNPS)
: a. Within 15 minates prior to withdrawal of any CEAs in regulating groups during an approach to reactor criticality, and
: b. At least once per 12 hours thereafter.
              "See Special Test Exception 3.10.2.
              #With K,ff greater than or equal to 1.
PALO VERDE - UNIT 1 3/4 1-28 CONTROLLED BY USER
 
CONTROLLED BY USER                                                          :
e A ' ., REACTIVITY CONTROL SESiEMS' REGULATING CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION                                                                ,
: 21. - -> 3.1. 3. 6 The regulating CEA groups shall be limited to 69    Gand_.to ervice            or'shown on FiguVe~3 the insertion limits ## shown_o0J_lgure 3.1-3 1s not.in1-4**      when, the CO hen the COLS$ is in' service. The(CEA
  ,)'                                          ~
_ insertion between the~Long Ters Stea'dy State' Insertion Limits and the Trans -pLtk l
tent Insertion Limits is restricted to:                      .
: a.      Less than or equal to 4 hours per 24 hour interval,
: b.      Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day interval, and
: c.      Less than or equal to 14 Effective Full Power Days per 18 Effective Full Power Months.
APPLICABILITY: MODES la and 2*#.
ACTION:
: a. With the regulating CEA groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 2 hours either:
: 1. Restore the regulating CEA groups to within the Ifmits, or
* 2. Reduce THERMAL POWER to less than or equal to that fraction ~of RATED THERMAL-POWER-which is allowed by the CEA group positi,on E)      using' Figures 3.1-3 or 3.1;4f ?a af.% fi?t4
: b. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals greater than 4 hours per 24 hour interval, operation may proceed provided either:
n    1. The Short_ Term Steady State Insertion Limits of Figure 3.1-3
                        @          or Figure 3.1-4'are not exceeded, or
: 2. Any subsequent increase in THERMAL POWER is restricted to less than or equal to 5% of RATED THERMAL POWER per hour.
                "See Special Test Exceptions 3.10.2 and 3.10.4.
                #With K,f greate                                                        --
d      '>      **CEAs oare fully,          w,r inthan ithdrawn          or equal accordance        to 1. 3.1-3.or Figure 3.1-4_ when with Flgure        .
d d , -->          withdrawn to at least_144.75 inches / 2 " f5
                ##fbwing a reactor power cutback in which (1) Regulating Group 5 or Regulating Groups 4 and 5 are dropped or (2) Regulating Group 5 or Regulating Groups 4 A9
    ~~          fand5aredroppedandtheremainingRegulatingGroups(Groups                        1,2,3,and4) sequentially inserted, the Transient Insertion Limit of Figure 3.1-3          or (F_igure3.1-4canbeexceededforupto2 hours.
PALO VERDE - UNIT 1                      3/4 1-29 CONTROLLED BY USER
 
REGULATING CEA INSERTION LIMITS l
i LIMITING CONDITICN FOR OPERATION l
3.1.3.6 The regulating CEA groups shall be ligted to the withdrawal sequence, specified overlap, and to the insertion limits shown on Figure 3.1-3**, with CEA insertion between the Long Term Steady Sute Insertion Limits and the Transient Insertion Limits restricted to:
: a. Less than or equal to 4 hours per 24 hour interval,
: b. Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day interval, and
: c. Less than or equal to 14 Effective Full Power Days per 18 Effective Full Power Month.
AppLICA8ILITY: MODES 1* and 2*d.
ACTION:
: a. With the regulating CEA groups inserted beyond the Transient Inser-tion Limits, except for surveillance testing pursuant to Specifica-tion 4.1.3.1.2, within two hours either:
: 1. Restore the regulating CEA groups to within the limits, or
: 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by.the CEA group position i                      using the above figure.                          '
: b. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals greater than 4 hours per 24 hour interval, operation may proceed provided either:
: 1. The Short Term Steady State Insertion Limits of Figure 3.1-3 are not exceeded, or
\
See special Test Exceptions 3.10.2 and 3.10.4, With K,ff greater than or equal to 1.0.
CEA's are considered fully withdrawn in accordance with Figure 3.1-3 when withdrawn to at least 144.75 inches (193 steps).
H      Following a reactor power cutback in which (1) Regulating Group 5 is          *
                                                                                        /.,
dropped or (2) Regulating Groups 4 and 5 are dropped and for cases (1) and (2) should the remaining pegulating Groups (Group 1, 2, 3, and 4)
      ! be sequentially inserted, the Transient Insertion Limit of Figure 3.1-3 can be exceeded for up to 2 hours. Also for cases (1) and (2), the
      ; specified overlap between Regulating Groups 3, 4 and 5 can be exceeded for up to 2 hours.
Amendment ,'tumber 9 3/4 1-29                      February 27,1984
 
CONTROLLED BY USER                                                ll REACTIVITY CONTROL SYSTEMS ACTION: (Centinued)
: c. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFP0 per 30 EFPD interval or greater than 14 EFPD per 18 Effective Full Power Months, either:
: 1. Restore the regulating groups to within the Long Term Steady State Insertion Limits within 2 hours, or
: 2. Be in at least HOT STANOBY within 6 hours.
SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits at least once per 12 hours except during time intervals when the PDIL Auctioneer Alarm Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulated times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at least once per 24 hours.
9 e
PALO VERDE - UNIT 1                3/4 1-30 CONTROLLED BY USER
 
i CONTROLLED BY USER
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                                                    'N N
TRACTION OF RATED THERMAL POWER                                                      /
h Mulia.A  Ll
                                                                                                                                                                                          . inR
      .                                                                                            FIGURE 3.1-3                                            ;.                    ,
CEA INSERTION LIMITS VS THERMAL POWER e(COLSS IN SERVICE)>                    -
l              PALO VERDE - UNIT 1                                                                    3/4 1-31 i                                                                    CONTROUJiD BY IJSER e
 
CONTROLLED BY USER                                                                              '
                                                                                          ~~''                                            ..
                                                                                                              ~*
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* l I I i i                I I I I i 1 .1              1I        I      I I        ,    e l                      1i      '    l l      l l l l 1                i i    'I l I i I                            i        i i      i l I i            i  i l i ll Il I i f                            I        l I l l 6 i i l i I i i i                              11 l
                                                                                                              ~
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il lI I I I I I l i I
      }                            l l                                                      l -l        I
                                                                                                              ~
a      e I                      l i 11 l l l l l l l l l l l                                  11      I l          U~    "
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                              ! l I I I I I I I I I I I l i I Il l l                                            ~'O~
* I i l I I I i l i l i l i i i i              b l 1 l l 1.
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  'Eg 4
l 1 i 1 li I I I l l l l l i l i I lz u l l l l lhil i i l              i l l/
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                                                                                                                    ,,    e
                                                                                                                          - m I lli l I i l i i - 1 I I i I i I/I                                                  _    e 7 1      I I I I l l ll ($1 i i l i i i /I I i i i i i i i i lEl I i i i i I /l l
                                                                                                              ~
3"O        .s 5
I l i l i I I I i l~l i I i ! l 1/1 1
                                                                                                              ~
3~*
h I                    I l i I I I              I i i Ib              I I    a i I/ I I            ,
g, e      E i                      i i i l I i i 1 i E I I I I 1/l I i                                                -
e :s l                          1i          l i    I        1 3ll            .I (l l I            ,,  e ,, *e g i                          i i          l                  1 M        i  l J/      I I    i        -
          \                    i l i I l
                                                              -1 l N ift                  i i
i
                                                                                                                ~*~      e i                  11 I i l I            ,
I I I i i IV I i                        i i    i l l l l lil i I l if i                                '
I  I  I          a      =  _
                            . L I I I I I i                      i I/I I I                    i  1  I l 8 il I I l                        Ml1 l l                      l  l  l                e i l I i j i i fl l GROUP 5 @ 60"                                  l  l  I      [
E ~0 g l gl g g 7 TATE              . SHORT tem 1 STEADY S        INSE2 TION LufIT          I  I  i      . e I  I  I          e
              !              I I i i If I I I I I I I I I
                              ! GROUP 5 9'108"                    LONG TIM! STEADI              I  I  I R
      .,                        STATE INSERTION Ln!IT                          .            I      l 1          -
i                    FTT F 1 I I I I i l i I I I                                    i !          o
                \        ?
                                  ?
o
                                          ?8 o      o Q
o RR o    o RRo RR.
o            o      o
                  \
                    ;                            TEACTION OF EATED THE??X. Porggg
                      \ \
FIGURE 3.1-4
                          \q                      CEA INSERTION LIMITS VS THERMAL POWER (COLSS_00T OF SERVICE)
                                                                                                                  =s PALO VERDE - UNIT 1                                        3/4 1-32 CONTROLLED BY USER
 
CONTROLLED BY USER 3/4.2 POWER DISTRIBUTION LIMITS                                  ,
3/4 2.1 LINEAR HEAT RATE LIMITING CONDITION FOR OPERATION 3.2.1 The linear heat rate shall not exceed 14.0 kW/ft.
APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.
j ACTION:
With the linear heat rate exceeding its limits, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on kW/ft; or (2) when the COLSS is not being used, any OPERABLE Local Power Density channel exceeding the linear heat rate limit, within 15 minutes initiate corrective action to reduce the linear heat rate to within the limits and either:
: a. Restore the linear heat rate to within its limits within 1 hour, or
: b. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 6 hours.
SURVEILLANCE REQUIREMENTS i                                                                          .
4.2.1.1    The provisions of Specification 4.0.4 are not applicable.-
4.2.1.2 The linear heat rate shall be determined to be within its limits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS) or, with the COLSS out of service, by verifying at least once per 2 hours that the linear heat rate, as indicated on all OPERABLE Local Power Density channels, is less than or equal to 14.0 kW/ft.
4.2.1.3 At least once per 31 days, the COLSS Margin Alarm shall be verified to l    actuate at a THERMAL POWER level less than or equal to the core power operating limit based on 14.0 kW/ft.
l l
l I    PALO VERDE - UNIT 1                      3/4 2-1 l                      CONTROLLED BY USER I
  ~
 
l
                                                                                      ~
CONTROLLED BY USER'                                                                        u,
                - ..    - - - ~ ~                                                                  . .;
d q      POWER DISTRIBUTION LIMITS                                  ~-
                                                                                                              ,                j cd 3/4.2.2 PLANAR RADIAL PEAKING FACTORS            .,F  ;            ,'
LIMITING CONDITION FOR OPERATION 3.2.2 The measured PLANAR RADIA'L PEAKING FACTORS (F"yl shall be less than or                    4 equal to the PLANAR RADIAL PEAKING FACTORS (Fxy) used in the Core Operating Limit Supervisory SystemTCOLSS) and in the Core Protection Calculators (CPC).                              ;
APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER.*
ACTION:
With an F"y exceeding a corresponding Fxy, within 6 hours either:
: a. Adjust the CPC addressable constants to increase the multiplier applied to planar radial peaking by a factor equivalent to greater c
than or. equal to F"y/F xy and restrict subsequent operation so that a c
margin to the COLSS operating limits of at least [(F"y/Fxy) - 1.0]
x 100% is maintained; or                  '                                                  '
1                                                                                                                      %,
C
: b. Adjust the affected PLANAR RADIAL' PEAKING FACTORS (Fxy) used in the                    ,
                        .COLSS and CPC to a value greater than or equal to the measured    -
l                        PLANARRA5IALPEAKINGFACTORS(F"y)or.
: c. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER.
SURVEILLANCE REQUIREMENTS 4.2.2.1    The provisions of Specificatica 4.0.4 are not applicable.
4.2.2.2 +The measured PLANAR RADIAL PEAkIrtG FACTORS (F"y) obtained by using the incore detection syster, shall be determined to be less than or equal to C
the PLANAR RADIAL PEAKING FACTORS (Fxy), used in the COLSS and CPC at the following intervals:                          '
;                  'a. After each ' fuel loading with THERMAL POWER greater than 40% but prior to operation above 70% of RATED THERMAL POWER, and
: b. ,At least once per 31 Effective Full Power Days.
            *See Special Test Exception 3.10.2.
9 PALO VERDE - UNIT 1                        3/4 2-2 l
COFTROLLED BV USER
 
l
  ',.        4 CONTROLLED BY USER                                                :
l s,
g          STRIBUTION L 3/4;2.3 AZIMUTHAL POWER TILT - T Q LIMITING CONDITION FOR OPERATION ~
3.2.3 The AZIMUTHAL POWER TILTq(T ) shalI be less than or equal to the AZIMUTHAL POWER TILT Allowance used in the Core Protection Calculators (CPCs).
APPLICABILITY: MODE I above 20% of RATED THERMAL POWER.*
ACTION:
: a. With the measured AZIMUTHAL POWER TILT determined to exceed the AZIMUTHAL POWER TILT Allowance used in the CPCs but less than or equal to 0.10, within 2 hours either correct the power tilt or adjust the AZIMUTHAL POWER TILT Allowance used in the CPCs to greater than or equal to the measured value.
i
: b. With the measured AZIMUTHAL POWER TILT determined to exceed 0.10:
l
!                      1. Due to misalignment of either a part-length or full-length CEA, within 30 minutes verify that the Core Operating Limit Supervisory System (COLSS) (when COLSS is being used to monitor the core power distribution per Specifications 4.2.1 and 4.2.4) is detecting the CEA misalignment.
: 2. Verify that the AZIMUTHAL POWER TILT is within its limit within 2 hours after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours and verify that the Variable Overpower Trip Setpoint has been I
reduced as appropriate within the next 4 hours.
: 3. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL POWER may proceed provided that I                            the AZIMUTHAL POWER TILT is verified within its limit at least once per hour for 12 hours or until verified acceptable at
                      ,      95% or greater RATED THERMAL POWER.
4 "See Special Test Exception 3.10.2.
l PALO VERDE - UNIT 1                        3/4 2-3 l
I l
CONTROU.5D BY IJSER
 
CONTROLLED BY USER POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.
4.2.3.2 The AZIMUTHAL POWER TILT shall be determined to be within the limit above 20% of RATED THERMAL POWER by:
: a.  . Continuously monitoring the tilt with COLSS when the COLSS is OPERABLE.
: b. Calculating the tilt at least once per 12 hours when the COLSS is inoperable.
: c. Verifying at least once per 31 days, that the COLSS Azimuthal Tilt Alarm is actuated at an AZIMUTHAL POWER TILT less than or equal to the AZIMUTHAL POWER TILT Allowance used in the CPCs.
: d. Using the incore detectors at least once per 31 EFPD to independently confirm the validity of the COLSS calculated AZIMUTHAL POWER TILT.
l l
1.
PALO VERDE - UNIT 1                      3/4 2-4 CONTRCUD BY USER
 
                                                                                    ~
CONTROLLED BY USER
                                                                                                      .1
                            --- ~ .                                                                  'j g POWER DISTRIBUTION LIMITS',
3/4.2.4 DNBR MARGIN
    - LIMITING CONDITION FOR OPERATION                                    -                    -
3.2.4 The DNBR margin shall be maintained by operating within the Region _of_
6 GLtDhe_teqdffementsZof Actionl63f_TableA3-1              3 Acceptable Operation of Figure 3.2-1 or APPLICABILITY: MODE 1 above 20% of RATED THERW L POWER.
ACTION:                                                                .                  .
With operation outside of the region of acceptable operation, as indicated by either (1) the COLSS calculated core power exceeding the COLSS calculated core power operating limit based on DNBR; or (2) when the COLSS is not being used, any OPERABLE Low DNBR channel below the DNBR limit, within 15 minutes initiate corrective action to restore either the DNBR core power operating limit or the ONBR to within the limits and either:
: a. Restore the DNBR core power operating limit or DNBR to within its limits within 1 hour, or
: b. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.2.4.1 The provisions of Specification 4.0.4 are not applicable.                -
4.2.4.2 The ONBR shall be determined to be within its ifmits when THERMAL POWER is above 20% of RATED THERMAL POWER by continuously monitoring the core power distribution with the Core Operating Limit Supervisory System (COLSS).or, with the COLSS out of service, by verifying at least once per 2 hours that the DNBR iMrcir., as indicated on all OPERABLE DNBR margin channels, is within the limit shown on Figure 3.2-2.
4.2.4.3 At least once per 31 days, the Co SS Margin Alarm shall be verified to actuate at a THERMAL POWER level less than or equal to the core power operating ifmit based on DNBR.
4.2.4.4 The following DNBR or equivalent penalty factors shall be verified to be included in the COLSS and CPC DNBR calculations at least once per 31 EFPD.
GWD Burnuo (HTU)                      DNBR Penalty (%)*
0-10                              0.5 10-20                              1.0 20-30                              2.0*
30-40                              3.5 40-50                              5.5 "The penalty for each batch will be determined from the batch's maximum burnup assembly and applied to the batch's maximum radial power peak assembly. A single net penalty for COLSS and CPC will be determined from the penalties associated with each batch accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.
PALO VERDE - UNIT 1 -                  3/4 2-5 CONTROU.9D BY 'JSER
 
r                                                                                                                                                                        1 l
CONTROLLED BY USER                                                                                                          L
                                                                                                                      \    i Sc 9.pciicaW L SEQ f'                                                                                                              !
[7                              FIGUR E 3.21 DNBR MARGIN OPERATING LIMIT BASED ON COLSS (COLSS IN SERVICE)
N'        -
i i              i
                        ,/                                                            i                . . .
J.      .            t 95        /
                      /    100          --
                                                  +-----
          '                                        r            f                  t"              '. t                    .      ._
E                                                                        --
i
                        $                              REGION OF                              ~~~ .              ~.
                                                                                                                    ~
l Q                              ACCEPTABLE                                                                                            !
OPERATION ta<    80  --                -
22 U5 c=
i z>                                                                                    . . .
i i
Pc                                  *
                    <w                                                                                                                                      i 5E
: a. x 80  --                                              -          --        - - - - - - - -
                                                                                                                                          ~
O u.                                                                                    ..
EO                                                                                          ~- -
l W>                                                                                                                                    i b                                                                        REGION OF Im                                                                                                                                    i 40                                                                  , UNACCEPTABLE ,,,__,
gg
                                  -                    -                              - - - -                                              .,            i
                                                  ,,                                            OPERATION                                                i o                                                                                                                                      i gg                    -
t l                  az 1
oC                                                                                                                                  '
W2 0    20  -
l                      N es 0
O                20            40                  60                        80                    100 __ ..          ,
!                                                        PERCENT OF RATED THERMAL POWER                                                              i l                                                                        [
FIGURE 3.2-1 l                                            DNBR MARGIN OPERATING LIMIT BASED ON COLSS (COLSS IN SERVICE)
PALO VERDE - UNIT 1                                3/4 2-6 CONTROLLED BY USER
 
CONTROLLED BY USER                                                                                                                                                                                                                            :
s              -                                          -
f.-                                                                                                                                                                          $ g dph 1Chrkg$                                                              Mk
                                /                                                                        .                            FIGURE 3.2-2
                                                                                                                                                                                                                                                                            *x                h
(    DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATOR (COLSS OUT OF SERVICE)                                                                                                                                                      \                        ;
                                                                                                                                                                                                                                                                                                      \                      I
                                                                                                                                                                                                                                                                                                      \
                                                                                                                                                                                                                                                                                                        '                    I
                              /
                            /                                                                                                                                                                                                                                                                            .                  .
0.60
::id:: I                                      ::!i . .                      .:      !                I.                          !.-                      -
l                I                                  -
i--
                                                                                                                                                                                                                                                                                                                )
                                                                                                                                                                                                    ..l.-
:.u-            j. :- .                "v:                                        !..              l            :          -l                                            1.      r!-    . ;l: - :;~                                                              .
        ~
                                            ".: !!' "dMEGION OF                                                          ..        .      i.                          !:                    ..'      - l-                'I :i . -lI:- b
                                                                      ..:ACCEPTABLEJ~'~"y~-..:p !-
  - -      --                              .            t-                                                                                                                                              :j'
                              ~ 0.55    :T
                                          .....j-
                                                                        " OPERATION ..                              - -
4
                                                                                                                                                                                                                  -l
                                                                                                                                                                                                                            .'j'.    ..j:.:.,.]:i, I:. n.:. . . .p:: : :
(S                                                                              . . . . . .
y-                ::j.,- a. ;l . _.:. _. p.                                  .    :- -                                            ;      ,-
                                        . -:.,;:        y                      .;.          -
: i.            ;
i                                -
                                                                                                                                                                                            ;.              ,                i        -
                                                                                                                                                                                                                                                      ,;;,- .:      ,,l.~                                        '
                                              !i                                                                                              ( .05,0.51) - !                                                              I
: (.25, 0.51) ..
_ .i....i. ..,
j                                !                                                                                                                                                  ;.
                        ,                            =.._._.l,...;                                          .
r
                        !                                .                              !                    I                        f.                              i                                                                  .I 0.50                                                                                        --
lz5                            .i i                            .:
                                                                                                                                                                                    ~
i
                                                                                                                                                                                                      ._.; ; . . l i              ;
t' f c:                      .l      -
                                                                                      ;j.    .              i- . . .
i,                          i                  {              !
* l-l<
                      .E                                  i
                                                                                                                            '                ;                          8 i              j                i              l-                        '
      ~
Ig l
{-p . .j .":--i                          ;
l          ---j - "-.
                                                                                                                                                                                        -8 r-y;                  ;j (.30,,,0.46) 1                          .
2                  . . . .
fQ 0.45                      -
                                                                                                              '.                                        ..t            :
                                                                                                                                                                                                                            ..l:. :.: l
                                                                                                                                                                                                                                                              -t.
x                  *
                                                                                                                                                            -                      -
* I
: l . . '. . ' .-                                                      i D
g            . i . l. . i                                          i. h                        !                .
: 1 REGION OF                                        i. -            ..            .4              :I                                    t 6.
                                                                                                                        . .                                              i. . UNACCEPTABLE .                                                                            I:.- ;':-
                                                                                                                                                                                                                                                                                                                    \
j._.
i l        Z            .: t                                                                                  :                I                  .a i:.dOPERATION                                            !              '
                                                                                                                                                                                                                                                                                                                        }
:s:p[....!:..
z- -                                            p. .ll;:                        11          - ~:i;                i._. 1 ; l-
                                                                                                                                                                                                            !              ~[_    :T
: .";l:. ';.
0.40                                                                                                                                                                                    -
M.Wi..:!r.:di!i:                                                    !                        : .!: 1 .l.. .:. i. .              .
i l'
Ji.
                                                                                                                                                                                                                                                                      .!                                                          i
: g.                    .8                                  .
l
[( .30,0.38)
                                              . . . . 1. .:... . . . . ;                5                                                    -
                                                                                                                                                                                        . _. . .                . . _ ,                        :.                                                                                  j l              {                                  - i                                    t-i,_...._..    ,
                                                                                                                                        ..i..                                                                                                  . .
J 1,                0.35                    i                              .                    ,
                                                                                                                                        -=
i t,                      .._. .                          .. . .. _.. ..._.                                      .. .                              .. .                                  .
l
                ,\.                    ._.                          . . _ . . .                                .      .        . . _ _ . . , .                      ..      ..                                                        .                      _ _ . . . .                                          t
                                                                                                                                                                                                                                                                  *                                                            .s
!                                                                                        I                                    t                                            t                                I                                  f 0.30                                                                                                                                                                                                                                                                              j 0.3                                          0.2                                0.1                                        0.0                                0.1                              0.2                                        0.3                          -
                            \                                                                                                          CORE AVERAGE ASl*
                                                                                            'SEE SECTION 3.2.7 FOR THE ASI OPERATING LIMITS l
                                                                                                                                                                                                                                                                                "~
                                                                                                                                                                                                                                                                                                    }                          ~
FIGURE 3.2-2
[                                      ONBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS l                                                                                                                      (COLSS OUT OF SERVICE)
I PALO VERDE - UNIT 1                                                                                            3/4 2-7 CONTROLLED BY USER 1
l t
 
CONTROLLED BY dSER                ''                                                          l l
                    ,},POWEROISTRIBUTIONLIMItS.
i                              3/4.2.5 RCS FLOW RATE l
l                              LIMITING CONDITION FOR OPERATION
                                                                                                                        ~
l                              3.2.5 The actual Reactor Coolant System total flow rate shall be greater than or equal to 164.0 x 10s lbm/hr.
_ . ,                      APPLICABILITY: MODE 1.                                                              .
1
\
l l                              ACTION
    ' ' ~ ~
With the actual Reactor Coolant System total flow rate determined to be less 1
than the above limit, reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REOUIREMENTS 4.2.5 The actual Reactor Coolant System total flow rate shall be determined
                              ..to be greater than or equal to its limit at least once per 12 hours. .
l l
l l
PALO VERDE - UNIT 1                  3/4 2-8
:                                                          CONTROLLED BY USER k-
 
                                                                                          .I r
CONTROLLED BY USER m
i ect POWER DISTRIBUTION LIMITSN l
3/4.2.6 REACTOR COOLANT COLD LEG TEMPERATURE
                                                                                          'l LIMITING CONDITION FOR OPERATION g 3.2.6 The ' reactor' coolant '' cold 1eg temperature Oshall'~ be within the Area of Acceptable Operation shown in Figure 3.2-3. W.
APPLICA8ILITY: MODE 18 and 2*
ACTION:
E4 With the'geactor' coolant' cold" leg temperature exceeding its . fait, restore the temperature to within its limit within 2 hours or be in HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.2.6 The reactor coolant cold leg temperature shall be determined to be within its limit at least once per 12 hours.
      *See Special Test Exception 3.10.4.
PALO VERDE - UNIT 1                    3/4 2-9 COWRMI.ED B'f USH
 
r-CONTROLLED BY USER FIGURE 3.2 3 REACTOR COOLANT COLD LEG TEMPERATURE vs CORE POWER LEVEL 580        ,      ,    ,    ,      ,      ,        ,                        !
575  .
l 570 ?O
* I                        568 568 585                                                                            '
e 560
                                        .      AREA OF ACCEPTABLE OPERATION
                                                                                /      562 55s .
              $ 550                                                            552        ,
O 540 .
[,      10 20  30 40 50    60 70 80 90 100 l
l                            CORE POWER LEVEL.% OF RATEDTHERMAL POWER FIGURE 3.2-3 REACTOR COOLANT COLD LEG TEMPERATURE VS CORE POWER LEVEL PALO VERDC - UNIT 1                    3/4 2-10 CONTROU.ED BY USER L
 
CONTROLLED BY USER                                          d i
I
[POWERDISTRIBtITIONLIMITS                                                          I i
3/4.2.7 AXIAL SHAPE INDEX LIMITING CONDITION FOR OPERATION 3.2.7 The core average AXIAL SHAPE INDEX (ASI) shall be maintained within the following limits:
: a. COLSS OPERABLL
                      -0.28 i ASI i 0.28
: b. COLSS OUT OF SERVICE (CPC)
                      -0.20 1 ASI i + 0.20 APPLICABILITY: MODE 1 above 20% of RATED THERMAL POWER *.
ACTION:
With the core average AXIAL SHAPE INDEX outside its above limits, restore the core average ASI to within its limit within 2 hours or reduce THERMAL POWER to less than 20% of RATED THERMAL POWER within the next 4 hours.
SURVEILLANCE REQUIREMENTS 4.2.7 The core average AXIAL SHAPE INDEX shall be determined to be within its limit at least once per 12 hours using the COLSS or any OPERABLE Coro Protection Calculator channel.
See Special Test Exception 3.10.2.
PALO VERDE - UNIT 1                    3/4 2-11 CONTROLLED BY US5R
 
CONTROLLED BY USER gt. c POEEb!S          BUTION LIMIT'S.
3/4.2.8 PRESSURIZER PRESSURE LIMITING CONDITION FOR OPERATION
                                                                                    ~
3.2.8 The pressurizer pressure shall be maintained between 1815 psia and 2370 psia.
APPLICABILITY: MODES 1 and 2.*
ACTION:
                                                                                          ~
With the pitssurizer pressure outside its above limits, restore the pressure to within its limit within 2 hours or be in at least HOT STANDBY within the next 6 hours.
SURVEILLANCE REQUIREMENTS 4.2.8 The pressurizer pressure shall be determined to be within its limit at least once per 12 hours.
c--  __    _                                                            -- . 22 . - _  . . . . .
                                                                                                          ~ ~^
              *See Special Test Exception 3.10.5 PALO VERDE - UNIT 1                    3/4 2-12 CONTROU.ED BY USSR
 
CONTROLLED BY USER 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION r
LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.
APPLICABILITY: As shown in Table 3.3-1.
l ACTION:                                                                .
As shown in Table 3.3-1.
i SURVEILLANCE REOUIREMENTS 4.3.1.1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-1.
4.3.1.2 The logic for the bypasses shall be den:enstrated OPERA 8LE prior to each reactor startup unless performed during the preceding 92 days. The total bypass function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
4.3.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.
Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the
          " Total No. of Channels" column of Table 3.3-1.
4.3.1.4 The isolation characteristics of each CEA isolation amplifier shall be verified at least once per 18 months during the shutdown per, the following tests for the CEA position isol tion amplifiers:
: a. With 120 volts A. $ '(60 Hz) applied for at least.30-seconds across the output, the reading on the input does not- changeby more than 0.015 volt D.C. with an applied input voltage of 5-10 volts 0.C.
4        PALC VERDE - UNIT 1                  3/4 3-1 CONTROLLED BY USER                                                -
L
 
I                                                                                              \
CONTROLLED BY USER INSTRUMENTATION SURVEILLANCE REQUIREMENTS (Continued)
                                                                                          .i
: b. With 120 volts A.C. (60 Hz) applied for at least 30 seconds across            '
the input, the reading on the output does not exceed 15 volts D.C.
l 4.3;1.5 The Core Protection Calculators shall be determined OPERABLE at least once per 12 hours by ve'rifying that less than three auto restarts have occurred on each calculator during the past 12 hours. The auto restart periodic tests Restart (Code 30) and Normal System Load (Code 33) shall not be, included in this total.
4.3.1.6 The Core Protection Calculators shall be subjected to a CHANNEL                l FUNCTIONAL TEST to verify OPERABILITY within 12 hours of receipt of a High CPC        '
Cabinet Temperature alarm.
l l
l
{
i l
l 1
l t
PALO VERDE - UNIT 1                    3/4 3-2 CONTROLLED BY USER e
 
r 1
P V
TABLE 3.3-1 2            Y g                                          REACTOR PROTECTIVE INSTRUMENTATION h                                                                                  MININUM g                                                    TOTAL NO. CllANNELS      CHANNELS        APPLICABLE
        ,                FUNCTIONAL UNIT                OF CHANNELS      TO TRIP        OPERABLE          MODES        ACTION g  I. TRIP GENERATION A. Process
: 1. Pressurizer Pressure - High      4                2              3                              2,3 O                2. Pressurizer Pressure - Low        4                2 (b)          3 1, 2 1, 2 2,3
: 3. Steam Generator Level - Low                                                                          #  #
a/SG            2/SG            3/SG            1, 2            2,3
  -4 g                4. Steam Generator Level - High      4/SG            2/SG            3/SG            1, 2            2,, 3, w
: 5. Steam Generator Pressure - Low    4/SG            2/SG            3/SG            1, 2, 3* , 4
* 2,3 p                6. Containment Pressure - High      4                2              3              1, 2            2,3 F"  x            7. Reactor Coolant Flow - Low        4/SG            2/SG            3/SG            1, 2            2,3    P w            8. Local Power Density - High                                                                          #  #
o                9. DNBR - Low 4
4 2 (c)(d) 2 (c)(d)
                                                                                      -3 3
1, 2 1, 2 2,3 a
2,3  a o U                                                                                                                              O
            ,_B._ Excore. Neutron _F_ lug f                1.
Variable Overpower Trip          4'              2              3              1, 2            2,3
: 2. Logarithmic Power Level - High                                                                          C
: a. Startup and Operating                                                                            #  #
4                2 (a)(d)        3              1, 2            2,3
  %                                                      4                2              3              3*,      4*, 5* 8      y
: b. Shutdown                    4                0              2              3, 4, 5        4 C. Core Protection Calculator System
: 1. CEA Calculators                    2                1              2 (e)          1, 2            6, 7
: 2. Core Protection Calculators                                                                          #  #
4                2 (c)(d)        3              1, 2            2,3,7 e        n                            .--.e    .  - - . - ~  ~~-mn-
 
h v                                                                                                                    -
k 2                                                        TABLE 3.3-1 (Continued) o
    . <                                                  REACTOR PROTECTIVE INSTRUMENTATION Q
M                                                                        -
MINIMIM i                                                            TOTAL NO. CilANNELS          CHANNELS APPLICA8LE g                  FUNCTIONAL UNIT                          OF CHANNELS    TO TRIP          OPERABLE    MODES      ACTION y        D. Suppl ementa ry _ Pro tec t.f on_Sys tes Pressurizer Pressure - High              4 (f)          2                  3        1, 2          8 II. RPS LOGIC -
Z q              A. Matrix Logic                                6              1                  3 Z
1, 2          1 q
  %                                                                6              1                3        3* , 4 * , 5* 8      %
8.' Initiation Logic                              4              2                4        1, 2          5      O g                                                          4              2                4        3* , 4* , 5*  8 III. RPS ACTUATION DEVICES                                                                                                M g              A. Reactor Trip Breaker                        4 (f)          2                4        1, 2          5 q                                                                4 (f)          2                4        3*, 4* , 5*  8 q              8. Manual Trip                                  4 (f)          2                4        1, 2          5 m                                                                4 (f)          2                4        3*, 48, 5*    8      g M                                                                                                                                ITI N                                                                                                                                N G
 
r CONTROLLED BY USER TABLE 3.3-1 (Continued)
TABLE NOTATIONS
          *With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.                    '
          #The provisions of Specification 3.0.4 are not applicable.
(a) Trip may be manually bypassed above 10-4% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is less than or equal to 10-4% of RATED THERMAL POWER.
(b) Trip may be manually bypassed below 400 psia; bypass shall be                            i automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.                                                                    ;
(c) Trip may be manually bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 1% of RATED THERMAL POWER.
(d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.
(e) See Special Test Exception 3.10.2.
i (f) There are four channels, each of which is comprised of one of the four reactor trip breakers, arranged in a selective two-out-of-four configuration (i.e., one-out-of-two taken twice).
ACTION STATEMENTS    ,
i ACTION 1      -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and/or open the protective system trip breakers.                                                    "
ACTION 2      -
With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or l
tripped condition within 1 hour. If the inoperable channel is bypassed, the desirability of maintaining this channel in the bypassed condition _shall be reviewed in accordance with              -
                  @          SpecificationC6.5.1. 63ih The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN.
                                                        , ( & q 'ca u s-Q t k ild 5[i U Sb3%h' l
l i
l PALO VERDE - UNIT 1                        3/4 3-5 L
 
CONTROLLED p..-          -
BY USER TABLE 3.3-1 (Continued)
ACTION STATEMENTS With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below:
Process Measurement Circuit          Functional Unit Bypassed / Tripped I      1.
Gtw    bahn k k w!"
Linear Power n
      #4                                                        VariableOverpoweri(RPS))i(,RPg (Subchannel or Linear)          Local Power ensity - High
                                                                                  )
DNBR - Low (
: 2. Pressurizer Pressure - High    PressurizerPressure-High.h)
(Narrow Range)                  Local Power Density - High (RPS),,
                                                                                              ~
DNBR - Low (g j
                    )    3. Steam Generator Pressure -
Low Steam Generator Pressure - Low Steam Generator Level 1-Low (ESF)
Steam Generator Level 2-Low (ESF)
                  /                          .
: 4. Steam Generator Level - Low    Steam Generator Level - Low (R/S)
{                  (Wide Range)
* Steam Generator Level 1-Low (ESF)
              ,                                              Steam Generator Level 2-Low (ESF)
S. Core Protection Calculator      Local Power Density - High((RPS)>
                                                                                            ~
DNBR-LowC(g ACTION 3    -
With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:
: a. Verify that one of the inoperable channels has been
'                              bypassed and place the other channel in the tripped condition within 1 hour, and I                        b. All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below:
(          Process Measurement Circuit          Functional Unit Bypassed / Tripped
                              ^Erert h det.<fe,sh w t                                n
: 1. Linear Power                    Variable Overpower (RPS))
k                                                Local Power Densit e
(Subchannel or Linear) p A'                                    ,
DNBR - LowfRPS)) y -ligh      '- ' (RPS);
                                                                                              -m
: 2. Pressurizer Pressure -          Pressurizer Pressure - High (RPS).
High (Harrow Range)            Local Power Density - High (RPS)      .
DNBR - Low (RPS);
                                              .                              ~
PALO VERDE - UNIT 1 3/4 3-6 CONTROLLED BY USER
 
                                                                                          ~
CONTROLLED BY USER TABLE 3.3-1 (Continued)
ACTION STATEMENTS
        ,            3. Steam Generator Pressure -                  Steam Generator Pressure - Low
(                  Low Steam Generator Level 1-Low (ESF) x                                                              Steam Generator Level 2-Low (ESF)
          )          4. Steam Generator Level - Low                  SteamGeneratorLevel-Low (RYS))
(Wide Range)                                Steam Generator Level 1-Low (ESF)
Steam Generator Level 2-Low (ESF)
: 5. Core Protection Calculator                    LocalPowerDensity-High,(US);
DNBR - Low (RPS)            v STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent                          -
STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied.
ACTION 4  -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes.
ACTION 5  -
With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continu's provided the reactor trip breaker of the inoperable channel is placed in the tripped condition within 1 hour, otherwise, be in at least HOT STAND 8Y within 6 hours; however, the trip breaker associated with the inoperable channel may be closed for up to 1 hour for surveillance testing per Specification 4.3.1.1.
ACTION 6  -
: a. With one CEAC inoperable, operation may continue for up to
(                          7 days provided that'at least once per 4 hours, each CEA i                          is verified to be within 6.6 inches (indicated position) of .alLother_CEAs_in j ts_ groupdAf ter 7 'daysToperation V{
        ,f (may continue provided that the conditions of Action Item 61b or 6.c are met.                ......
:                  b. With both CEACs inoperable and COLSS in D c_e, operation may continue provided that:
                            . WithinT:hourn-                                    -
m a)  Operation is restricted to the limits shown in Figure 3.3-1. The DN8R margin required by Specification 3.2.4 is replaced by this
!                                      restriction when both CEAC's are inoperable
  .                                    and COLSS is in operation.
,                                b)    The Linear Heat Rate Margin required by Specification 3.2.1 is maintained.
{
l                      \        c)    The Reactor Power Cutback System is placed out l                                      of service.                                                              .
i
                                                        .,_.- --              y                          ,,
                  $,          f Lt f&    YeUs    Ao V*              %$ U ade                            :i+~un
: 3. i . c w. .:u exwa'eh h10~.3              k,anA Nyolv O1'ma w
W"
* U. w ,Y ' O .
l*d TE!:PHLa
* O E'
PALO VERDE - UNIT 1                        3/4 3-7 CONTROLLED BY USER
 
r CONTROLLED BY_ USER oh*                                    TA8LE 3.3-1 (Continued) '
N ACTION STATEMENTS              )
: 2. Within 4 hours:
a)    All full-length and part-length CEA groups are withdrawn to and subsequently maintained at the                        i
                                              " Full Out" position, except during surveillance                      ,
testing pursuant to the requirements of Specifica-
* tion 4.1.3.1.2 or for control when CEA group 5                        '
may be inserted no further than 127.5 inches withdrawn.
b)    The "RSPT/CEAC Inoperable."_ addressable constant
                              ,g_            in the CPCs is set totDe indicated that bothw '% m, epv.2M .
(CEAcrs are inopefabTe.
QA.
c)    The Control Element Drive Mechanism Control System (CEDMCS) is placed in and subsequently maintained in the " Standby" mode except during CEA group 5 motion permitted by a) above, when the CEDMCS may be operated in either the " Manual Group" or " Manual Individual" mode.
: 3. At least once per 4 hours, all full-length and part-length CEAs are verified fully withdrawn except during surveillance testing pursuant to Specification 4.1.3.1.2 or during insertion of CEA group 5 as permitted by 2.a) above, then verify at least once per 4 hours that the inserted CEAs are aligned within 6.6 inches (indicated position) of all other CEAs in its group.
                            /    4. Following a CEA misalignment with both CEA gg J /                inoperable and COLSS in operation, operation may continue provided that within 1 hour:
{                    .
l The power is reduced to 85% of the pre-misaligned          ;
    .                                power but need not be reduced to less than 50% of            -
                        ,            RATED THERMAL POWER. This power restriction replaces l              the power restriction of Specification 3.1.3.1, l              otherwise Specification 3.1.3.1 remains applicable.
                    .!    c. With both CEACs inoperable and COLSS out-of-service,                  I l          operation may continue provided that:                                  I f            1. Within 1 hour:                                                    I l                                    a)    The existing CPC value of the CPC addressable
                  ;                        constant "BERR1" is multipled by 1.19 and the resulting value is re-enter'ed into the CPCs.
l
                    \                b)    The Reactor Power Cutback System is placed out
                        \                    of service                                                          '
l
                          \  N c)    The COLS$ out of service Limit Line, Specifica-tion 3.2.4, is not applicable to this mode of N ~ operation,                                                    j
                                                          ~
                                                                                . . .      __,/
l PALO VERDE - UNIT 1                      3/4 3-8 CONTROLLED BY USER l
t
 
CONTROLLED
                                        ~'
BY USER TABLE 3.3-1 (Con'.inued)' '
ACTION STATEMENTS
                                /
l'
: 2. Within 4 hours:                                                                    I vY?
f'                                                                                                .;
                        /
a)    All full length and part length CEA groups are withdrawn to and subsequently maintained at the
                                                  " Full Out" position, except during surveillance I                          testing pursuant to the requirements of Specifi-
                        }                          cation 4.1.3.1.2 or for control when CEA group 5    t i                        may be inserted no further than 127.5 inches          \
withdrawn.
i b)    The "RSPT/CEAC Inoperable" addressable constant I
in the CPCs is set to be indicated that both          '
CEAC's are inoperable.
                            !              c)    The Control Element Drive Mechanism control              k l                      System (CEDMCS) is placed in and subsequently              !
maintained in the " Standby" mode except during CEA group 5 motion permitted by a) above, when            f f                        the CEDMCS may be operated in either the " Manual          ,
                        !                        Group" or " Manual Individual" mode,                        t
                        !                                                                                      \
l'
: 3. At least once per 4 hours, all full length and part                  i length CEAs are verified fully withdrawn except                      !
during surveillance testing pursuant to Specifica-                      i tion 4.1.3.1.2 or during insertion of CEA group 5 as                      s permitted by 2.a) above, then verify at least once per 4 hours that the inserted CEAs are aligned within                        ,    .
6.6 inches (indicated position) of all other CEAs in                          i its group.          -
j f
: 4. Following a CEA misalignment with both CEAC's and COLS5                          !
inoperable, operation may continue provided that within                        !
I hour:
                                                                                                                        /
The power is reduced to 85% of the pre-misaligned power but need not be reduced to less than 50% of RATED THERMAL
(
POWER. This power restriction replaces the power                '
restriction of Specification 3.1.3.1, otherwise Speciff ,
cation 3.1.3.1 remains applicable.                        -
_                          - - -- -- .J ACTION 7  -
With three or more auto restarts, excluding periodic auto restarts (Code 30 and Code 33), of one non-bypassed calculator during a 12-hour interval, demonstrate calculator OPERABILITY by performing a CHANNEL FUNCTIONAL TEST within the next 24 hours.
ACTION 8  - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore an inoperable channel
      .                        to OPERABLE status within 48 hours or open an affected reactor trip breaker within the next hour.
i        PALO VERDE - UNIT 1                            3/4 3-9 l
I CONTROLLED BY USER                                                                              ,
1
                                                                                                                                  )
 
                                                                                                                                                                                              .                                                              .                  I CONTROLLED BY USER                                                                                                                                                                                                    !
i 1
                                                                                        -                                          ~~~.~
* p.*                                                                                          g
                                                                    -                                                                                                        ~,                                                                                                    ;
FIGURE 3.31                                                                                    'x.                                                                        l
                ?;3                        ,e ONBR MARGIN OPERATING LIMIT 8ASED ON COLSS FOR                                                                                                                                                                                    :'
            , l 0 .'
                                    ,/                                                  8OTH CEAC'S INOPERA8LE                                                                                                                      \                                    .
                          /
                            /                                                                                                                                                                                                                      ,                      s
                      /                                                                                                                                                                                                                                                  t l
14') /                                                                        1                                i              .                  n                  .                      s l dj .                                      d. .                      . . . . . . .                'h              .    , ;. ... l , . . . '                                        . ., ' . .
                                                                                                          .{j
                                      .l.
l              l          :
                                                                                                                                                                                    ;e-                                  :-              ::r.            1
                        . i: .!.                        .j. ; L, : .                              .
I                  F I          fl- 0 :                                              1                            ,
j                      = ~ r :-- , .            p                        ..                  j                  .                                            .. ,                                                              .
                                                            ,      7. ",                                .
1
                                        ;;....[
                                              ". .                    a.      t            :
                                                                                                                    .        i        .                        .                      :                    (100,118.7)_
u-              "            !' r . ;                  '      ... ;. . .. l...... i. .                                                  1        i~-
e l
                      ". ".. . : i. : : :: .. }l
                                                                      ,      j ."          :
                                                                                            .            :                  l        .-        ..u. ... (96,112.7) ,.;.;:
                                                                                                                                                                                                                  .  . l!      '.j. :.          .
l .,/                        .          g-      .. .
                      . . . ! :. . r "t . :. ".....
r-    . . . . .          .. .
: h. ,.,. l
                                                                                                                                                                                                      .u
                                                                          .g..                                                          .
2j                        ,,,t,.. ..                        I                g
* e
                                                                                                                                                    ..g..g.                  ~
                                                                                                                                                                                                        , , , g '.P . .. . 7 O g                                    1 100                      ,
Qw                                    .;                    j                c            .                                                l  ..                            ,;.,                            .
                                      ...;. . ... t R E G IO Nd O l ' - "* ,- "! " . '. . .. [t" * '-                                                                                                :;r "' . - - "
w                    _
                                                                                                                                                  '"                            -[......  ;"          ""
                                                          .: ACCEPTABLE                                                                                                                                                -
8                        .        l          ..-      j OPERATION                            5      i                .
I
                                                                                                                                                                            >: j. p ,j,                  .
                        ...:.;              .:r.a                                                      . . . . . . . . . . .                      ,"
* g                                                                . . . . . .          . . .                                                                                ,                    ,,
d w            80            .        .
                                                                                                                        - ~
i                  , (73,4,73,4) . .-                        ..              . . .
                                                                                                                                                                                                                                  -+-        .
                                                                      ...i..
_r                    .. . ..-._ ; _ _                                        ;                            '
                                                                                                                                          . . . . . _ ! . _. . ' . . . ... r.- .-.
so                                            .
: l.                              .
(w gr W                    . ..;.. 7
: d ; .                ..J...      .                                                              .          .            . . . . . . .
                                                                                                                                                                                                    ^
8                                    ,-
l                                                              .                  ;                  e                                                                                ;
gg              60                      .
lf . .            ,
                                                                                                                                                          ...... .. t .
wg                            ..          i,                  ,                ,            .              ,
                                                                                                                                                .                  e
                                                                                                                                                                                                            ,i.                  .-
                                          , ,,                c .. . . _.                    .......,                                    ....g.
ua                                    i                  !                e            ;                                                                                                                                    :
y                                                                                                  .-                                                                                                          :
mz                                        '                                                                                                    i                                                                                  -
L-                  .              .l.-            ;
                                                                                                                                            . .g.                          ,
g                40                                                              .
UNACCEPTABLE- ' -                                                          - ---
.1                        ..                . . . . . . . . . . .
O P ER ATIO N . _,. .
8                          _ _ . _ _ _ . . . _ .                                          ._ _
i.
                                                                                                                                      - - . --                                                                    - ~                        - - -
20                                                  - - -              --                    -
            ,i f                                I                                f                                  f                                          f 0                                        20                              40                              60                                80                                        100 s                                                                PERCENT OF RATED THERMAL POWER
                                                                                                                                                                                                                                                        /
pal.0 VERDE - UNIT.1._ .. '                                                                                                                                                              ' .
                                                                                              -..._ 3/4 3-10                                      , _ . .
CONTROLLED BY USER
 
b
                                              ~
f,        1 5
j                            h      TABLE 3.3-2 f.
l          5                                              REACTOR PROTECTIVE INSTRtMENTATION RESPONSE TIMES h            FUNCTIONAL UNIT                                                              RESPONSE TIME I. TRIP GENERATICM                                                                                        .
c z
_A.__ Process  -                                                              ,
q                        1. Pressurizer Pressure - High i
51].1 seconds
                            "                        2. Pressurizer Pressure - Low
                                                                                                                        $ 1.15 seconds          b
: 3. Steam Generator Level - Low                                    $ 1.15 seconds Z                                      4. Steam Generator Level - High                                    1 1.15 seconds          2
: 5. Steam Generator Pressure - tow                                  1 1.15 seconds
: 6. Containment Pressure - High                                    1 1.15 seconds O                                      7. Reactor Coolant Flow - Low i
1 0.65 second O
                                                                                                                                                %F F
y==                                    8. Local Power Density - High+ M                      -
V                F'".
5:a
_                y--
E            [                          a. Neutron Flux Power from Excore Neutron Detectors            < 0.75 second*          M U            a
: b. CEA Positions
: c. CEA Positions: CEAC Penalty Factor 3%.35Isecond**          Q
                                                                                                                        <0.75jsecond**
: 9. DNBR - Low Dk
              $                                        a. Neutron Flux Power free Excore Nautron Detectors            <:
o%
0.75'gsecond*
f C                                        b. CEA Positions
: c. Cold Leg Temperature 31.35:second**
4                  g g                                                                                                        $ 0.75 secondff E
: d. Hot Leg Temperature                                          < 0.75 secondif        M
: e. Primary Coolant Pump Shaft Speed                            il0.75                M N                                          f. Reactor Coolant Pressure from Pressurizer                30.75j,secondi secondiff  M
: g. CEA Positions: CEAC Penalty Factor                          5{0.75jsecond**
                                              .B.__Excore Neutron Flux                              ,                          s
: 1. Variable overpower Trip                                        1 O.55 second*
: 2. Logarithmic Power Level - High
: a. Startup and Operating f
130.551second*
: b. Shutdown                                          -
10,.55jsecond" 4
                                          " i agiha.it.JM2 fer :gr<I h (4h*'a h4
 
L')
          .                                                                                                        V              y 8
* TABLE 3.3-2 (Continued) 3I
    ;    5                                                      REACTOR PROTECTIVE INSTRtHENTATION RESPONSE TIMES '{
M m
M        FUNCTIONAL LMIT                                                            .                RESPONSE TIME
                          .C._ Core _ Pr_ o t_ec_t i ca. .Ca l cu l a to r___Sys t em z                      __        _        .      _        _ _ .
Q                      1. CEA Calculators                                                        Not Applicable
          "                      2. Core Protection Calculators                                                Not Applicable D. Supplementary Protection System Pressurizer Pressure                    Hiah                              1 15 econd.                                            2 S
T II. RPS LOGIC                                                    /
IU4#/                                                                N Q
F A. Matrix, Logic
: 8. Initiation Logic Applicable
                                                                                                                  . Applicable p
U III. RPS ACTUATION DEVICES                                            h g                A. Reactor Trip Breakers                                                          ; Applicable                                          Q g                          8. Marmal Trip                                                                    : Appilcable N-
                      =
Neutron detectors are exempt i                                                          ise time of the neutron flux signal                    d C              Hg} Parti = at the ch===i shan i w Josponent in channel.
e fro. the input or first eiectr "'"
                                                                                                                    -- -                                                C
                                                                    ~ ^
                                ~ . . -              - -                                                                                                        ^~
FT1                      Response time shall be measure                                                          eptable CEA sensor response shall be i                pyy y        @ [((An        demonstrated by compilance wit
                      #The pulse transmitters measur _                                                          e time testing. The response time
      .              . shall be measured from the pulse shaper input.                                        _        _ _ _ ,
                < / ## Response time shall be measured from the output of the resistance temperature detector (sensor). RTD response time shall be measured at least once per 18 months. The measured response time of the                                              i
          @h li slowest RID shall be less than or equal to 13 seconds. Adjustments to the CPC addressable constants given in Table 3.3-2a shall be made to accomodate current values of the RTD time constants. If the RID time constant for a CPC channel exceeds the value corresponding to the penalties currently in i      use, the affected channel (s) shall be declared inoperable until penalties appropriate to the new
                  \ , time constant are installed.                                                                  -
                  ### Response time shall be measured from the output of the pressure transmitter. The transmitter response                                                .-
time shall be less than or equal to(0.7)second.
I L a y la.t. St.L G r r u(ML t-pf0*
 
TASLE 3.3-2      (Cont'd)
REACTOR PROTECTIVE IItSTRI4tENTATION Resp 0ItSE TIES lieutnmi detectors are exempt from repsonse time testing. The response time of the neutron flux signal
        ,          portion of the channel shall be measured from the detector output or fram the input of the first electronic
      ,ag)f        couronent in the channel.
                            -~~-        _                                  -
                                                                                            ~ . _ _    - -
[**      T'.  '                                                                                      . - __ . . . ,
osition transmitters are~emempt from response time testing. The response time shall be measured
                    ' as the in y t to the CpC. CEAC or signal isolator.
k** Response times are verified using CpC Response Time Test Software, and are for. hardware delays only.
The pulse transmitters measuring pump speed are exempt from response time testing. The response time shall g, , , _ be measured from the pulse shaper input.
      >'      r          ~ _
y i ##    Response time shall be measured from the output of the resistance temperature detector (sensor). RTD
=          ;      response time shall be measured'at least once per 18 months. The measured response time (P ) of the slavest y-RTD shall be less than or equal to 6.0 seconds.                                                T
~
              -### Response time shall be measured from the output of the pressure transmitter. The transmitter response time shall be less than or equal to (0.7) seconds.
w.
E%
Q*
y:z
==
  .o i
 
CONTROLLND BY USER TABLF 3.3-2a
    -=,
D                            INCREASES IN BERRO, BERR2, AND BERR4 VERSUS RTO OELAY TIMES
                                                                              ~N~.\ h BERR0            BERR2                  BERR4 RTO DELAY TIME                      INCREASE          INCREASE                              I INCREASE s
(%)                (%)                  (%)      \
(t)                                                                                  !
t < 8.0 sec                        0                0                      0 870 sec < t < 10.0 sec            2.5              2.0                    1.0            -
10.0sec<t{13.0sec                  6.0              4.0                    6.0            l NOTE: 8 ERR term increases are not cumulative.      For example, if the time constant changes from the range of 8.0 < t i 10.0 see to the range 10.0 < t < 13.0, the BERR0 increase from its original (t < 8.0 see)-
value is T.0 not 2.5 + 6.0.
y\
K_                                _ _  .
s l
l          PALO VERDE - UNIT 1                    3/4. 3-13 CONTROLLED BY USER
 
b
                                ,                                                                                              TABLE 4.3-1
                                ?
O REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS 0
m CHANNEL      MODES IN Wi!CH CHANNEL      CHANNEL                                FUNCTIONAL    SURVEILLANCE E                          FUNCTIONAL UNIT                                                  CHECK      CALIBRATION                                  TEST          REQUIRED E                                TRIP GENERATION O                              -
I.
A. Process O
: 1. Pressurizer Pressure - High                          S            R                                  H              1, 2 2.
3.
Pressurizer Pressure - Low Steam Generator Level - Low S
S R                                  M              1, 2          Z R                                  H              1, 2
: 4. Steam Generator Level - High                          5            R                                  H              1, 2
: 5. Steam Generator Pressure - Low                        5            R                                  H              1, 2, 3*, 4*
F" f"                            1                                    6. Containment Pressure - High                          S            R                                  H              1, 2 r
p
                                ';'                                  7. Reactor Coolant Flow - Low                            5            R                                  M              1, 2          M Q                              5                                    8. Local Power Density - High                            S
                                                                                                                                                        -~
I ~D (2, 4)7 R'(4, 5)                      M. R (6)      1, 2          O W                                                    '
: 9. DNBR - Low                                            S      D (2, 4), R (4, 5)                                      1, 2          W f                                                                                                                                      M (8), S (7)                            M, R (6) ./)
C                                                              B. Excore Neutron Flux                                                WMFit7dm Cd**ddM5O**
g                                                                    1. Variable Overpower Trip                                S      D (2, 4), M (3, 4)                      M              1, 2          C m                                                                                                                                      Q (4)                                                                  M y                                                                    2. Logarithmic Power Level - High                        5      R (4)                                    M and S/U (1)    2  3, 4, S m
C. Core Protection Calculator System
: 1. CEA Calculators                                      S            R                                  M. R (6)      1, 2
: 2. Core Protection Calculators                          S      D (2, 4), R (4. 5)                      M (9), R (6)  1, 2 M (8), S (7) e
_ _ = _ . , . -            m..
 
l l
1    ?
g                                                                      TABLE 4.3-1 (Continued) o I      <                                      REACTOR PROTECTIVE INSTRUNENTATION SURVEILLANCE REQUIREMENTS E
o CHANNEL      MODES IN nAIICH CHANNEL                CHANNEL                          FUNCTIONAL    SURVEILLANCE FUNCTIONAL UNIT                                                        CHECK              CALIBRATION                            TEST            REQUIRED
      -4 c1  **            D. Supplementary Protection System                                                                                                                                O Pressurizer Pressure - High                                5                      R                        H              1, 2 Z
q            II. RPS LOGIC z
N                  A. Matrix Logic                                                    N.A.                    N.A.    -                M              1, 2, 3* , 4 * , 5*          %
B. Initiation Logic                                                H.A.                    N.A.                      M              1, 2, 3* , 4 * , 5*
l                                                                                                                                                                                            g--
l r-"" w l      A      III. RPS ACTUATION DEVICES
* f""
gyg ad  u ggg g)
D    h (t9          A. Reactor Trip Breakers                                          N.A.                    N.A.                      M, R (10)g      1, 2, 3*, 4*,      5*        Q
  @                  B. Manual Trip                                            -
N.A.                    N.A.                      M              1, 2,    3*, 4*,  5*        @
                                                                                                                                ~
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c m                                                                                                                    .
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  ,*      w    '4  ,.I                                }
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          ~
                                                                                    ?          :                                    <
                  's
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A
                                                                                                                                                      -            -~~m=            .-e-n.www      -
      "                        -    i,    %                                    k
 
CONTROLLED BY USER TABLE 4.3-1 (Continued)
TABLE NOTATIONS With reactor trip breakers in the closed position and the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.
(1)    -
Each STARTUP or when required with the reactor trip breakers closed and the CEA drive system capable of rod withdrawal, if not performed in the previous 7 days.
d (2)      -
Heat balance only (CHANNEL FUNCTIONAL TEST not included), above 15%
of RATED THERMAL POWER; adjust the 'tinear Wwer % vel, the CPC delta T power and CPC nuclear power signals to agree with the calorimetric calculation if absolute difference is greater than 2%. During PHYSICS              .
TESTS, these daily calibrations may be suspended provided these                        ,
calibrations are performed upon reaching each major test power plateau                  ,
and prior to proceeding to the next major test power plateau.
(3)    -
Above 15% of RATED THERMAL POWER, verify that the linear power sub-                    I channel gains of the excore detectors are consistent with the values used to establish the shape annealing matrix elements in the Core Protection Calculators.
(4)    -
Neutron detectors may be excluded from CHANNEL CALIBRATION.
(5)    -
After each fuel loading and prior to exceeding 70% of RATED THERMAL POWER, the incore detectors shall be used to determine the shape a
annealing matrix elements and the Core Pr'otection Calculators shall A,'                          use these elements.
Y a                (6)    -
This CHANNEL FUNCTIONAL TEST shall include the injection of simulated process signals into.the channel as close to the sensors as practicable i'
to verify OPERABILITY including alarm and/or trip functions.
S        (7)    -
Above 70% of RATED' THERMAL POWER, verify that the total steady-state                1 RCS flow rate as indicated by each CPC is less than or equal to the actual RCS total flow rate determined by either using the reactor coolant pump differential pressure instrumentation ,or by calorimetric calculations nd if necessary, adjust the CPC addressable constant
                                                                                      ~
flow coeffic nts such that each CPC indicated flow 1s less than or                    -
equal to th actual flow rate. The flow measurement isncertainty may be includedirLthe BERR1_ term in tile CPland_is_equallto_or_ greater, than 4%. (,Ciu4tGMut4 chs9aud & NC NPh
: d. (8)      -
Above 70% of RATED THERNAL POWER, verify that the tEab steady-state
_                            RCS flow rate as indicatedsby each CPC is less than or equal; to the
      ,                      actual RCS total flow rate determined by either using the reactor s%                            coolant pump differentral pre'ssure instrumentation and the ultrasonic
/^ -
^
                        . flow meter adjusted pump curves 1or calorimetric calculations!
The monthly CHANNEL FUNCTIONAL TEST shall include verification that U,0            (9)    -
the correct values of addressable constants are installed in each s            OPERABLE CPC per Specification 2.2.2.
4 (10) -    At least once per 18 months and following maintenance or adjustment of the reactor trip breakers, the CHANNEL FUNCTIONAL TEST shall
    ,          j            include independent verification of the undervoltage and shunt trips.
PALO VERDE - UNIT 1                        3/4 3-16                    .
e C                  ..
CONTROMD BY USER
                                - - -            - _ -        - -- .                      __..___-_----.___s
 
l
                                      ~
CONTROLLED BY USER                                                    :
g      ,
s                l Bde      -
INSTRUMENTATION      s
                          ~ _    _      m.,
3/4.3.2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION l
LIMITING CONDITION FOR OPERATION 3.3.2' The Engineered Safety. Features Actuation System (ESFAS) instrumentation            ;
channeis and' bypasses shown in Table 3.3-3 shall be OPERABLE with their trip              i setpoints set consistent with the values shown in the Trip Setpoint column of              i Table 3.3-0 add with RESPONSE TIMES as'shown in Table 3.3-5.                              I APPLICA8ILITY: As shown in Table 3.3-3.
s.
ACTION:
: a. With an ESFAS instrumentation channel trip setpoint less conservative than the value shown in, the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.                                                  -
: b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.    .
                      -SURVEILLANCE REOUIREMENTS
              ,        4.3.2.')    Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST cperations for the MODES and at the frequencies shown in Table 4.3-2.
4.3.2.2 The logic for the bypasses shall be demonstrated OPERABLE during the at' power CHANNEL FUNCTIONAL TEST of channels affected by bypass operation.
The' total bypass function shall be demonstrated OPEP.ABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by bypass operation.
4.3.2.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No.      ~
1 of Channels" Column of Table'3.3-3.
av          5
                                        \j 9
1 1
PALO VERDE - UNIT 1                      3/4 3-17 CONTROLLED BY. USER
 
t l
s                                                                            TABLE 3.3-3
                                !                                                        5
                                                                                          <                                        ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION b
I                                                                                                MINIMUM c-                                                                    TOTAL NO.      CHANNELS    CHANNELS                  APPLICABLE 5--a ESFA SYSTEM FUNCTIONAL UNIT                            OF CHANNELS    TO TRIP    OPERABLE                      MODES        ACTION O
* I. SAFETY INJECTION (SIAS)                                                                                                            O A. Sensor / Trip Units                                .                                                                            O
: 1. , Containment Pressure - High , 4                          2          3                        1,2,3,4          13*, 14*
N                                                                                                2. Pressurizer Pressure - Low                4            '2          3                        1, 2, 3(a), 4    13*, 14*  %
B. ESFA System Logic-r                                                                          e*                                                                                                                                                      r-
: 1. Matrix Logic                              6              1          3                        1,2,3,4          17        F T                                                                                                                                                      571 O                                                                          g                      2. Initiation Logic                          4(c)          2(d)        4                        1,2,3,4          12 Q
W                                                                                                  3. Manual SIAS (Trip Buttons)                4(c)          2(d)        4                        1,2,3,4          12        g C. Automatic Actuation Logic                      2              1          2                        1,2,3,4          16 II. CONTAINMENT ISOLATION (CIAS)
M                                                                                              A. Sensor / Trip Units                                                                                                            FT1 N                                                                                                                                                                                                                                '
s y
: 1. Containment Pressure - High              4              2          3                        1,2,3            13*, 14*,
: 2. Pressurizer Pressure - Low                4              2          3                        1,2,3(a)          13*, 14*
B. ESFA System Logic
: 1. Matrix Logic                              6              1          3                        1,2,3            17
: 2. Initiation. Logic                        4(c)          2(d)        4                        1,2,3,4          12
                                                                                                                                                                                                                                                      .-  i d
* e          e      .*                    .,
9 e                                                    % w          -e. wee-#w--e                          -
 
i g                                                                                                TABLE 3.3-3 (Continued)                -
5
  ~
                      <                                                                              ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION N
                      =                                                                                                        *
                                                                                                                                                    ~
                            ,                                                                                                                        HINIMUM c                                                                                                TOTAL NO.      CHANNELS      CHANNELS    APPLICABLE
                      -4 ESFA SYSTEM FUNCTIONAL UNIT                  OF CHANNELS    TO TRIP        OPERABLE        MODES    ACTION O                    H II. CONTAINNENT ISOLATION (Continued)                                                                          O
: 3. Manual CIAS (Trip Buttons)      4(c)          2(d)                        1,2,3,4 Z                                                                                                                                  - -
4 12 Z
H                                                                                    . Manual SIAS (Trip Buttons)      4(c)          2(d)          4            - -1,2,3,4            H y                                                                                                                                          - - ---              .
15)
C. Automatic Actuation Logic          2              1            2            1,2,3,4      16 F                  R                                                    III. CONTAINHENT SPRAY (CSAS)                                                                                  F F                    +
r M                    y                                                          A. Sensor / Trip Units                                                                                  M U                    !$
Containment Pressure --                                                                          Q W                                                                                      High - H10h                    4              2              3            1,2,3        13*, 14* g B. ESFA System Logic M                                                                                  1. Hatrix Logic                  -6              1            3            1,2,3        17      C M                                                                                                                                                                                      m
: 2. Initiation Logic                4(c)          2(d)          4            1,2,3,4      12      m
: 3. Manual CSAS (Trip Buttons)      4(c)          2(d)          4            1,2,3,4      12 N
C. Automatic Actuation Logic          2              1              2            1,2,3,4      16 O
A
 
I s                                                                TABLE 3.3-3 (Continued)
  ! o
    <                              ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION W
* R i
MINIMUM c                                                                TOTAL NO. CHANNELS  CHANNELS      APPLICABLE 5
    -4 ESFA SYSTEM FUNCTIONAL UNIT                                  OF CHANNELS  TO TRIP    OPERABLE        MODES                          ACTION O  "
IV. MAIN STEAM LINE ISOLATION (MSIS)                                                                                                            O O          A. Sensor / Trip Units                                                                                                                  o-Z q              1. Steam Generator Pressura -                      4/ steam      2/ steam  3/ steam      1, 2, 3(b), 4                    13*, 14* Z y                  Low                                              generator    ' generator generator q
O p
2- ste liigh ceaer ter te ei -                        4/ste -
generator 2' te --
generator 3'ste -
generator
: 1. 2. 3. 4                      13*. 14*
O
: 3. Containment Pressure - liigh                    4            2          3              1,2,3,4                          13*, 14*
O  s.      B. ESFA System Logic O
. E                                                                                                                                                    W N              1. Hatrix Logic                                    6            1          3              1,2,3,4                          17 f
C              2. Initiation togic                                4cc)          2(a)      4              2, 2, 3, 4                      12 C
: 3. Manual MSIS (Trip Buttons)                      4(c)          2(d)      4            - 1,2,3,4                          12 N          C. Automatic Actuation Logic                            2            1          2              1,2,3,4                          16      %
e
                                                                                                        **-      e -,e- .m o see en aw mee
 
O                                                                            O                                      O i
* TABLE 3.3-3 (Continued)
I 5        .
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION si k                                                                                                  MINIMUM
      ,                                                                        TOTAL NO.      CilANNELS CHANNELS APPLICABLE c    ESFA SYSTEM FUNCTIONAL UNIT                                        OF CHANNELS    TO TRIP  OPERABLE  MODES    ACTION 5
V. RECIRCULATION (RAS)
      ]
O O                ^- Seaser'Tria "aits o
Z q
t Reueiino water Stor^9*
Tank - Low                                      4              2        3        1,2,3      13*, 14* Z B. ESFA System Logic                                                                                              M
                  @    1. Matrix Logic                                      6              1        3        1, 2, 3) 4 17
: 2. Initiation Logic                                4(c)            2(d)      4        1,2,3,4    12
: 3. Manual. RAS                                      4(c)            2(d)      4        1,2,3,4    12 i              4                                                                                                                      W d                C. Automatic Actuation Logic r.y.e e .:):n/
2              1        2        1,2,3,4    16 C        VI. AUXILIARY FEEDWATER (SG-1)(AFAS-1)
{
A. Sensor / Trip Units                                                                                            M FT1 N                    1. Steam Generator #1 Level -                                                                              %
Low                                              4              2        3        1,2,3      13*, 14*
ect        2. Steam. Generator IA Pressure - SG2 > SG1                            4              2        3        1,2,3      13*, 14*
pf'      f,. h 1 .. c.... d..i -
u; , . . ( 4 w d .tL
 
                                                                                                      ~    '
o                                                                            m g's                                                                  TABLE 3.3-3 (Centinu:d) e,              y
  .o                      f                        ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION N
o          I MINIMUM
  ;p, g
f' E SF  4 SYSTEM FUNCTIONAL UNIT TOTAL NO.      CilANNELS    CilANNELS    APPLICABLE OF CilANNELS    TO TRIP      OPERABLE        MODES    ACTION 7                VI.      AUXILIARY FEEDWATER (SG-1)(AFAS-1) (Continued) g                        B. ESFA System Logic U                              1. Matrix Lngic                        6              1            3            1,2,3 O  w
: 2. Initiation Logic                    4(c) 17 b
2(d)'        4              1,2,3,4    12
: 3. Manual AFAS                          4(c) y                                                                          2 2(d)        4              1,2,3,4    15 2
Automaic,gc,tuationLogic                                1            2            1,2,3,4    16              d
%                  VII...AUXIL,IA,RY FEEDWATER (SG-2)(AFAS-2)
O F" U
: a. see er/ Trig u its                                                                                            O
: 1. Steam Generator #2 Level -
Low                                  4            2            3            1,2,3      13*, 14*
m                              2. Steam GeneratorsA U  "                                            -
4 PressurI h .SGI > SG2;v3 tra St b(NMr64;4          2            3            1,2,3      13*, 14*        Q W                            B. ESFA System L'ogic        7                                                                                      g d                                  1. Mat.rix Logic                        6            1            3            1,2,3      17 2.
{                                      Initiation Logic                    4(c) 2(d)          4            1,2,3,4    12              C
: 3. Manual AFAS                          4(c)          2(d)          4            1,2,3,4    15 C. Automatic Actuation Logic                  2            1            2            1,2,3,4    16              N f
VIII.      LOSS OF POWER (LOV) N                                                                        .. _
A. 4.16 kV Emergency Bus Under-
        !.                        voltage (Loss of Voltage)                4/ Bus        2/ Bus        3/ Bus        1, 2, 3    13*, 14* .
          ;                  B. 4.16 kV Emergency Bus Under-                                                                                    i l'
voltage (Degraded Voltage)                4/ Bus        2/ Bus        3/ Bus        1, 2, 3    13*, 14*
IX.      CONTROL ROOM ESSENTIAL FILTRATION            2              1            1            All Modes # 18*
I
 
CONTROLLED BY USER TABLE 3.3-3 (Continued)
TABLE NOTATIONS (a) In MODES 3-6, the value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is reduced, provided the margin between the pressurizer pressure and this value is maintained at less than or equal to 400 psi; the setpoint shall be increased automatically as pressurizer pressure is increased until the trip setpoint is reached. Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 500 psia.
(b) In MODES 3-6, the value may be decreased manually as steam generator pressure is reduced, provided the margin between the steam generator pressure and this value is maintained at less than or equal to 200 psi; the setpoint shall be increased automatically as steam generator pressure is increased until the trip setpoint is reached.
(c) Four channels provided, arranged in a selective two-out-of-four configuration (i.e., one-out-of-two taken twice).
(d) The proper two-out-of-fouitcombination.                                                            I
* l The provisions of Specification 3.0.4 are not applicable.                                          I b7#' ^ After the initial ~ criticality of Unit 2 or Unit 3.
                  ~
                                                                                ~
ACTION STATEMENTS ACTION 12 -
With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
ACTION 13 -
With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or tripped condition within 1 hour. If the inoperable channel is                ;
bypassed, the desirability of maintaining this channel in the bypassed condi ion shall be reviewed in accordance with g                      Specification        Z13Qd) The channel shall be returned to (OPERABLt sta          no later than during the next COLD SHUTDOWN.
S.ap'ca.c sr,R With a channel process measurement circuit that affects multiple
* functional units inoperable or in test, bypass or trip all associated functional units as listed below.
l I                                Process Measurement Circuit                -
,                  ~(            1. Steam Generator Pressure -
Low Steam Generator Pressure - Low Mfd Steam Generator Level *1-Low (ESF) 4        {                                                        Steam Generator Levelv2-Low (ESF)
: 2. Steam Generator Level              Steam Generator Level - Low (RPS)
(Wid.e Range)
Steam Generator Levelv1-Low (ESF)
Steam Generator Level"2-Low (ESF)
PALO VERDE - UNIT 1                                3/4 3-23 i
: l.                                    CONTRO'. LED BY USER
 
CONTROLLED BY USER TABLE 3.3-3 (Continued)
ACTION STATEMENTS (Continued)
ACTION 14 -          With the number of channels OPERABLE one less than the Minimum Channels OPERABLE, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied:
: a. Verify that one of the inoperable channels has been bypassed and place the other inoperable channel in the tripped condition within 1 hour.
: b. All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below:
          ,                  ,/ Process Measurement Circuit            Functional Unit Bypassed / Tripped j 1. Steam Generator Pressure -            Steam Generator Pressure - Low (895) j          Low LA                                                    Steam Generator Levelil - Low (ESF)
Steam Generator LevelG2 - Low (ESF) k    2. Steam Generator Level - Low    Steam Generator Level - Low (RPS)
(Wide Range)                    Steam Generator Level:1 - Low (ESF)
Steam Generator LevelW2 - Low (ESF)
STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP and/or POWER OPERATION may continue if one channel is l                                restored to OPERABLE status and the provisions of ACTION 14 are j                                satisfied.
ACTION 15 -            With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE
      "                          status within 48 hours or be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.
ACTION 16 -            With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to I hour for surveillance testing provided the other channel is OPERABLE.
ACTION 17 -            With the number of OPERABLE channels one less than the Minimum
!                                Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the l                                next 6 hours and in COLD SHUTDOWN within.the following 30. hours.
s ACTION 18 -            With the number of OPERABLE channels one less than the Minimu[n
! BOP '
Number of Channels, operation may continue for up to 6 hours.
After 6 hours operation may continue provided at least 1 train N
          \                      of efsential filtration is in operation, otherwise, be in HOT        ,
STANDBY within the next 6 hours and in COLD SHUTDOWN within the.
                                      -lowing 30 hours.                                          f PALO VERDE - UNIT 1                            3/4 3-24 CONTROLLED BY USER ~
 
(oc
* TABLE 3.3-4
                                                                                                                                        '1                            2 2,
5                                                                    ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES i;R B
m ESFA SYSTEM FUNCTIONAL UNIT                                                                  TRIP SETPOINT          ALLOWABLE VALUES
            .                    I.                  SAFETY INJECTION (SIAS)
E                                            A.                        Sensor / Trip Units
                                                                                                                              , - - - ~ .                . - . .,
: 1. Containment Pressure - High        -4 ' < 3.0 psig      g < 3.2 psig N
: 2. Pressurizer Pressure - Low        4'. 1837 psia (  y(      1822 psia Not.Appiicble
                                                                                                                                    -                        ~
B.                        ESFA System Logic                                                    Not Appliiable
: Z                                                    C.                        Actuation Systems                            Not Appitcabie          Not Appitcabie      Z II. CONTAINHENT ISOLATION (CIAS)                                                                                                        d N                                                    A.                        Sensor / Trip Units                                                                      N O        m                                                                      1. Containment Pressure - liigh        -E ~< 3.0 psig N  q.
                                                                                                                                                      ~
                                                                                                                                                        < 3.2 psig'N        %#
A                                                                      2. Pressurizer Pressure - Low          k [ 1837 psia h    1      k1822 psi          p==
M        $                                            B.                        ESFA System Logic                            Not Applicable    'NotAppIlcable
                                                                                                                                                                          ~
FT1 C.                        Actuation Systems                            Not Applicable          Not Applicable      U
  @                                III. CONTAINHENT SPRAY (CSAS)                                                                                                            W A. Sensor / Trip Units N
C                                                                                  l. Containment Pressuret Higi - liigh 'f 18.5 psib          20 1 8.9 psig )
{
  .M                                                    B.                      ESFA System Logic                            Not Applicable          Not Applicable      M M                                                      C.                      Actuation Systems          -
Not Applicable          Not Applicable      M IV. MAIN STEAM LINE ISOLATION (MSIS)
A.                        Sensor / Trip Units                                --
: 1. Steam Generator Pressure - Low      $f 1919 psia (3)              912 psiaI3) i
: 2. Steam Generator Level - liigh      .k < 91.0% NR(2)    1T < 91.5%      \ T NR(2) j
: 3. Cor.tainment Pressure - liigh      .['    3.0 psig y # W1        3.2 psig , /
B.                      ESFA System Logic                            Not Applicable          Not Aiplicable l
    '* M e- Og[;c.wr.C .; %L .
 
n h
l
                                                  =
TABLE 3.3-4 (Continued)              g s
ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES E                                      ESFA SYSTEM FUNCTIONAL UNIT                                                            TRIP VALUES                          ALLOWABLE VALUES h                                      V. RECIRCULATION (RAS)
E                                            A. Sensor / Trip Units
[                                                  1 Reiueling Water Storage Tank - Low [7 4% of                                                            -fi      > % of Span >i[)
h                                              B. ESFA System logic                                                          N2 App 1Tcable                      Not A511cib'le O            e                            j            C x.. Actuation System                                  (gp n3. }                Not Applicable                      Not Applicable            O l      O                                                  VI. @ n va w sXILfARY> FEE 0 WATER (SG-1)(AFAS-13 Z
I H
j
                                                                ^- Seasar/Tri, Unit,
: 1.      Steam Generator #1 Level - Low eg. A ,25.8% WR
                                                                                                                                                              ~~
I4)                    7-
                                                                                                                                                                                %/ > 25.3% WR
                                                                                                                                                                                              ~I4) s z
M N                                                            2.      Steam Generator a Pressure GD                            $ < 185sid          p #/            Mk < 192 psid ' '                  N Q                                                                  c'SG2 > SG1Xy42> y d ')-liqh <
e                                                    C-Q
{                                          B. ESFA System Logic                                                          Not Applicable                      Not Applicable m            y;                                        J.,Jctua,gonSys,tems                                  (ac gg)                    Not Applicable                      Not Appilcable            m U            $                                    VII.$UXILIARY2FEEDWATER(SG-2)(AFAS,2)                                                                                                                  U g                                        g              A. Sensor / Trip Units
* g
                                                                                                                                                                                -F > 25.3% WR I4)
_ . .I4) f                                            {
i
: 1.      Steam Generator #2 Level - Low                            -:f,1, > 25.8% WR l
f
: 2.      Steam Generator A Pressure t-J                                                                    T' < 192 psid /
C                                              \                    <SG1 > SGa am s m q . y . 9 71'u< 185 psida                            '=        -                  C M                                                \        B. ESFA Syst'em Logic                                                          Not Applicable                      Not Applicable M
y                                                  Q    ,
C._ Actuation Systems
                                                                                                                      ~~
Not Applicable                      Not Applicable            y
                                                          'VIII. LOSS OF POWER' ~                                                                    - - - - - _ .                              ~
A. 4.16 kV Emergency Bus Undervoltage                                                                                              n
                                                        /            (Loss of Voltage)                                                          > 3250 volts                        > 3250 volts      N B. 4.16 kV Emergency Bus Undervoltage                                          2930 to 3744 volts                  2930 to 3744 volts (Degraded Voltage)                                                          with a 35-second
* with a 35-second    I maximum time delay                  maximum time delay  '
                                                      \IX. CONTROL ROOM ESSENTIAL FILTRATION                                                    > 2 x 10 .a pCi/cc                  < 2x10 s pCf/cc w-        _ _ _
e
 
CONTROLLED BY USER
                                                                                        ''~-
                                      / TABLE 3.3-4 (Continued)                                                      ;
                  /p'                                TABLE NOTATIONS I
                                                                                                    \
(1) In MODES 3-6, value may be decreased manually, to a minimum of 100 psia, as pressurizer pressure is redJced, provided the margin between the pres-surizer pressure and this value is maintained at less than or equal to              i  .
  '. mb                400 psi; the setpoint shall be increased automatically as pressurizer                  ;
i        i            pressure is increased until the trip setpoint is reached. Trip my be                  i l        l            manually bypassed below 400 psia; bypass shall be automatically reinoved              t i            whenever~ pressurizer pressure is greater than or equal to 500 psia.    .
f I    1 (2) % of the distance between steam generator upper and lower level narrow                l range instrument nozzles.
f (3) In MODES 3-6, value may be decreased manually as steam generator pressure            l is reduced, provided the margin between the steam generator pressure and          j this value is maintained at less than or equal to 200 psi; the setpoint          I i              shall be increased automatically as steam generator pressure is in:reased      ;
        ;              until the trip setpoint is reached.
i (4) % of the distance between steam generator upper and lower level wide                /
              \
range instrument nozzles.                                                    /
w'                l t
                                                                                        -                        e i
I r
I e
e o
PALO VERDE - UNIT 1                      3/4 3-27
            ~ ~
CONTROLLED BY USER
 
CONTROLLED BY USER                                                ll TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION                      RESPONSE TIME IN SECONOS
: 1. Manual
: a. SIAS Safety Injection (ECCS)                      Not Applicable Containment Isolation                        Not Applicable Containment Purge Valve Isolation            Not Applicable
: b. CSAS Containment Spray                            Not Applicable
: c. CIAS                                      .
Containment Isolation                        Not Applicable
: d. MSIS Main Steam Isolation                          Not Applicable
: e. RAS Containment Sump Recirculation                Not Applicable (f S ,.D      f. AFAS' ~                                                N Auxiliary Feedwater Pumps                    Not' Applicable g,          M . C.C.65 Ced e d U N                                      Nu fighAL._                    l GD          9.esas% m,yc3 FAik                    g              W = kphaLu l
l PALO VERDE - UNIT 1                    3/4 3-28 CONTROLLED-BY USER
-        -                                                                                  ~
 
CONTROLLED BY USER A  "
TABLE 3.3-5 (Continued) u
                                                                                                      .s
                                    ~
ENGINEERED !aFETY FEATURES RESPONSE TIMES f                                              _-
INITIATING SIGNAL AND FUNCTION                          RESPONSE TIME IN SECONDS
: 2.      Eressurizer Pressure - Low f                                                        .
: a. Safety Injectior.,(HPSI)                        1 30*/30**                                        l;
: b. Safety Injection (LPSI)                        1 30*/30**
c    Containment Isolation                                                                    ,
t                                                                                                                        '
      .                                1. CIAS actuated mini purge valves            < 10.6*/10.6**                                    i 4
1
: 2. Other CIAS actuated valves
        \                                                                              531*/31**                              !          !
l i              3.      Containment Pressure - High
                                                                                                                              /
: a. Safety Injection (HPSI)                          1 30*/30**
: b. Safety Injection (LPSI)                          1 30*/30**                          !
i
: c. Containment Isolation i
i                  1. CIAS' actuated mini purge valves            < 10.6*/10.6**
i                2. Other CIAS actuated valves                I 31*/31**                    i t
: d. Main Steam Isolation
: 1. MSIS actuated MSIV's                        < 5.6*/5.6**
i                  2. MSIS actuated MFIV's#                      7 10.6*/10.6**
                                                                                      ~
i
!              l              e. Containment Spray Pump J                                                                          1 27*/17**                  .
: 4. Containment Pressure - High-High
: a. Containment Spray
(                                                                      1 33*/23**
{      5. Steam Generator Pressure - Low l
j                a. Main Steam Isolation,                                                      '
: 1. MSIS actuated MSIV's                      < 5.6*/5.6**            .
: 2. MSIS actuated MFIV's#                      510.6*/10.6**
l                      l6.      Refeling Water Tank - Low                                    *            ,
f          a. Containment Sump Recirculation                  ,,45*/45**
t        7. Steam Generator Level - Low I                  a. Auxiliary Feedwater (Motor Drive)                i 46*/23**
I
(                    b. Auxiliary Feedwater (turbine drive) 1 30*/30**
PALO VERDE - UNIT 1                        3/4 3-29 l
CONTROLLED BY USER                                                                      ~
 
  .                                                                                                                                          .o
                                                                                                                            .                I CONTROLLED BY USER                                                                          :  i TABLE 3.3-5 (Continued)                                              4
                                  ~ ,              ~ ~ . ~ _            _-.            . . - . -    --
l,
                              -                    ENGINEERED SAFETY ~ FEATURES RESPONSE TIMES
                                                                                                  ~
INITIATING SIGNAL AND FUNCTION                            RESPONSE TIME IN SECO'NDS                      .!
              /                                                                                                                '
l;          8. Steam Generator Level - High I
l I
: a. Main Steam Isolation
: 1. MSIS actuated MSIV's                        < 5.6*/5.6**                                        !
c,
: 2. MSIS actuated MFIV's#                      310.6*/10.6"*
                                                                                                                              /
: 9.      Steam Generator AP-High-Coincident With Steam Generator Level Low                                            i
: a. Auxiliary Feedwater Isolation                    i 16*/16**
from the Ruptured Steam Generator
(            10. Control Room Essential Filtration Actuation                5 180*/5 180**
I'
: 11. 4.16 kV Emergency Bus Undervol-tage (Oegraded Voltage)
Loss of Power 90% system voltage                1 35.0
                      \ 12. 4.16 kV Emergency Bus Undervoltage (loss of Voltage)
                        \              Loss of Power                                  1 2.4 l                        TABLE NOTATIONS l
* Diesel generator starting and sequence loading delays included. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.
lod.
aa uquence          'M                              .
! ' (*b '?'              ** Diesel generator starting 7 delays not included. Offsite power available.
:                            Response time limit includes movement of valves and attainment of pump or i                            blower discharge pressure.                                            .
Qy (#MFIVvalvestestedatsimulatedoperatingconditions;valvestest static flow conditions to < 8.6/8.6 seconds.                                            /
* See Apchcds SC Q. (< l2.es p:nse Ti re S .
l                        PALO VERCE - UNIT 1                      3/4 3-30 l
f l                                              CONTROLLED BY USER                                                        '
 
i TABLE 3.3-5        (Cont'd)                            ,s,'
I                                      ENGINEERED SAFETY FEATURES RESPONSE TIMES
                                                                                                            ~
                      ,/'                                    ~~
fINITIATING SIGNAL AND FUNCTION                                    RESPONSE TIME IN SECONOS 4              ,
l
,            l2. Pressurizer Pressure-Low
: a. Safety Injection (ECCS)                                        <    */    **
            ,'          b.- Containment Isolation                                          <    */    **
f I
l 3.          Containment Pressure-High                                                                    l'
: a. Safety Injection (ECCS) i                                                                                  < _
                                                                                                  */    **
: b. Containment Isolation                                          <    */    **              ,
f                  c. Main Steam Isolation                                          <    */    **
l i
      !            4. Containment Pressure--High-High                                                            j
: a. Containment Spray                                              <    */    **
I
    !              5. Steam Generator Pressure-Low                                                              t 1                  - a. Main Steam Isolation                                          .<    */    **
,'      t
        ;          6. Refueling Water Tank-Low.
            \            a. Containment Sump Recirculation                                <          **
                                                                                                  */                  l 1
l i
: 17. Steam Generator Level-Low                                        -
f j        a. Emergency Feedwater                                              <    */    **
l l
{ 8.      Steam Generator Water Level-High                                                            i j        a. Main Steam Isolation                                            <
_    */    **
I9. Steam Generator AP-High-Coincident With Steam Generator Level Low I
              ,          a. Emergency Feedwater Isolation from Ruptured                    <    */    **          '
Steam Generator                                                                          +
t                                                                                                        I
            \r                                                                                                    l
              '.                                                                                                i
                                ~ "' ' ~        ..
Amendment Number 9 February 27, 1984 3/4 3-28 L
 
i
    ! ,                                                    TABLE 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS m                                                .                                  CHANNEL    MODES FOR MlICH
      '                                                    CHANNEL          CHANNEL  FUNCTIONAL      SURVEILLANCE g  ESFA SYSTEM FUNCTIONAL UNIT                        CHECX        CALIBRATION      TEST        IS REQUIRED I. SAFETY INJECTION (SIAS)
O              a- Seasarnr'a uaits
: 1. Containment Pressure - High      S            R            H            1,2,3,4 o
d                  2. Pressurizer Pressure - Low        S            R            H            1, 2, 3, 4      H B. ESFA System Logic                                                                                  N p                  1. Matrix Logic                      NA          NA            M            1, 2, 3h m                  2. Initiation Logic                  NA          NA            M            1, 2, 3, 4 0    h            3. Manual SIAS                      NA          NA            M            1,2,3,4          U SAU-)
E              C. Automatic Actuation Logic                NA          NA            M(1),<(2f(3)) 1, 2, 3, 4      W II. CONTAINMENT ISOLATION (CIAS)
C                  Sensor / Trip Units                                                                                C M              A.
M                  1. Containment Pressure - High          S            R            H            1,2,3            m
: 2. Pressurizer Pressure - Low        S.          R            H            1, 2, 3          N B. ESFA System Logic Matrix Logic                                                              1, 2, 3h'
            @      1.                                      NA          NA            M
: 2. Initiation Logic                  NA          NA            M            1, 2, 3, 4
: 3. Manual CIAS                      NA-          NA            M            1, 2, 3, 4
                  ' 4. Manual SIAS                      NA          NA            M            1,2,3,4)_
g>
 
b              _.
i E                                                TABLE 4.3-2 (Continued)
G                      ..
                            <-          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIRENENTS 9
El
                              .                                                                                    CilANNEL    MODES FOR WlICil c-                                                            CHANNEL    CllANNEL  FUNCTIONAL      SURVEILLANCE
                    .      5  ESFA SYSTEM FUNCTIONAL UNIT                              CHECK  CALIBRATION        TEST        IS REQUIRED O          H II. CONTAINMENT ISOLATION (Continued)                                            c, O
h      C. Automatic Actuation Logic                    NA      NA M(1)j Q        1, 2, 3, 4 O
H                                                                                                                              H
;                y              III. CONTAINMENT SPRAY (CSAS)                                                                                  y-O r-r"        4:'
A. Sensorerri, Units
: 1. Containment Pressure --
o r-
* y-ITI High - liigh                              S      R            H              1, 2, 3 m
O          I m                                                                                                                  O g                      B. ESFA System Logic g
h    1. Matrix Logic                              NA      NA            M              1,2,364)        N C                          2. Initiation Logic                          NA      NA            M              1,2,3,4          C N                                                                                                                              M lTI                        3. Manual CSAS                              ;NA-    HA            M              1, 2, 3, 4 m
N                                                                    '
L  '. L2                  N h
C. Automatic Actuation Logic                    NA      NA            M(1) (2)      1, 2, 3, 4 I                                I l                                                      >  -
 
i e
l @                                                                TABLE 4.3-2 (Continued)                                          ,
                                                    . E
                                                      <                                ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
                                                      ,                                                                                                CilANNEL      MODES FOR WHICil
: c.                                                                      CilANNEL      CilANNEL  FUNCTIONAL        SURVEILLANCE ESFA SYSTEM FUNCTIONAL UNIT                  CHECK      CALIBRATION      TEST        IS REQUIRED H                          IV. MAIN STEAM LINE ISOLATION (MSIS)                                                                O O                                      A. Sensor / Trip Units                                                                        O Z
l H                                          1. Steam Generator Pressure -
Z q
;                                              y                                              Low                            S        R            H              1,2,3,4 O                                          2. Ste    ceaer tor Levei - aioh  S        R            a
                                                                                                                                                              ~
: 1. 2. a. 4      O g                                    3. Containment Pressure - High-    S        R            H              1, 2, 3, 4
,                                              m    _                                                                                                                                m
                                                      $                                B. ESFA System logic                                                                          U
: 1. Matrix' Logic                  NA        NA          M              1,2,3,4          W
: 2. Initiation Logic              NA,        NA          M              1,2,3,4 C                                                                                                                                      C M                                          3. Manual MSIS                    NA        NA          M              1,2,3,4          g i ''
                                                                                                                                                            .5A(2)                    M h        C. Automatic Acutation Logic          NA        NA M(1)g2)      )  1, 2, 3, 4      N e
 
O'
      !                                                          TABLE 4.3-2 (Continued) i i 5 5          ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
        <  }
E  '
l      S                                                                                                        CHANNEL          MODES FOR MIICH
        .                                                                              CilANNEL    CHANNEL  FUNCTIONAL            SURVEILLANCE c-                                                                              CllECK    CALIBRATION      TEST            IS REQUIRED fESFASYSTEMFUNCTIONALUNIT p    ]
V. RECIRCULATION (RAS)
O              a-  seaser/Triau ait-O Z                                    Reruenng water Stora9-                                                                                      Z H                                    Tank - Low                                      S        R          H                    1, 2, 3          y B,  ESFA System Logic                                                                                                            N Matrix Logic                                    NA      NA          M          h        1, 2, 3Cp
: 2.              InihationLogic                                  HA        NA          H                    1, 2, 3, 4 h              3.              Manual RAS                                      NA        NA          M                    1, 2, 3, 4        O C. Automatic Acutation                ic                                            ('h M(1 '-(2 [3)                          E N          VI.
Enug<,,e q                            o?6&h!)
NA      NA 1, 2, 3, 4 f
(AUXILIARYJFEEDWATER(SG-1)(AFAS-1);
C                                                                                                                                                C M              A. Sensor / Trip Units                                                                                                          g
: 1.              Steam Generator #1 Level '      ' '
Low                                            S.        R          H                    1,2,3            N
: 42. Steam Generator P Nin't.
Q Pressure SG2 > SGI S        R          H                    1, 2, 3 0 !e .f 4 2 %'y. 0.%<vr!ll
                                                                                                              --e
 
  .                                                                            f                    -
g              h
    !            k                                                                    TABLE 4.3-2 (Continued) l                      ENG NEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
        ;f,                                                                                                                      CllANNEL        MODES FOR WHICH ig                                                                                      CHANNEL          CHANNEL      FUNCTIONAL          SURVEILLANCE m          ESFA SYSTEM FUNCTIONAL UNIT                                                  CHECK        CALIBRATION          TEST          IS REQUIRED h          VI.      AUNLIARYFEEDWATER(SG-1)(AFAS-1)(Continued) h                    B.;ESFSSystemLogic H                        1. MatrhLogic                                                NA            NA      g      M                1, 2. Q[')                  O 2
: 2. Initiation Logic
                                                                \3 NA            NA              M                1,2,3,4                    0
: 3. Manual AFAS                                                NA            NA H      s/p M(1),4(p 1,](3) 1,2,3,4 2
CyAutomati,cActuationLogigpg                                        NA            NA      TM                          2, 3, 4                  d              -
              ]VII.(AUXILIARY!FEEDWATER(SG-2)(AFAS-2)>                      F                                                                                                N l                            A. Sensor / Trip Units O
R                        1. Steam Generator #2 Level -                                  .
m    {                              Low                                                        S            R              H                1,2,3 m
Q    J, u'
                .i 01
: 2. Steam _ Generator. @--wm .
iA Pressure SGI > SG2'i                                    S            R              H                1, 2, 3                    Q
  @                          B. ESFA$yNhmL$iskN ""                                                                                                                        W
: 1. Matrix Logic                                              NA                                              1,2,3;[4)
C                                  Initiation Logic NA
                                                                                                                            )H          .
N 1      2.                                                            NA            NA              M          ,
1, 2, 3, 4 C
: 3. Manual AFAS                                            .NA              NA              M    st,(2.)    1, 2, 3, 4                M Y  C. Automatic Actuation Logic .                                    NA            NA
                                                                                                                        />$) M(1) (2) (3)
                                                                                                                          ~
1, 2, 3, 4                E
                                                                                                                                            ~
VIII. LOSS OF POWER (LOV)                                                            - - - -
                                                                                                                ~.
      .                    A. 4.16 kV Emergency Bus Under-                                                                                  '
                                                                                                                                                    ~s s.
voltage (Loss of Voltage)                                    ,
5              R              R                1, 2, 3, 4 N s B. 4.16 kV Emergency Bus Under-voltage (Degraded Voltage)                                    S              R              R                1, 2, 3, 4          i e
e e
w
 
CONTROLLED BY USER TABLE 4.3-2 (Continued)
Op3                _____--
TABLE NOTATION            .
                                                                                                                ~ _ .
s
                        / (1) Each train or logic channel shall be tested at least every 62 days j            on a STACGERED TEST BASIS.
(2) Testing of automatic actuation logic shall include energization/
deenergization of each initiation relay and verification of proper                                ;
l                    operation of each initiation relay.                                                                -
k              (3) A subgroup relay test shall be performed which shall include the
            \                    energization/deenergization of each subgroup relay and verification
              !                  of the OPERABILITY of each subgroup relay. Relays listed below are i                exempt from testing during POWER OPERATION but shall be tested at                                  '
                ;              least once per 18 months during REFUELING and during each COLD i              SHUTDOWN condition unless tested within the previous 62 days.
h                                                                                                          '
i                  !                  ACTUATION DEVICES THAT CANNOT BE TESTED AT POWER TRAIN A                                    TRAIN B l
ESF            ACTUATION                  ESF                  ACTUATION FUNCTION        DEVICE                      FUNCTION            . DEVICE i SIAS A              K108
!                      i SIAS A            K409                      SIAS SIAS B B'            K108 K409 t
CIAS A          K202                      CIAS B                K204
                      / CIAS A              K204                      CIAS B                K205
                  /'CIASA                  K205                      CSAS B                K304 CSAS A          K304                      MSIS B                K305
            ,/ -- MSIS A                  K305              ,      MSIS B                K404
            '            MSIS A            K404                      AFAS la                K113
          /              AFAS 1A          K211                      AFAS 1B                K' 211-l                        AFAS 2A          K112                        AFAS 2B                K112-l i
In the case of the following relays which are tested durino power coeration, one or more pieces of equipment cannot be actuated, but can be racked out, bypassed or etc., which will not preclude the relay from being tested but will not actuate the locked out. equipment associated with the relay:
    .                    SIAS A          K401                        SIAS B                K301 SIAS A          K410                        SIAS B                K308 i                    SIAS A          K412                        CIAS B                K203 j                    CIAS A          K203                        CIAS B                K210
:                  CIAS A          K210                        RAS B                K104 i                  RAS A            K104                        RAS B                K312
        \                RAS A.          K312                      RAS B                  K405
          \              RAS A            K405
            \ AFAS 1A                      K113
                      'NN
                                              ~                            -
PALO VERM - UNIT 1                    3/4 3-36 l
                                    ~                                                                                        ~
i                                            CONTROLLED BY USER' l
 
I i
b                  1 p~.---_..          TABLE 4.3-2  (Cont'd)                                    l
                                              ~~                      _-            -
x
              '                                                                                  3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM                      N ~~!'
INSTRUMENTATION SURVEILLANCE REQUIREMENTS TABLE-NOTATION
    ,  (1) Testing of automatic actuation logic shall include energization/de-ener-
  ,          gization of each initiation relay and ~ verification of proper operation of each initiation relay.
(2) Testing of the actuation relays shall include the energization/de-energi-zation of each actuation relay and verification of proper operation of each actuation relay.
                ~,_
f Amendment Number 9 February 27, 1984 3/4 3-35
 
CONTROLLED BY USER                                                                        !
                -            w                                                                                                I.
    ,d ([ INSTRUMENTATI g                                                                                                    I 3/4.3.3 MONITORING INSTRUMENTATION                                                                          l RADIATION MONIT') RING INSTRUMENTATION LIMITING CONDITION FOR OPERATION AN3 S uc v6 LL4NCA flCStu CEM6t)G 3 /4 3 3.[    [acibi.HGu. (-t W *t @ 'Lu h atw t.W m 3T3.1 Theradiationmoni$ringinstrumentationchannelsshownin (Table 3.3-6shallbeOPERABLEwiththeiralarm/tripsetpointswithinthesX $.#.-
specified limits.                                                              \            4{[2/uwA 5,g.({
            ,    APPLICABILITY: As shown in Table 3.3-6.                                          k
                                                                                                    \
  -@      j    ACTION:
: a. With a radiation monitoring channel alarm / trip setpoint exceeding                          ,
the value shown in Table 3.3-6, adjust the setpoint to within the                            '
limit within 4 hours or declare the channel inoperable.                    ;
: b. With the number of channels OPERABLE one less than the Minimum            !
Channels OPERABLE requirement, take the ACTION shown in Table 3.3-6.
                                                                                                                ~
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. .
i l
l' SURVEILLANCE REQUIREMENTS l        i I    4.3.3.1 Each radiation monitoring instrumentation channel shall be                        f demonstrated OPERABLE by the performance of the CHANNEL' CHECK, CHANNEL                l CALIBRATION, and CHANNEL FUNCTIONAL TEST operations for the MODES and at the                            ,
frequencies _shown in Table 4.3-3.                                                /
                                                                        ~
                                                                                  ~
                                                                                              /
l PALO VERDE - UNIT 1                    3/4 3-37 COF TIOLLD B'' USSR
                                                                                                .                              1
 
N W
f%.-                                                  .
                                                                                                                                                          .y          / TABLE 3.3-6                                                  ~ ~ ~
o                                                                            RADIATION MONITORING INSTRUMENTATION y
                                                                                                ,/y HINIMUM                                                                  .                                          I M.                                                                  .          CHANNELS        APPLICABLE                                        ALARM / TRIP  MEASUREMENT                      f y                      INSTRUMENT                                                OPERABLE            MODES                                          SETPOINT          RANGE            ACTION              i e            1. Area Monitors E                      A. Fuel Pool Area RU-31                                1                **
                                                                                                                                                                                                                            <15mR/hr      10 8 to 10*mR/hr      22 & 24 4                      B. New Fuel Area RU-19                                  1
* 515mR/hr      10 8 to 10*mR/hr      22 Q                e-.                    C. Containment RU'148 &
RU-149                                              2                                                                  <10R/hr                                          Q 1,2,3,4                                                          1R/hr to 107R/hr O                                        D. Containment Power Access Purge Exhaust RU-37 &
27 RU-38                                                1                #                                                                10 8 to 10 *mR/hr    25
                                                                                                                                                                                                                            ~<2.5mR/hr
                                                    ==l                                                                                                                                                                                                                        d g                                        E. Main Steam
: 1)    RU-139 A&B                                                      1,2,3,4 0
I""                si
: 2)    RU-140 A&B 1
1                1,2,3,4 10 3 to 104R/hr 10 3 to 10*R/hr 27 27 I"''
r==                *f              2. Process Monitors m                  Yi I""
m
;                                                  g                  8!/-
A. Containment Building Atmosphere RU-1                                    2                1,2,3,4                                                                                              Q 23 & 27 W                                                1)    Particulate                                                                                                      <2.3x10 8pCf/cc 10 8 to 10 *pci/cc                  @
Cs-137 C                \                              2)    Gaseous                                                                                                      <6.6x10 =pci/cc 10        to 10 8pci/cc                C M                  \                                                                                                                                                ~ Xe-133                                                M
: m.                \
i      m y                                          B. Noble Gas Monitors Control Room Ventilation y
g'-I                        Intake RU-29 & RU-30 1      !.      ALL MODES                                      <2x10.sp Ci/cc      10 s to 10 8pci/cc 26
: 3. Post Accident Sampling System                                              1,2,3                                          N.A.                N.A.                28 e
1###l.
                                                                                                                                                            ,i
                                                                        !                  *With fuel in the storage pool or buildthg.
                                                                        !                ^^With irradiated fuel in the storage pool.
I
                                                                                          #When purge is being used.
{              ##Three (3) times background in Rem / hour.,
                                                                              '        ###The Minimum Channels Operable will be defined in the Preplanned Alternate Sampling Program                                                                                              -'
of Specification 6.16. .                                        !:
1
                                                                                                                                                            \i      .
: i. !
 
CONTROLLED BY_ USER                                                                        H C ~ 'N' TABLE 3.3-6 (Continued)
                                                                                                  \.s ACTION STATEMENTS                          N
          ;          /                                                                                    N l ACTION 22 -      With the number of channels OPERABLE less than required by the
                    /
          * '                            Minimum Channels OPERABLE requirement, perform area surveys of                              -
the monitored area with portable monitoring instrumentation at        ,
least once per 24 hours.
i l    ACTION 23 -    With the number of channels OPERASLE less than required by the j                      Minimus Channels OPERABLE requirement, comply with the ACTION j                      requirements of Specification 3.4.5.1.
l        ACTION 24 -    With the number of channels OPERABLE less than required by the j                        Minimum Channels OPERABLE requirement, comply with the ACTION i
requirements of Specification 3.9.12 or operate the fuel build-ing essential ventilation system while handling irradiated fuel.
ACTION 25 -    With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, comply with the ACTION            i i                          requirements of Specification 3.9.9.                                      '
L                                                                                                      L ACTION 26 -    With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, within 1 hour initiate                k
{'
l                          and maintain operation of the control room emergency ventilation j                        system in the recirculation mode of operation.
I ACTION 27 -    With the number of OPERABLE Channels less than required by the I f Minimum Channels OPERABLE requirement, either restore the        .
f                          inoperable channel (s) to OPERABLE status within 72 hours, or:              i
: 1. Place moveable air monitor in-line.
l
: 2. Prepare and submit a Special Report to the Commission                  !
pursuant to Specification 6.9.2 within 30 days following                  ',
the event outlining the action taken, the cause of the                    !
inoperability, and the plans and schedule for restoring                    i t
          !                                  the system to OPERABLE status.                                              !
ACTION 28 -    With the number of OPERABLE Channels one less than required by                      i the Minimum Channels OPERABLE requirement, either restore the inoperable channel (s) to OPERABLE status within 7 days, or:                            ,
: 1. Initiate the Preplanned Alternate Sampling Program of Spec-
                  \          __            ification 6.16 to monitor the appropriate parameter (s).
    - -                  N            2. Prepare and submit a Special Report to the Commission
                          \\                pursuant to Specification 6.9.2 within 30 days following the event outlining the action (s) taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.
N                                                          -
N      s N                    __
PALO VERDE - UNIT 1                  3/4 3-39 CONTROLLED BY USER
 
m                  /                                                            TABLE 4.3-3                          ~^%---
j                                RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS m              /
E
                                                      /                                                                                              CilANNEL        MODES FOR WilICll                                        -
                                        '          /                                                                      CilANNEL    CllANNEL  FUNCTIONAL        SURVEILLANCE                                              -
                                                  /      INSTRUMENT
_CilECK    CALIBRATION      TEST          IS REQUIRED y        /      1. Area Monitors                                                                                                                                                      .
O  ~;                    A. Fuel Pool Area RU-31                                          S            R            H                  **
Q Q          f              8. New Fuel Area RU-19                                        S            R            M
* Z                          C. Containment Power Access Purge Exhaust H                                RU-37 & RU-38                                              P#          R          P###,W##            ##
l H
N                          D. Containment RU-148 &                                                                                                                                            N Q
F  R                      E.
RU-149 Main Steam RU-139 A&B 5            R            H              1,2,3,4 Q
y-
* RU-140 A&B                                                                                                                                                      F S            R            H              1,2,3,4                                              /    F.
M                    2. Process Monitors                                                                                                                                                      E O                          A. Containment Building I
I g        !
Atmosphere RU-1
                                                                                                                                                                                                                            /
g
: 1) Particulate                                            S            R            H            1,2,3,4                                            I l                      2) Gaseous                                                S            R            H            1,2,3,4 C        i                8. Controi Room                                                                                                                                        i j
C
(/)      l                      Ventilation Intake                                                                                                                                              y) m      ;                        RU-29 & RU-30                                              S            R            H          All MODES                                          f m
N                    3. Post Accident Sampling System                                    N.A.        R            ***        1,2,3                                                          %
                                                            "With fuel in the storage pool or building.                                                                                                                                    ~
                                                          **With irradiated fuel in the storage pool.                                                                                                                                    *
                                                          ***The functional test should consist of, but not be limited to, a verification of system sampling capabilities.                                                                                                                                  j
                                                            #If purge is in service for greater than 12 hours, perform once per 12-hour period.                                                                      1
                                                          ##When purge system is in operation.
                                                          ###The functional test should consist of, but not be limited to, a verification of system                                                                  [''
isolation capability by the insertion of a simulated alarm condition.                                                                              /                          .
                                                                                                                                                                                                                /
                                                                  ~~              _.,n..-w~            " ~ ~ ~ ~                  '
_---___n,___                                                                                                                    --- - _ . - _ _ - - - - - -
                                                                                                                                                                                                                            - . . _,,_n
 
Y      .    -
                                                                                        ~
CONTROLLED BY USER                                                          q A
INSTRUMENTATION                                                                                    ,hI        -
INCORE DETECTORS                                                                                        I
                                                                                                                  ~
l'.
LIMITING CONDITION FOR OPERATION I
3.3.3.2 The incore detection system shall be OPERABLE with:                              '
                                                                                                                \
:                  a.        At least 75% of all incore detector locations, and l;                b.      A minimum of two quadrant symmetric incore detector locations per core quadrant.
An OPERABLE incore detector location shall consist of a fuel assembly containing a fixed detector string with a minimum of four OPERABLE rhodium detectors or an OPERABLE movable incore detector capable of mapping the location.
l            APPLICABILITY: When the incere detection system is used for monitoring:
: a.      AZIMUTHAL POWER TILT,                                              -
: b.        Radial Peaking Factors,
: c.      Local Power Density,
: d.      DNB Margin.
1            ACTION:                                                          .
1                                                                                            .
With the incore detection system inoperable, do not use the system for 4                theaboveapplicablemonitoringorcalibrationfunctions.j C
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
j' SURVEILLANCE REQUIREMENTS                                                                                      -
4.3.3.2 The incore detection system shall be demonstrated OPERABLE:
: a.      By performance of a CHANNEL CHECK within 24 hours prior to its use if the system has just been returned to OPERABLE status or if 7 days ~
or more have elasped since last use and at least once per 7 days thereafter when required for monitoring the AZIMUTHAL POWER TILT,-
radial peaking factors, local power density or DNB margin:
: b.      At least once per 18 months by performance of a CHANNEL CALIBRATION operation which exempts the neutron detectors but includes all electronic components. The fixed incore neutron detectors shall be calibrated prior to installation in the reactor core.
l PALO VERDE - UNIT 1                        3/4 3-41 l
CONTROLLED BY USER                                                                      l
 
CONTROLLED BY USER q                j,          -
                                      -~ N
        'g. 34' (, INSTRUMENTATION'
          ,s  ,
iT'N            , SEISMIC INSTRUMENTATION
              $      c QMITIh3 CONDITION FOR OPERATION th? 5'tD&h4.fil CC-ituttF.9 ''i        .
          '        , w. 3. s wuc-=%
                                                                                        ,_      3.e,y g u
                                                                                                  ~
                    ' 3.3.3.3 The seismic' monitoring _ instrumentation shown iii~Tabl~e~3.~3-7 stat 1 be .
                                                                                      ~
i
                't.iOPERABLE.                                                                                ,
APPLICABILITY: At all times.
          .,    s      ACTION:                                                    .
: a. With one or more seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining
  ,,.                                the cause of the malfunction and the plans for restoring the instryment(s) to OPERABLE status.
    -v                          b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.3.3.1      Each of the above seismic monitoring instruments shall be
{%)~}            demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in    .
Table 4.3-4.
l                        4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a l                        seismic event (greater than or equal to 0.02g) shall have a CHANNEL CALIBRATION performed within 5 days. Data shall be retrieved from actuated instruments and l
analyzed to determine the magnitude of the vibratory ground motion. A Special
                      . Report shall be prepared and submitted to the Commission pursuant to Specifi-cation 6.9.2 within 10 days describing the magnitude, frequency spectrum, and l
resultant effect upon facility features important to safety.,
N w -_:                          -
n g
PALO VERDE - UNIT 1                    3/4 3-42 CONTROLLED BY USER                                          l l                                          ___ ___      _ _ _    _.
 
a <                                                                                                                        ~
9                                                        .
T                                                    .,
('
CONTROLLED ~BY                  N USER
                            ,p
                                          ~
TABLE 3.3-7        N
                                                                                      \ x SEISMIC MONITORING INSTRUMENTATION            N                              '
                /                                                                  MINIMUM N
i INSTRUMENT                                ';
        ] INSTRUMENTS AND SENSOR LOCATIONS                                        OPERABLE                    i t
                                                                                                                  \
l          1.      TriaxialAccelerometers                    s                                          j
: a. Tendon Gallery Floor, 55' level                      1
: b. R.C.P.. Motor Housing, 129'6" level                                                  s 1
      }                      c. Steem Gs.wrator Base,101'9" level          '
1
      ;                      d. Control Building Floor, 74' level                    1 l
I                      e. Auxiliary Building Floor' 40' level                  1                        ,'
i                      f. 25' E. of Turbine 81dg. W. side x                                            i j                            189'9" 5. of Turbine Bldg. S. Sidei l                          on ground (Ref. Plant N.)                                                      l
                                                                          %              1 l
: 2.      Peak Reading Accelerograph s
f i                                                                                                        1
: a. Aux. Bldg., Valve Gallery, Class                                          j j                          ,1 Pipe, 78'7" level                                  1 i
: 3.      Seismic Triggers                                                                  f'          ~
i
.        I                a. Tendon Gallery Floor, 55' level l          1                      (Setpoint 0.010 g)                                  1                  !
l          !'
: b. Containment Operating Floor,140'                                      l level (Setpoint 0.020 g)                            1
: 4.      Digital Cassette Recorders
                          -a. Control Room Area, 140' level                        1                i l                          b. Control Room Area, 140' level                        1                I
: c. Control Room Area, 140' level                        1                p
: d. Control Room Area, 140' level                        1
: e. Control Room Area,140' level                                          i 1                i
: f. Control Room Area, 140' level                        1                !
i
\
            !5.            Seismic Switches                                                          I
              \
l              \          a. Tendon Gallery Floor, 55' level                                    l 1            /
l                \
I                  \.                              Horizontal      Vertical                      /
                                                                                                  /
l                  'N Setpoint OBE      0.18 g          0.17 g                    '
i                                Setpoint SSE      0.31 g                                  j 0.34 g                /
x                          z 3
l                PALO VERDE - UNIT 1                        3/4 3-43 L
CONTROLLED BY USER
 
CONTROLLED BY USER r
TABLE 4.3-4 n-- -
m.__
                                                                                                    's
                          /,,/  SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS
                        /
CHANNEL
                      /                                                      CHANNEL    CHANNEL    FUNCTIONAL y      INSTRUMENTS AND SENSOR LOCATIONS                  CHECK  CALIBRATION      TEST          s            .
I        1. Triaxial Accelerometers                                                                \
f                  a. Tendon Gallery Floor, 55' level          N.A.        R        SA i                  b. R.C.P., Motor Housing, 129'6" level      N.A.        R        SA
,                                c. Steam Generator Base, 101'9" level        N.A.        R        SA i
    ~,        i                d. Control Building Floor, 74' level        N.A.        R        SA
: e. Auxiliary Building Floor 40' level        N.A.        R        SA
: f. 25' E. of Turbine Bldg. W. side x 189'9" 5. of Turbine Bldg. S. Side j                        on ground (Ref. Plant N.)
l t
N.A.        R        SA                            l f                                                                                                                      1 l          !.              2. Peak Reading Accelerograph 1
                                                                                                                                \
: a. Aux. Bldg., Valve Gallery, Class                                                          j l                                      1 Pipe, 78'7" level                      N.A.        R        NA
: 3. Seismic Triggers
: a. Tendon Gallery Floor, 55' level          N.A.        R        SA                          i I
: b. Containment Operating Floor, 140'                                                          '
          ;                          level                                    N.A.        R        SA
          '.              4. Digital Cassette Recorders I
: a. Control Room Area, 140' level            M          R        SA
: b. . Control Room Area, 140' level            M          R        SA
: c. Control Room Area, 140' level            M          R        SA
: d. Control Room Area, 140' level            M      -
R        SA                      /
i
: e. Control Room Area, 140' level            M          R        SA                    /
: f. Control Room Area, 140' level            M          R        SA                  /
: 5. Seismic Switches                                                                      '
: a. Tendon Gallery Floor, 55' level                                            I' M          R        SA        /
i i
N l
PALO VERDE - UNIT 1                    3/4 3-44 CONTROLLED BY USER l
 
CONTROLLED BY USER                                                            ,
4        INSTRUMENTATION'
              ~.                                                                                              -
METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION I.'.P      <u CJ E. .L.IVE fDt :C':e'.9 'C 3'y.3.3.4        Fuku sl.onid Las.hc An%,                      c ., Q- C .-H %. e1:.lu, Og                                                                      ..e                        -
i 3.3.3.4 The meteorological monitoring instrumentation channeliF'shown'in - _'
ops      Table 3.3-8 shall be OPERABLE.
      ' APPLICABILITY: At all times.
ACTION:
: a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS C\
Vi          4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and
        .'    CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5.
            \
N-_____._____..-
O PALO VERDE - UNIT 1                    3/4 3-45 CONTROLLED BY USER l                                                                  ..                . _ _ _ _ _ _ _ _            _
 
1 l
CONTROLLED BY USER
                                                ./                                                        l TABLE 3.3-8                                          l
                  ,/,._._.-
y
              /                        METEOROLOGICAL MONITORING INSTRUMENTATION        'y
          !                                                                                '\
        ''f MINIMUM                .
INSTRUMENT                            LOCATION                    OPERABLE
    /          1. WIND SPEED I
~
: a. 0* tc 50 mph,            Nominal Elev. 35 feet              1 f                                                                                                      )
: b. 0* to 50 mph,            Nominal Elev. 200 feet              1
: 2. WIND 01,RECTION                                                                  .
: a. 0*-360*-180*,            Nominal Elev. 35 feet              1
      \i
: b. 0*-360*-180*,            Nominal Elev. 200 feet              1
        \    3. AIR TEMPERATURE - DELTA T
: a.    -6*F to 6*F,              Nominal Elev. 35 feet-200 feet      1 x
                    \
j
              " Wind speeds less than 0.6 MPH will be reported as 0.
l PALO VERDE - UNIT 1                    3/4 3-46                                              l CONTROLLED BY USER-                                                      !
 
CONTROLLED BY USER
                ,-                              TABLE 4.3-5        N..,
METEOROLOGICAL MONITORING INSTRUMENTATION          -
        ,!                                SURVEILLANCE REQUIREMENTS j
f                                                      CHANNEL        CHANNEL i      INSTRUMENT                                      CHECK        CALIBRATION        'N i
      '                                                                                            \
: 1. WIND SPEED                                                                        's,          .'
: a. Nominal Elev. 35 feet                  D                SA
                                                                                                          \
        ;          b. Nominal Elev. 200 feet                D                SA                          \
                                                                                                              \i
        ;    2. WIND DIRECTION
        !        a. Nominal Gev. 35 feet                    D              SA                            ,
                                                                                                            /
: b. Nominal Elev. 200 feet                  0              SA                        ,/
                                                                                                        /
: 3. AIR TEMPERATURE - DELTA T                                                            /
f
: a. Nominal Elev. 35 feet - 200 feet        0              SA              ,,
                                                                                              /
                                                                                            /
N                                            , . . '
em          o PALO VERDE - UNIT 1                  3/4 3-47 CONTROLLED BY USER
 
CONTROLLED BY USER
          ' INSTRUMENTATION (b
l REMOTE SHUTDOWN SYSTEM LIMITING CONDITION FOR OPERATION
                                                                                                \  \              ,
                                                                                                        \
3.3.3.5 The remote shutdown system disconnect switches, power, controls and              \  i monitoring instrumentation channels shown in Table 3.3-9 shall be OPERABLE.
    ,          APPLICA8ILITY: MODES 1 and 2.
                                                                                                                }
ACTION:
: a. With the number of OPERABLE remote shutdown monitoring channels h                        less than required by Table 3.3-9, restore the inoperable channel (s) to OPERABLE status within 7 days, or be in HOT STANDBY within the next 12 hours.
i                b. With one or more remote shutdown system disconnect switches or power
* j                    or control circuits inoperable, restore the inoperable switch (s)/
j                    circuit (s) to OPERABLE status or issue procedure changes per Speci-g fication 6.8.3 that identifies alternate disconnect methods or power or control circuits for remote shutdown within 7 days, or be in HOT STANDBY within the next 12 hours.
: c. The provisions of Specification 3.0.4 are not applicable.
l
            \
            $URVEILLANCE REQUIREMENTS
            /
l 4.3.3.5 The Remote Shutdown System shall be demonstrated operable:
By performance of the CHANNEL CHECK and CHANNEL CALIBRATION I        a., operations    at the frequencies shown in Table 4.3-6 for each remote shutdown monitoring instrumentation channel,                                      j
: b.                                                                                    i By operation of each remote shutdcwn system disconnect switch and                '
k            power and control circuit including the actuated  components at least
              'N        once per 18 months.                              '
                                                                                                    /
~                        ~                                                                /
l PALO VERDE - UNIT 1                      3/4 3-48 l
CONTROLLED BY USER
 
                                                                                                .am.
b
                                                                                                                          ~ . .
                                                                                          . _ _ . _ _                            x
                                                                                                                                    %~~...
                          ,                                                        TABLE 3.3-9                                              N
                          $                        REMOTE SilUTDOWN INSTRUMENTATION, DISCONNECT SWITCilES AND CONTROL CIRCUITS
                        ;g
                                                                                                                                                    \
MINIMUM is m
                                                                        ,                            READOUT                      CHANNELS          \-
INSTRUMENTATION                                                        LOCATION                    OPERABLE                1
: 1. Log Neutron Power Level                                            Remote Shutdown Panel        2
                        $    2. Reactor Coolant flot Leg Temperature                                                                                    \
Remote Shutdown Panel        1/ loop
                        -4    3. Reactor Coolant Cold Leg Temperature O                    9    4. Pressurizer Pressure Remote Shutdown Panel Remote Shutdown Panel 1/ loop 1                              O O                          5. Pressurizer Level Remote Shutdown Panel
: 6. Steam Generator Pressure                                          Remote Shutdown Panel 2
2/ steam generator              Q Z
q 7.
8.
Steam Generator Level Refueling Water Tank Level Remote Shutdown Panel Remote Shutdown Panel 2/ steam generator 2                              2
: 9. Charging Line Pressure                                                                                                        q N                          10. Charging Line Flow Remote Shutdown Panel Remote Shutdown Panel 1
1                          l i
N O                    m
: 11. Shutdown Cooling fleat Exchanger Temperatures
: 12. Shutdown Cooling Flow Remote Shutdown Panel Remote Shutdown Panel 2
2
                                                                                                                                                              /
                        )    13. Auxiliary Feedwater Flow Rate                                      Remote Shutdown Panel                                I 2/ steam generator /
m                    v                                                                                                                                /
m O                    a                                                                                                                              '
O tn                                                                                                                                                  '
tn
                                                                                                                                                                  =<
C                                                                                                                                  -                          C m                                \                                                                                              ~
: g.                      m m                                      ..
                                                  '                        c                                        -          .                    .
m 50                                                      _.._          ;. _ .
                                                                                                              ~~ ~
                                                                                                                                                                  ;10 1
1
                                                                                                                                                                          +
 
CONTROLLED BY USER                            @                                              l''
f-                      ____~~~.._              _
y DISCONNECT SWITCHES                                      LOCATION                                              '
I 1.
SG 1 line 2 Atmospheric Oump Valve SGB-HY-178A RSP          \                                          l i
: 2. SG 2 line 1 Atmospheric. Dump                    RSP              i                                        '
Valve SGB-HY-185A                                                    .
                                                                                                                                  ~
: 3. Auxiliary Spray Valve CHB-HV-203              -
RSP
                                                                                                \
                                                                                                  'g
: 4. Letdown to Regenerative                          RSP                    \
Heat Exchanger Isolation, CHB-UV-515 kg              5. Reactor Coolant Pump                                RSP
                                                                                                    'g Controlled Bleedoff, CHB-HV-505                                            \
: 6. Auxiliary Feedwater Pump                            RSP                        \
B to SG 1 Control Valve AFB-HV-30                                            1
: 7. Auxiliary Feedwater Pump                            RSP                          '
              ,          B to SG 2 Control Valve, AFB-HV-31                                              ,
i
            .      8. Auxiliary Feedwater Pump                            RSP B to SG 1 Block. Valve, AFB-HV-34                                                  \
: 9. Auxiliary Feedwater Pump                                                              \
RSP                                i B to SG 2 Block Valve, AFB-HV-35
: 10. Pressurizer Backup Heaters Banks                    RSP i
i BIO, B18, A05 Control
: 11. Safety Injection Tank 2A                                                                  \
RSP                                      t.
Vent Control SIB-HV-613                                                                    \
: 12. Safety Injection Tank 2B                          RSP Vent Control SIB-HV-623                                                                      \
: 13. Safety Injection Tank 1A                          RSP 1
Vent Control SIB-HV-633                                                                        '\.
: 14. Safety Injection Tank 18                            RSP Vent Control SIB-HV-643
  .              15. Safety Injection Tank Vent                          RSP Q1ves Power Supply SIB-HS-18A
          .        16. SG 1 line 2 ADV                                    .RSP l              SGB-Hi-1788 i                  17. SG 2 lir.= 1 ADV                                    RSP l                        SGB-HY-1853                                                                                          1
                                                                                                                              )
: 18. Control BLDG Battery Room 0                        PHB-H3206                                          !'
Essential Exhaust Fan 'HJB-JOIA'
: 19. Control BLDG Battery Room B                        PHB-H3207                                            i Essential Exhaust Fan 'HJB-JOIB'
: 20. Battery Charger 0 Control                          PHB-3209 AND PKD-H14 Room Circuits                                                                                      ,
: 21. ESF Switchgear Room                              PHB-3205 Essential AHU HJB-ZO3                                                                            '
: 22. LPSI Pump SIB-Pol Breaker                          PBB-504F Control
: 23. Diesel Generator B Breaker                          PBB-5048                            -
Control
: 24. Essential Spray Pond Pump SIB-P01                  PBB-504C          ,./
                                                                                                    /
Breaker Control                                                  ''
PALO VERDE - UNIT 1                  3/4 3-50 CONTROLLED BY USER
 
CONTROLLED BY USER                                                      6'
                                                                                                        /
yy.
        '                                                                                      SWITCH DISCONNECT SWITCHES                                                          LOCATION 25.
Essential Chiller ECB-P01                                              PBB-504G Breaker Control
: 26. E-PBB-504J-4.16KV Feeder                                                PBB-504J                              -
l Breaker to Load Center L-32                                                                                      (
: 27. E-PBB-SO4H-4.16KV Feeder                                                PBB-504H                                  ;
Breaker to Load Center L-34                                                                                      ,
: 28. E-PBB-504N-4.16KV Feeder
  /(g                Breaker to Load Center L-36 PBB-504N      -
: 29. Auxiliary Feedwater Pump AFB P01                                        PBB-5045                ,
l Breaker Control
;              30. Essential Cooling Water                                                PBB-504M                                j Pump EWB-P01 Breaker Control                                                                                  .
: 31. E-PGB-L32B2-480V Main                                                  PGB-L3282                                !
Feeder Breaker to Load Center L-32                                                                  i          j
: 32. E-PGB-L3482-480V Main                                                  PGB-L3482                      i        '
Feeder Breaker to Load Center L-34                                                                    '
: 33. E-PGB-L36B2-480V Main                                                  PGB-L36B2 Feeder Breaker to Load Center L-36
: 34. Charging Pump No. 2 CHB-P01                                            PGB-L32C4 Supply Breaker CHB-P01
: 35. Diesel Engine Control                                                  DGB-C01 Switch 2A                                                                                                .
: 36. Diesel Engine Control                                                  DGB-C01 Switch 2B
: 37. Diesel Generator Control                                                DGB-C01 Switch                                                                      .                          t f
: 38. Diesel Generator Essential                                              DGB-C01                        j Exhaust Fan HPB-J01                                            '
                                                                                                                          +
: 39. Diesel Generator Fuel Oil                                              DGB-C01                      .
Transfer Pump DFB-P01                                                                              1
: 40. Battery Charger 83              .                                      PHB-H3425                  ,
Control Room Circuits-PKB-H16
: 41. Battery Charger B                                                      PHB-M3627                ,
Control Room Circuits PKB-H12
: 42. 125 VDC Battery B Breaker                                              PKB-M4201 Control Room Circuits                                                                          ,
: 43. 125 VDC Battery 0 Breaker                                                PKD-M4401 Control Room Circuits
: 44. CS Pump B Discharge to                                                  PHB-M3804 i
SD HX B SI-HV-689                                                                          .
{            45. Shutdown Cooling LPSI Suction                                            PHB-M3605          .,
SIB-HV-656
: 46. LPSI-CS- from 50 HX B                                                    PHB-M3810 X-Tie SIB-HV-695
: 47. Shutdown Cooling Warmup                                                  PHB-M3806 l                  Bypass SIB-HV-690 l            48. LPSI-CS- to 50 HS B                                                      PHB-M3414 Crosstie SIB-HV-694 l            49. 50 HX "B" to Rc Loops                                                  .
PHB-M3415
!                  2A/2B SIB-HV-696 l
                                                                      ^
                                            ,,,._    s e=--      *-
PALO VERDE - UNIT 1                        3/4 3-51 CONTROLLED BY USER
 
l CONTROLLED BY USER
                                                                                  ~
f~'~~~..                ~~'"  -
                                                                      .-.-                    Y                            I SWITCH l  DISCONNECT SWITCHES                                      LOCATION s
i
: 50. LPSI-50 HX "B" Bypass                              PHB-M3803
                  ,        SI-HV-307                                                                                    -
l  51. LPSI Pump "B" Recire                                  PHB-M3609
                ,        SI-UV-668 l    52. LPSI Pump "B" Suction                                PHB-M3805 b          i
()                        from RWT SI-HV-692
            /      53. 50 Cooling LPSI Pump "B"                            PHB-M3604 i            Suction SI-UV-652                                                                      ,!
l        54. 50 Cooling LPSI Pump "B"                            PKO-844 Suction SI-UV-654 I
: 55. LPSI Header "B" to RC Loop                          PHB-M3606 2A SI-UV-615 1
: 56. LPSI Header "B" to RC Loop                          PHB-M3621 28 SI-UV-625
: 57. VCT Outlet Isolation                                NHN-M7208 CHN-HB-501
: 58. RWT Gravity Feed                                    NHN-M7209                            ;
CHN-HV-536
: 59. Shutdown Cooling Temperature                        PHB-M3412 Control SIB-HV-658                                                                  'i
: 60. Shutdown Cooling Heat Exchanger                      PHB-M3413 Bypass Valve SIB-HV-693                                                            '
: 61. 4.16 KV Bus PBB-504                                  PBB-504K Feeder from XFMR NBN-XO4                                                        i
: 62. 4.16 KV Bus PBB-504                                  PBB-504L Feeder from XFMR NBN-XO3                                                      /
: 63. Electrical Penetration Room B                        PHB-M3631 ACU HAB-206
: 64. Control Room HVAC Isolation Dampers                                          '
HJB-M01/HJB-H55
: 65. 0.S. A. Supply Damper HJB-M02                                          '
: 66. 0.S.A. Supply Damper HJB-M03
: 67. R.C.S. Sample Isolation Valve SSA-UV-203                            '
: 68. R.C.S. Sample Isolation Valve SSB-UV-200                          '
i                                                                                      ,
PALO VERDE - UNIT 1                          3/4 3-52 CONTROLLED BY USER
 
CONTROLLED BY USER
                                                                              ~
                                        - -                  ._-        =n      4 SWITCH    .
: CONTROL CIRCUITS                                                #
LOCATION
: 1. Auxiliary Feedwater Pump B to S/G 1            RSP Isolation Valve AFB-UV34
: 2. Auxiliary Feedwater Pump B to S/B 1            RSP i
Control Valve AFB-HV30                                              -
: 3. Auxiliary Feedwater Pump B to S/G 2            RSP                  !
Isolation Valve AFB-UV35                                              +
    .          4. Auxiliary Feedwater Pump B to S/G 2            RSP                    ;
    /c              Control Valve AFB-HV31 (f,        5. Auxiliary Feedwater Pump                        PBB-SO4S i
AFB-P01
: 6. Charging Pump No. 2                            PGB-L32C4 CHB-Pol                                                      *
: 7. Pressurizer Auxiliary Spray                    RSP Valve CHB-HV203
: 8. Pressurizer Backup Heater Bank                  RSP
: 9. Letdown to Regen HX Isolation                  RSP Valve CHB-UVS15
: 10. RCP Cont Bleedoff                              RSP l                  Valve CHB-UV505
: 11. Volume Control Tank Outlet                        NHN-M7208 Isolation Valve CHN-UV501
: 12. RWT Gravity Feed Isolation                        NHN-M7209 Valve CHE-HV536
: 13. S/G 1 line 2 Atmospheric Dump Valve Control      RSP SGB-HIC 1788
: 14. S/G line 2 Atmospheric Dump                    RSP Valve SGB-HY-178A                            -
: 15. S/G 1 line 2 Atmospheric Dump                    RSP Valve SGB-HY-1788                                                  '
: 16. S/G 2 line Atmospheric Dump                      RSP Valve Control SGB-HIC-1858
: 17. S/G 2 line 1 Atmospheric Dump                    RSP l                  Valve SGB-HY-185A l            18. S/G 2 line Atmospheric Dump                      RSP 1
Valve SGB-HY-1858
: 19. Diesel Generator B Output                        PSB-5048 Breaker
: 20. Diesel Generator Building                          DGB-801 Essential Exhaust Fan H08-J01
: 21. Diesel Generator B Fuel Oil                        CGB-B01 Transfer Pump DFB-P01
: 22. 4.16 KV to 480V LC L34                            PSB-504H Supply Breaker
: 23. 4.16KV to 480V LC L32                              PBB-504J Supply Breaker
: 24. 4.16KV to 480V LC L36                              PBB-504N Supply Breaker
: 25. 480V LC L32                                        PGB-L32B2 Supply Breaker
: 26. 480V LC L34                -
PGB-L3482 Supply Breaker l
PALO VERDE - UNIT 1                  3/4 3-53 1
CONTROttED BY USER
 
J u CONTROLLED BY USER
                                                                                                        ~
                                                                                                      ,s;,
                      '/
SWITCH    ',              -
j 'C ONTROL CIRCUITS                                                      LOCATION
[/27. 480V LC L36                                                        PGB-L36B2 i          Supply Breaker
                / 28. Battery Charger PKB-H12 PHB-M3627 j            Supply Breaker                                                                              ~
j        29. Battery Charger PKD-H14 PHB-M3209 j              Supply Breaker l      *30. Backup Battery Charger PHB-M3425              '
          /                  PKB-H16 Supply Breaker
  'g5 I                                                                                                            !
: 31. Essential Spray Pond Pump                                        PBB-SO4C I    i                  SPB-Pol                                                                                !
: 32. Essential Cooling Water Pump                                  PBB-504M EWB-Pol j              33. Essential Chilled Water                                -
PBB-504G i
Chiller ECB-E01                                                                        .
{'          34. Battery Room D Essential PHB-M3206 Exhaust Fan HJB-J01A                                                                  ,
: 35. Battery Room B Essential                                        PHB-M3207 i
Exhaust Fan HJB-J018
: 36. ESF Switchgear Room B                                                                '
PHB-M3203 Essential AHU HJB-ZO3                                                              !
: 37. Electrical Penetration Room B                                    PHB-M3631 ACU Fan HAB-ZO6
* 38. SIT Vent Valves Power                                            RSP Supply SIB-HS-18A                                                                ,
: 39. SIT 2A Vent Valve                                                RSP
          ;                SIB-HV613
: 40. SIT 28 Vent Valve                                                                ,
RSP                ,
          !                SIB-HV623        .
: 41. SIT 1A Vent Valve                                                RSP l
SIB-HV633                                                                        '
: 42. SIT 18 Vent Valve                              .
                                                                                        . RSP                i SIB-HV643                                                                          i
: 43. LPSI Pump B PBB-504F S18 Pol
            ;      44. Containment Spray Purap B                                                                      '
PHB-M3804 Discharger to 50 HX "B" l                          Valve SIB-HV689
: 45. LPSI Containment Spray from                                  PHB-M3B10 l                    SD HX "B" X-tie Valve SIB-HV695 l            46. Shutdown Cooling LPSI Suction                                    PHB-M3605        -
Valve SIB-HV656                                                  -
: 47. Shutdown Cooling Warmup Bypass                                  PHB-M3806 Valve SIB-UV690
: 48. LPSI Containment Spray to
                                                                                    - - PHB-M3414 SD HX "B" X-tie Valve SIB-HV694
                                                                                                        ,/
r
_m=*
PALO VERDE - UNIT 1                    3/4 3-54                    .
CONTROLL'ED BY USER                                                                    '
i
 
                  ./
p g                                    -
N.L Q                        I SWITCH
            /      CONTROL CIRCUITS                                    LOCATION i            49. 50 HX "B" to RC Loops                        PHB-M3415 3                  2A/28 Valve SIB-HV696
: 50. LPSI SD HX "B" Bypass                        PHB-M3803 i                  Valve SIB-HV307                                                                    -
i          51. .LPSI Pump B Recire.                              PHB-M3609 1                Valve SIB-UV688                                                    1
            !      52. LPSI Pump B Suction                          PHB-M3805 From RWT SIB-HV692                                            ~
j      53. RC Loop to Shutdown f    i              Cooling Valve SIB-UV652 PH8-M3604                !
i      54. RC Loop to Shutdown                          PKO-B44                    i Cooling Valve SIB-UV654                                                  1
: 55. LPSI Header 8 to RC                          PHB-M3606 j              Loop 2A Valve SIB-UV615                                                    {
i
:        56. LPSI Header 8 to RC i                            PHB-M3621                    I i                Loop 28 Valve SIB-W625                                                      l
          !'        57. SDC "B" Temperatura Control Valve                PHB-M3412                      !          !
SIB-HV-658                                                                    i i
: 58. Control Room Ventilation Isolation                                                ;
Dampers HJB-M01/HJB-M55                                                        :
            .      59. 0.S.A. Supply Damper HJB-M02                                                  '
i      60. O.S. A. Supply Damper HJB-M03                                                  -        '
            } 61. Diesel Generator "B" Emergency Start                                                  f 1 62. Normal Offsite Power Supply Breaker                P88-504K                      1 1 63. . Alternate Offsite Power Supply Breaker                PB8-504L                    /
i 64. Battery "B" Breaker                                PKB-M4201                  /
: 65. Battery "0" Breaker                              PKO-M4401              ;
: 66. RCS Sample Isolation Valve SSA-UV-203                              '
{  67. RCS Sample Isolation Valve S$8-UV-200        '
                                                                                          /
: 68. Shutdown Cooling Temp. Control
                                                  '                    .PH8-M3412                                    '
s        Valve SIB-HV-658                                          -
: 69. Shutdown Cooling Heat Exchanger Bypass          PH8-M3413
                    \
Valve. SIB-HV-693
                          \                                                                                      :
                              'x b
e PALO VERDE - UNIT 1                    3/4 3-55
 
e                          n
                                                                                                            , _ . - -_                                                        f)
                                                                                                                                            ~~
TABLE 4.3-6                    '
                                                                                                                                                                  ; ,_ x g
                                                  !2 O
m REMOTE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS                                          N.\    'N
                                                    *                                                                                                                                        \
                                                                                                                                                                                              \
c                                                                                              CHANNEL                CHANNEL
                                                  -s INSTRUMENT                                                                              CHECK          CALIBRATION            \
r          1.            Log Neutron Power Level                                                                                            \
M                R                      \
: 2.            Reactor Coolant Hot Leg Temperature (2)                                    M                R
: 3.            Reactor Coolant Cold Leg Temperature (2)                                    M                R                          i
: 4.            Pressurizer Pressure                                                                                                    \i M                R S.            Pressurizer Level                                                          M              .R
: 6.            Steam Generator Pressure                                                    M                R 8
t 5:          7.            Steam Generator Level                                                      M                R
: 8.            Refueling Water Tank Level                                                  M                R
                                                  $          9.            Charging Line Pressure                                                      M                R                            '
i
: 10. Charging.Line Flow                                                                    M                R                            i      ,
: 11. Shutdown Cooling Heat Exchanger Temperatures                                          M                R l
: 12. Shutdown Cooling Flow                                                                  M                R                        j
                                                ,            13. Auxiliary feedwater Flow Rate                                                          M                R                      /            "
l 1
                                                                                                                                                                                              /
e N                                                                    ,                                                          /
I                                                          ''N                                                                                                                            .
l                                                                                .                                                                                                  -
                                                                                                                                        *"^*'e  . , , .    .,r- w* #'
                                                                                                                                                                            ,/              f..x
                                                                                                  .m.          -,mm,e....
 
1 g'
                                    ,_                        - ~ ~ ~ . . .                                '..        .
                              .                2Jg, 9                              -.          .
REMOTE SHUTDOWN MOSI-TORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION
        )
3.3.3.5 The remote shutdown controls and monitoring instrumentation channels                                  i
      !                  shown in Table 3.3-6 shall be OPERABLE with readouts displayed external to the control room.                                                                                                    )          ,
l 1                APPLICABILITY: '10 DES 1, 2, and 3.                                                                            .
                                                                                                                                            ~
l                        ACTION:                                                                                                      i 5                      a. With the number of OPERABLE remote shutdown monitoring channels less than required by Table 3.3-6, either restore the inoperable channel to OPERABLE status within 7 days or be in HOT SHUTDOWN within the next 12 hours.
: b. With one or more remote shutdown system instrumentation control circuits required by Table 3.3-6 inoperable, restore the inoperable
            ,                        circuit to OPERABLE status within 7 days, or be in at least HOT STANDBY within the next 12 hours.                                                            ,
l              ;                c. The provisions of Specification 3.0.4 are not applicable.
              \
                \                                                                                                                ,
l                                                                                                              '
SURVEILLANCE REQUIREMENTS                                                                              -
                  \
                    \ 4.3.3.5.1 Each remote shutdown monitoring instrumentation' channel shall be i demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRA-l (TIONoperationsatthefrequenciesshowninTable4.3-3.                                                  /
                                                                                                                        /
4.3.3.5.2 Each remote shutdown system instrumentation control circuit shall be demonstrated OPERABLE by verifying its capability to perform its intended function (s                once per 18 months.                                        .
l l
l l
l l
lI
;                                                                  ,                            Amendment Number 9 February 27, 1984
'l 1
 
                                                                                                                                    ~--
w TABLE 3.3-6                                                              's REMOTE SHUTDOWN MONITORING INSTRUMENTATION.                                                                      x
                                                                                                                                                                              \
                                                                                                                                                                                \
Readout            Measurement                            Minimum Channels Instrument                                                                    Location                Range                                        OPERABLE
: 1. Logarithmic Neutron Channel                                                                          RSP*          2 x 10 200%                                            \  1
                                                                                                                                                                                              \
: 2. Reactor Coolant Hot Leg Temperature                                                                    RSP
                                                                                                                          ~
50 - 750*F                                            1/ loop
      , 3. Pressurizer Pressure                                                                                  RSP          15 - 3000 psia                                                1 j 4. Pressurizer Level                                                                                      RSP.              0 - 100%
1
                                                                                                                                                                                              )1
    ' 5. Steam Generator Pressure                                                                              RSP            0 - 1385 psig-                          1/ steam generator
                                                                                                                                                                                            /
R        6. Steam Generator Water Level                                                                            RSP                0 - 100%                            1/ste'am generator e
Y    \ 7. Refueling Water Tank Level                                                                            RSP                0 - 100%                                  J              l a                                                                                                                                                                              f
: 8. Charging Flow                                                                                          RSP              0 - 150 gpm                            /                    1
: 9. Charging Pressure                                                                                      RSP            0 - 3000 psig                      7                          1
                                                                                                                                                                ./ r s'              ,}
Control Circuits                                                ,                                                                      ,/                --
Pressurizer Heater f'
n 5[ g*    .RSP - Remote Shutdown Panel c a.
ghs :3                                                            - - . -
xa                                                                        .
  -E GRT 2 .'
 
i 6
([,
TABLE 4.3-3                                                  !
REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE RE0VIREMENTS
    '/                  INSTRUMENT CHANNEL                CHANNEL CHECK                CALIBRATION
,1.      Logarithmic Neutron Channel                          M                    hR              '
: 2.      Reactor Coolant Hot Leg Temperature                  M                    R
: 3.      Pressurizer Pressure                                  M                    R
: 4.      Pressurizer Level                                    M                    R
: 5.      Steam Generator Pressure                              M                    R
: 6.      Steam Generator Water Level                          M                    R
: 7.    ' Refueling Water Tank Level                            M                    R'
: 8.      Charging Flow                                        M                    R
: 9. Charging ~ Pressure                                    M                    R
                        %                                                        s'
* N s
                                                                  >-f Amendment Number 9 February 27, 1984 3/4 3-42
 
                                        ,                                                                                                          !I M    tINSTRUMENTATIONS -)                                                                                                        l'
  ,                    POST-ACCIDENT MONITORING INSTRUMENTATION                                                                                  i LIMITING CONDITION FOR OPERATION I
3.3.3.6 The post-accideni; monitoring instrumentation channels shown in Table 3.3-Q sfia1The OPIRABLE.
7 APPLICABILITY: MODES 1,.2, and 3.
                                                                                                                        ~
ACTION:
7 g                            i
: a. With one or more accident monitoring instrumentation channels                                                l inoperable, take the action shown in Table 3.3-10.                                s f,. #
k        \
: b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS (6) 4.3.3.6 Each post-accident; monitoring instrumentation channel shall be i                      demonstrated OPERA 8Lt ny performance of the CHANNEL CHECK and CHANNEL i
[ CALIBRATION operations at the frequencies shown in Table 4.3-y-f, e
i e                                /"
b p
4 P
PALO VERDE - UNIT 1                  3/4 3-57 j-      .
I
                                          -...-._m . .- .  ,r r.,. r__      .. ,  __.-_., , ,x  _ _ -% ._ .
_#_-_s _.,    ,-...--.-
 
                                                                        -.  .              _ _ _ .          . - . _    .  .=
i i
i POST-ACCIDENT MONITORING INSTRUMENTATION                                                      l LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-7                    l shall be OPERABLE.                                                                              l APPLICA8ILITY: MODES 1, 2 and 3.
ACTION:      p--
: a. With the number of OPERABLE accident monitoring channels less than '
                                                                                                                                  \
the Required Number of Channels shown in Table 3.3-7, either restore \
the inoperable channel to OPERABLE status within 7 days, or be in HOT SHUTDOWN within the next 12 hours.                                          .)
: b. With the number of OPERABLE accident monitoring channels less than                  l the Minimum Channels OPERABLE requirements of. Table 3.3-7; either              /  i restore the inoperable channel (s) to OPERABLE status within 48 hours          /
or be in at least HOT SHUTDOWN within the next 12 hours.                      /      i
: c. The provisions of Specification 3.0.4 are not applicable.                  /
x
                                          ~ ~~-                                                  ;;= ' f '/
SURVEILLANCEREQUIREMENTE 4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrat-
                                                                                                  ~
ed OPERABLE by performance of the CHANNEL CHECK arid CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-4.
l l
I Amendment Number 9 February 27, 1984 3/4 3-43 i
    . , _ _ , _ _ _ _ , _ , . .,                . , , , _ . ~ , - -
 
G%                                                                        ~'
TA8LE 3.3- M ActsotSt o                                                                              POST-MONITORING    INSTRUMENTATION p                                                                                      -
5
                                                              ;5i
                                                                                                                                      '                                    REQUIRE 0        HINIMUM NUMBER OF        CHANNELS g h INSTRUMENT (I(bu/rafiM                                                  )                              CHANNELS          OPERA 8LE    ACTION.
: 1.            Containment Pressure                                          2              1          , 29,30                    b            ,
b
                                                              ^
: 2.            Reactor Coolant Outlet Temperature - Thot (Wide Range)        2              1/ loop      29,30                      ff  ' l 'U
                                                              "                                3.                                                                                                                                  S.00 Reactor Coolant Inlet Temperature - Tcold (Wide Range)        2              1/ loop
                                                                @ 6A.                                "d5M'd'{5A n",'M*"ge                                                      2              1            29,30
: 5.            Pressurizer Water Level                                        2              1            29,30
: 76.              Steam Generator Pressure                                      2/ steam      1/ steam    29,30 generator      generator 37.
Steam Generator Water Level - Wide Range                      2/ steam      1/ steam    29,30 generator      generator.
[ 't h,                                          Refueling Water Storage Tank Water Level                      2              1            29,30 g so 9.                                        Auxiliary Feedwater Flow Rate                                    2              1            29,30
:: 10. Reactor Cooling System Subcooling Margin Monitor                            2              1            29,30
                                                                  **l211. <Piessufjfte Safety Valve Position Indicator                                                        1/ valve      1/ valve    29,30 33 12              Containment Water Level (Narrow Range)                          2              1            29,30    .
11 13. Containment Water Level (Wide Range)                                          2              1            29,30 is l'4. Core Exit Thermocouples                                                        4/ core        2/ core      29,30 quadrant      quadrant
                                                                                          > 15.                Reactor Vessel Water Level                                ,
2*            1a          31,32 17 l'6.            Neutron Flux Monitor (Power Range)                              2              1            29,30 J
                                                                                                *A channel is eight sensors in a probe. A channel is OPERABLE if four or more sensors, two or more in the upper four and two or more in the lower four, are OPERABLE.
 
                                                          @              TABLE 4.3-@ N g                                POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
N                                                                                      CHANNEL    CHANNEL
%    INSTRUMENT (2fustraticta! OnCty)                                                      CHECK CALIBRATION
[      1. Containment Pressure                                                            M        R h    2. Reactor Coolant Outlet Temperature - That (WId* "*"9*)                          M        R e      3. Reactor Coolant Inlet Temperature -T cold IwId* I*"8*)                H        N
: 4. Pressurizer Pressure - Wide Range                                                  M        R cy                                                                              t        N
  > s, hl,fr.u. essurizer rte ectortWater sy., tem Pres 2nra Level                                                    M        R
      '6.7  Steam Generator Pressure                                                        M        R 7.5 Steam Generator Water' Level - Wide Range                                        M        R
      '8.') Refueling Water Storage Tank Water Level                                        M        R
{    9.c Auxiliary Feedwater Flow Rate                                                    M        R y    10): Reactor Coolant System Subcooling Margin Monitor                                  M        R E cellJ2 d'FiissurliEn Safety Valve Position Indicat~or M        R 12.0 Containment Water Level (Narrow Range)                                            M        R 13.11 Containment Water Level (Wide Range)                                            M        R 143s' Core Exit Thermocouples                                                          M        R ISA Reactor Vessel Water Level                                                        M        R 16.t? Neutron Flux Monitor (Power Range)                                              M        R t
4
                                                    . ..            =                ,,.
 
f-'
[                                    TABLE x
ACTION STATEMENTS
        /
      / ACTION 29 -            With the number of OPERA 8LE Channels one less than the Recuired
[                        Number of Channels in Table 3.3-10, either restore the Inoperable Channel (s) to OPERA 8LE status within 7 days, or be in HOT
    /.                      ' SHUTDOWN within the next 12 hours.
                                                                                                            \
ACTION 30 -
With the number of OPERA 8LE Channels one less than the Minimus                k Channels OPERABLE in Table 3.3-10, either restore the Inoperable f                            Channel (s) to OPERA 8LE status within 48 hours or be in at least i
HOT SHUTDOWN within the next 12 hours.
ACTION 31 -        With the number of OPERA 8LE Channels one less than the Required Number of Channels, either restore the system to OPERA 8LE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-
    \                        lining the action taken, the cause of the inoperability and the                      '
    \                        plans and schedule for restoring the system to OPERA 8LE status.
      \ ACTION 32-            With the number of OPERA 8LE Channels one less than the Minimum
      \                      Channels OPERABLE in Table 3.3-10, either restore the inoperable                '
        \
channel (s) to OPERABLE status within 48 hours if repairs are
        \ \
feasible without shutting down or:
            \                1. Initiate an alternate method of monitoring the reactor N                    vessel inventory;
                \
s            2. Prepare and submit a Special Report to the Commission pur-suant to Specification 6.9.2 within 30 days following the
                  \s              event outlining the action taken, the cause of the inopera-bility and the plans and schedule for restoring the system to OPERABLE status; and
                      \
x    3. Restore the system to OPERA 8LE status.at the next scheduled
                          '\        refueling.                                                                  .
N N,                                                                      /
                                                                                                    /
                                      \                                                        .,
pW
                                                                                        ~
6 PALO VERDE - UNIT 1                      3/4 3-60
 
M.
                                ,r
                          ,INSTRUMENTATIONy K                                                            C " Q,, 4
                                                                                                                            '~
                          , FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION A O SR O M tL @ v Cth o W M9FS                              -
h                -~~
3/43.3.I Fia CNedionb@ammMA P          3.3.3.7 As a minimum, the fire detection instrumentation for each FPER I          detection zone shown in Table 3.3-11 shall be OPERA 8LE.
i
                -1          APPLICA8ILITY: Whenever equipment protected by the fire detection instrument j        is required to be OPERA 8LE.
ACTION:
                    >-            a. With any, but not more than one-half the-total in any fire zone
                    !'                  Function X fire detection instrument shown in Table 3.3-11 inoperable, i                '!
restore the inoperable instrument (s) to OPERA 8LE status within 14 days or within the next I hour estabitsh a fire watch patrol to e
                    ;                    inspect the zone (s) with the inoperable instrument (s) at least once l                    ;
per hour, unless the instrument (s) is located inside the containmenti then inspect that containment zone at least once per 8 hours or i
monitor the containment air temperature at least once per hour at i                  the locations listed in Specification 4.6.1.5.
b.
With more than one-half of the Function X fire detection instruments in any fire zone shown in Table 3.3-11 inoperable, or with any
                  /                    Function Y fire detection instruments shown in Table 3.3-11 inoperable.
[                      or with any two or more adjacent fire detection instruments shown in
                #.                      Table 3.3-11 inoperable, within 1 hour establish a fire watch patrol I-                        to inspect the zone (s) with the inoperable instrument (s) at least j              l                        once per hour, unless the instrument (s) is located inside the l
l containment, then. inspect that containment zone at least once per
,                                        8 hours or monitor the containment air temperature at least once per hour at the locations listed in Specification 4.6.1.5.            '
: c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS
      ]
4.3.3.7.1 Each of the above required fire detection instruments which are
!        !              accessible during plant operation shall be demonstrated OPERA 8LE at least once l        l              per 6 months by performance of a CHANNEL FyNCTIONAL TEST. Fire detectors
    .j                  which are not accessible during plant operation shall be demonstrated OPERA 8LE by the performance of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN                    *
                        -exceeding 24 hours unless performed in the previous 6 months.
l 4.3.3.7.2 The NFPA Standard 720 supervised circuits supervision associated s'            with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.
                          %-                ~
PALO VERDE - UNIT 1              _
3/4 3-61                                      .
                                                            ~..                                            /
                                                                            'p-l i
 
                                                                                    --                      N      s 7
7 TABLE 3.3-11                  N r
j                                                              FIRE DETECTION INSTRUMENTS                                            'N      .
I i              FIRE        ELEVATION                    INSTRUMENT LOCATION                      TOTAL NUMBER OF INSTRUMENTS
* ZONE f                                                                                                    HEAT                  FLAME            SM0KE                      ',
L l                                                                                                      (x/y)                                  (x/y)
I                                                                            BUILDING - CONTROL i                                                                                                                                                                            !, .
1                  74'              Essential Chiller Rs. -                                                            24/0 i
f                                                                      Train A                                                                                                    !
l
          .j                      2                  74'              Essential Chiller Rs. -                                                            21/0                          ;
i Train 8 SA                74'              Cable Shaft - Train A                                                                1/0
              \'
36                74'              Cable Shaft - Train 8                                                                1/0
                  ;            86A          74-156'4"                Deadspace Compartment -                            0/1                                0/3 i
Train A                                                  .
868          74-156'4"                Deadspace Compartment -                            0/1                                0/3                        /
Train 8 4A              100'                Cable Shaft - Train A                                                                1/0                      l
}'
    ,                  ;      48              100'                Cable Shaft - Train 8                                                                  1/0                    !
                          \
1 SA              100'                ESF Switchgear Room -              .              0/5                                0/5                    l i'
Train A                                                                                              I i
* 58              100'                ESF Switchgear Room -                                                                                        I 0/5                                0/5
                              !                                              Train 8                                                                                            !'
i t                                                                                                                                                I j 6A 100'                DC Equip. Rs. - Train A                                                              2/0                ;
(Channel C) i
                          /68                    100'              DC Equip. Rs. - Train 8                                                                2/0              ,
i                                                  (Channel D)                                            _.                                      l t                                7A              100'              DC Equip. Rs. - Train A                                                                2/0              l (Channel A)                                                                                    l
!                        ' 78                  100'                DC Equip. Rs. - Train 8                                                                2/0
                        !                                                    (Channel B)                  .
                      ,i                                                                                                                                                .
                      / 8A                    100'                Battery Am. - Train A                              0/2                                0/2
;                    I                                                      (Channel C) i
                  /            88              100'              Battery Rm. - Train 8                              0/2                                0/2
      .'          \                                                        (Channel 0)
PALO VERDE - UNIT 1                                          3/4 3-62 l
 
                          /p_                                                                      %
TA8LE 3.3-11 (Continued)                      ~~ s                          ;
FIRE DETECTION INSTRUMENTS
        .I                                                                                                                                i FIRE  ELEVATION        INSTRUMENT LOCATION-                  TOTAL NUMBER OF INSTRUMENTS
* ZONE HEAT      FLAME          SMOKE i                                                                                                                            .
(x/y)      (x/y)          (x/y)
          ,i 9A      100'-        Battery Rs. - Train A                      0/2                    0/2 h;                                              (Channel A) f                    98    '100'          . Battery Rs. - Train 8                    0/2                    0/2            I (Channel B)                          .                                        -
10A      100'          Remote Shutdown Rs. -                      0/1                    1/1
:                                                      Train A                                                                      l i
108      100'          Remote Shutdown Rs. -                    0/1                      1/1 I  -
Train 8                                                                    i i
11A      120'          Cable Shaft - Train A                                              1/0      l 118      120'          Cable Shaft - Train 8                                              1/0 14      120'          Lower Cable Spreading Rs.                                      ~
l              ,                                              System 1                      0/1                    0/6
!                                                              System 2                      0/1                    0/6 Systeni 3
* 0/1                    0/8        '
j                                            System 4                      0/1                    0/8      i j                                          System 5                      0/1                    0/8 l
i.
System 6                      0/1                    0/8      !
I 15A      140'          Cable Shaft - Train A                                              1/0    I i
                      ! 158        140'          Cable Shaft - Train 8                                              1/0 i
'                    l 17          140'          Control Rs. - MC8's & Relay                4/0                    113/0 j                                Cabinets
                  /
j j        18A      160'          Cable Shaft - Train A                                              1/0 188      160'          Cable Shaft - Train 8                                              1/0
(\          20        160'          Upper Cable Spreading Rs.
r
                \\                                            System 1                      0/1                      0/12 System 2                      0/1                    0/8
                        \                                    System 3                      0/1                    0/8 System 4                      0/1                    0/8 System 5                      0/1                    0/8 l
x ' ~ ~ %"m                        ._., --~~~~~~'~~ . _ _.-- .._  . - .
PALO VERDE - UNIT 1                        3/4 3-63                                                              I
\
 
l i
                            - ~ ~
                                                                  ~ ~ <
            -[                                      TA8LE 3.3-11 (Continued)                      ' s, 7
j FIRE DETECTION INSTRUMENTS l                                                                                                                      .
I f.
FIRE    ELEVATION    INSTRUMENT LOCATION ZONE TOTAL NUMBER OF INSTRUMENTS *                  !I HEAT      FLAME      SMOKE                  f (x/y)      (x/y)      (x/y)                  J''
BUILDING - DIESEL GENERATOR k
21A      100'      Diesel Generator - Train A          0/3    0/4 218        100'      Diesel Generator - Train 8          0/3    0/4 22A        100'      Diesel Generator Control Rs. -                            1/0          i
,                                            Train A l
228        100'      Diesel Generator Control Rs. -                            1/0 4
Train 8                                                              1 g
24A        115'      Combustion Air Intake Rs. -                              1/0 Train A 248        115'      Combustion Air Intake Rs. -                              1/0 Train 8                                                                  ,
23A        131'      Fuel 011 Day Tank -                0/1 Train A                                                                  +
t l
238        131'      Fuel 011 Day Tank -                0/1                                      i t-                                            Train 8 1
25A        131'      Exhaust Silencer Rs. -                      3/0 Train A                                      -
i 258        131'      Exhaust Silencer Am. -                      3/0                                    ,
Train 8                                                                  '
BUILDING - FUEL
* i
          .        28    ,    100'      Spent Fuel Pool Cooling and                              3/0 Cleanup Pump Areas                                                    8 l
                                                                                                                    ^
BUILDING - AUXILIARY 88A    51'-6"      West Corridors                                          6/0 888    51'-6"      East Corridors                                          6/0            l 32A    51'-6"      LPSI Pump Rs. - Train A                                  0/2 328    51'-6"      LPSI Pump Rs. - Train 8                                  0/2 PALO VERDE - UNIT 1                    3/4 3-64
 
                                                                                                                ~
_____.-- -      ^%
TABLE 3.3-11 (Continued)
                                                                                                                  '~
                                                                                                                    's
                  /                                                                                                    s      ,
                ,/. '                                            FIRE DETECTION INSTRUMENTS
                                                                                                                            \
i            FIRE    ELEVATION            INSTRUMENT LOCATION            TOTAL NUM8ER OF INSTRUMENTS
* ZONE                                                                                                    '
HEAT FLAME        SMOKE          '
(X/y)    - - -- (x/y)        (x/y) l r
34A            70'          ECW Pump Rs. - Train A                                          2/0 348            70'          ECW Pump Rs. - Train 8                                          2/0 g
35A            70'          Shutdown Cooling Ht. x Chgr.                                    4/0 Train A                                                                              -
358            70'          Shutdown Cooling Ht. x Chgr.                                    4/0 Train B 36            70'            Reactor Makeup and Boric                                        1/0 Acid Makeup Room i
t 37C          70'& 88' Piping Penetration Rs. -                                              5/0                    f Train A                                                                            I I
37D          70'& 88' Piping Penetration Rs. -                                              4/0                  i Train 8 4
                      , 378            70'            Corridors - East
    \                                                                                                                11/0                I 37A          70'            Corridors - West                                              11/0 39A'        88'            Pipeways - Train A                                                                I
                                                                                                ,                    8/0              ,
398          88'            Pipeways - Train 8
;                                                                                                                    8/0 i
l                        42A        100'              Elect. Penetration Rs. -          0/1                          0/25 l                                                        Tr. A (Chan. C) 42B        100'              Elect. Penetration Rs. -          0/1                          0/24
,                                                        Tr. B (Chan. 8) 42C        100'              Corridors - East & Southeast      0/2                          3/35 4
420        100'              Corridor - West                    0/1                          2/29        . - .          . . -
46A        100'.            Charging Pump and Valve                                        0/3 Gallery Rs. - Trai;4 A                                                i 468        100'              Charging Pump and Valve                                        0/3 Gallery Rs. - Train 8 46E        100'              Charging Pump and Valve                                        0/3 Gallery Am. - Train E
;                      PALO VERDE - UNIT 1                            3/4 3-65                    -
      -----4                -..y  ,_..m.  . - , ,
 
                                                    ,                    6.--e=
                                  -                          TABLE 3.3-11 (Continued)                                  'N-s x
                        ,-                                FIRE DETECTION INSTRUMENTS                                              N
                ./
                /. FIRE        ELEVATION      INSTRUMENT LOCATION                  TOTAL NUMBER OF INSTRUMENTS
* j        ZONE HEAT                              FLAME  SMOKE j                                                                      (x/y)                              (x/y)  (x/y)      ..
bI I            47A        120'        Elect. Penetration Rs. -                          0/1                          0/28        I l                                          Tr. A (Chan. A)                                                                        {
i i
478      ~
120'        Elect. Penetration Rs. -                          0/1                          0/24          i Tr. 8 (Chan. 0) 48        120'        ECW Surge Tanks Corridor -                                                      3/0            i Tr. A & 8 508        120'        Valve Gallery                                                                    1/0              I, 518        120'        Spray Chemical Storage Tk Rs.                                                                      I 1/0              ,
I 52A        120'        Central Corridor - West                          0/1                            5/17
                                                                                                                                                  \  ,
520        120'        Central Corridor - East                          0/1                            7/18                !
54        120'        Reactor Trip Switchgear Rs.                                                      6/0 1/0                  j ,
568        140'        Storage and Elect. Equip.-                                                      6/0                  !
Rs. - East                                                                                    t
                                                                                                                                                    ,  t 57I        140'        Clothing Issue and Men's                                                        5/0                    I Locker Rs.
f  ,
57J        140'        Women's Locker, Clean Storage                                                    7/0                    i and Lunch Res.                                                                                    t 57N        140'        Corridor Area                                                                    4/0                    !
BUILDING - CONTAINMENT **
66A&668 100' & 120 Southwest and Southeast                                1/0 Perimeter 67A&678 ~ 100'          Northwest and Northeast                          1/0 Perimeter                                                                          '
66A        120'        Southwest Perimeter                              1/0 668        120'        Southeast Perimeter                              1/0 67A&678 120'            Northwest and Northeast                          3/0 Perimeter PALO VERDE - UNIT 1                          3/4 3-66 1
 
                                                                . _ .                                      >-                        m.
                                'p-4                                                                -
TA8LE 3.3-11 (Continued)
                                                                                                                                                                's FIRE DETECTION INSTRUMENTS
                ' ,1                                                              +
j              FIRE    ELEVATION              INSTRUMENT LOCATION                                  TOTAL NUMBER OF INSTRUMENTS *
                ;              ONE HEAT                FLAME            SMOKE (x/y)                Tx7yl            (x/y)              *
.''                                                                                                                                                                              \
{            63A        120'                No. 1 RCPs and SG Area                                                                          6/0                  \
                  !          638        120'                No. 2 RCPs and SG Area                                                                          6/0 1
    ^                                                                                                                                        ~
(b        i          66A&668
                            . 67A&678      140'      .      Southwest, Southeast,                                      1/0 Northwest and Northeast                                                -                                              I 3                                                Perimeters
{
f          63A        140' .            .No. 1 RCPs and SG Area s
5/0                            [
f                                      '
i, 638        140!                No. 2 RCPs and SG Area                                                                          5/0
,                                                                                                                                                                                              i 4
70        140' s              Refueling Pool and Canal Area                                                                  4/0
                                                                                                                                                                                              'l.
i' 71A        140'                North Preaccess Normal AFU                                                                      2/0                                6
                                                **                Area                                                                                                                          .-
q                                                                                                                                                  ,
718      '140                  South Preaccess Normal AFU                                                                    2/0                                j
(                                                            Area                                                                                                                        ?
                                    .                                                                                                                                                          I e
MAIN STEAM SUPPORT STRUCTURE                                                                                                    j, i'
72            80'              Turbine Driven Aux. Feedpump                                                                  0/3                              ::
Rs.                                                                                                                              ;
t 73            80'.              Notor Driven Aux. Feedpump Rs.                                                                1/1                            ;        ,
                                                                                                                                                                                          ;          4
                            ' 74A        100',120' Nain Steam Isol. & Dump Valve                                                                            0/6                        I
                                            & 140'                Area North
                                                                                                                                                                                      /
                          + 748        100'',120'          Main Steam Isol. & Dump. Valve                                                                0/6                    -
                                            & 140'                Area South        1 4
OUTSIDE AREAS 83                              Condensate Storage Tank Pump                                                                  2/0        ,
House                              ' t                                                                      -
                                                                                                                                                                    /
h
                                                                                                                                            ~ '
PALO VERDE - UNIT 1                                        3/4 3-67.        .
9 l'
 
t
                            'p_.--                                      ~ ~ ~
                        ,-                              TA8LE3.3-11(Continued) _ _-Tr~ x NNs' ,_
FIRE DETECTION INSTRUMENTS                  '  -
[ FIRE    ELEVATION      INSTRUMENT LOCATION                                                                      '
ZONE                                                  TOTAL NUMBER OF INSTRUMENTS *                                .
HEAT      FLAME        SMOKE l                                                                                                                  '
f (x/y)      (x/y)        Tx7yJ                          ,
i i
            !          84A                  Spray Pond Pump House -
                                                                                                                      ;              ii Train A                                                1/0            ',
f                                                                                                                              '
848                  Spray Pond Pump House -
g s
{l
                                                                                                                                      .c Train 8                                                  1/0                ,        ;
t i
P The fire detection instruments located within the containment are not required to be OPERABLE during the performance of Type A containment leakage rate tests.
                    *(x/y):
x is the number of instruments associated with early fire detection and notification only.
y is the number of instruments associated with actuation of fire
                              , suppression systems and early fire detection and notification.
                                  ~~~_._      __
PALO VERDE    UNIT 1                      3/4 3-68
 
e                    ..
J --
INSTRUMENTATION
                                                                                        ~
1" %fSlitk                hD LOOSE-PART DETECTION INSTRUMENTATION                                .
g              LIMITINGCONDITIONFOROPERAh!ON ti ." 9; F.htu Ae V (JsM,'f:csq st'.rf.                                              '
                                                                        ~m-                                    -
w                        4-
      '@      f pt.3.~3.3.8
                        .s.1.s    eu  inn oose Th~el  _wa part waar~iitim~shal1~b~e~0PERABLE detection s                                  with all sensors j
(                      specified in Table 3.3-12.
j                                                                                                                                    s i
l          APPLICA8ILITY: MODES 1 and 2-I
                                                                                                                                                  \
t          ACTION,                                                                                                                        \          \
Ofb i
: a.                                                                                                                      k t
i                          With one or more loose part detection system channels inoperable for                                              (
          ,                          more than 30 days, prepare and submit a Special Report to the                                                      i      ,
Commission pursuant to Specification 6.9.2 within the next 10 days                                                  j    !
i
'                                    outlining the cause of the malfunction and the plans for restoring                                                        ,
i                          the channel (s) to OPERABLE status.
: b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
            '                                                                                                                                        /
SURVEILLANCE REQUIREMENTS
              \    4.3.3.8 Each channel of the loose part detection system shall be demonstrated l                  OPERABLE by performance of:                                                                                              /
                                                                                                                                                /
l                \
!                ,i          a.      a CHANNEL CHECK at least once per 24 hours,                                                  '
'                  \        b.
j
        ,            i a CHANNEL FUNCTIONAL TEST at least once per 31 days, and                          /
                      ._                                                                                      '_ , -                                        1
                          \c.      a CHANNEL CALIBRATION at least once per 18 months,                  p 4
I                  PALO VERDE - UNIT 1                                3/4 3-69
 
t
                          "          ~
                                              -~~~~ ,.
                  /,,                    TA8LE 3.3-12
              /                                                    ~s''~- s_                                    j
        /
LOOSE PARTS SENSOR LOCATIONS
                                                                            ''N                            -
      /    INSTRUMENT NO.                              LOCATION JSVNYE - 1 UPPER VESSEL A (STUD BOLTS)
    ,      JSVNYE - 2 UPPER VESSEL B (STUD BOLTS)
JSVNYE - 3 LOWER VESSEL A (INCORE N0ZZLE)                  s
                                                                                                        \
    ,      JSVNYE - 4 LOWER VESSEL B (INCORE N0ZZLE)                    \
JSVNYE - 5 SG-IA (HOT LEG)
JSVNYE - 6 SG-18 (COLD LEG 1A)                                !
JSVNYE - 7 SG-2A (HOT LEG)                                /
JSVNYE - 8 SG-2B (COLD LEG'2A)
                                                                                                  /
                                                                                                /
                                                                                        ,/
                                                                                      /
e p/
                                                                  /'
e e
PALO VERDE - UNIT 1              3/4 3-70
 
l
                  /    . INSTRUMENTATION                                              -
RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION                            N LIMITING CONDITION FOR OPERATION
[f)
'                      3.3.3.9 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERA 8LE with their alarm / trip setpoints set to
                      ' ensure that the Ifmits of Specification 3.11.2.1 are not exceeded. The alarm / trip setpoints of these channels shall be determined and adjusted in                                              ,
accordance with the methodology and parameters in the 00CM.
APPLICA8ILITY: As shown f.n Table 3.3-13.
i,
                      . ACTION:
: a.        With a low range radioactive gaseous effluent monitoring instrumenta-tion channel alarm / trip setpoint less corservative than required by the above Specification, immediately suspend the release of radioac-tive gaseous effluents monitored by the affected channel, or declare the channel inoperable, or change the setpoint so it is acceptably conservative.                                                                              i
: b.      With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown                      ,
in Table 3.3-13. Restore the inoperable instrumentation to OPERABLE A
                                        . status within 30 days and, if unsuccessful, explain in the next Semi-annual Radioactive Effluent Release Report why this inoperabfifty was not corrected within the t'ine specified.
: c.      The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
                                                                                                                        /
SURVEILLANCE REQUIREMENTS 4.3.3.9 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE' CHECK, CHANNEL CALIBRATION, and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-8.
                                                                                                    ,/
4 PALO VERDE -JUNIT 1                  -
3/4 3-71 f
 
                                                                                                                                                                            ~~'..._
QR D9 D
x,.'
TABLE 3.3-13                                                          .
[o '.                                                                                                          RADIDACTIVE GASE00S' EFFLUENT HONITORING INSTRUMENTATION x.
                              *g                                                                                                                            *
                                                                                                                                                                                                                                                          \
m                                                                                                                                          MINIMUM CHANNELS INSTRUMENT                                                OPERABLE e
APPLICA8ILITY              ,    ACf!ON
                            . g              1.                                GASEOUS RADWASTE SYSTEM 4
y                                                  a.                                  Noble Gas Activity Monitor -
                                  ;                                                                                  Providing Alarm and Automatic l                                                                                  Termination of Release #RU-12                                          1                #                            35
: b.                                  Flow Rate Monitor                                                      1              #                              36 f
i i      2.                                GASEOUS RADWASTE SYSTEM EXPLOSIVE GAS                                                                                                                                            /
                                                                                                                                                                                                                                                                  /
i                                        MONITORING SYSTEM                                                                                                                                                              '
j k                                        a.                                  flydrogen Monitor                                                        2              **
39          ,
y\                                                                                                                                                                      -
y\                                                b.                                  Oxygen Monitor                                                            2              **
39
                                                                                                                                                                                                                                                      /
                                          \'
                                                                                                                                                                                                                                                    /
                                                                                                                                                                                                                                                      /
                                                                                                                                                                                                                                                  /
s                                                                                                                                                                                                .!
s f
                                                                                                                                                                                                                                    ,Y i
r e
e        .
9
                                                                                                                                                                                                        --_m      -- .-~.m.._ m--.m                                      -.
 
                                                                    ^
qq                                                  .
                                                                                                              *\ ~
                                                                                            ~~~--q,          g TABLE 3.3-13 (Continued)                          N RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION                    ,
g                                                                                                                  \
      $                                                            MINIMUM CHANNELS
                                                                                                                              \
7            INSTRUMENT                                          OPERABLE          APPLICABILITY        ACTION -
E  3. CONDENSER EVACUATION SYSTEM                                                                                                  \
[      A. Low Range Monitors                                                                                                        !
: a. Noble Gas Activity Monitor #RU-141                1                  1,2,3,4            37
: b. Iodine Sampler                                    1                  1,2,3,4            40                            i'
  .                c. P, articulate Sampler                            1                  1, 2,-3, 4          40 l
: d. Flow Rate Monitor                                1                  1,2,3,4            36
: e. Sampler Flow Rate Measuring Device l
1                    1,2,3,4            36                          l l    g      8. High Range Monitors                                                                                                          1
      +          ,
l w
e
: a. Noble Gas Activity Monitor #RU-142                1                    1,2,3,4            42 d            b. Iodine Sampler                                  1        ,          1, 2, 3, 4          42                      f I
: c. Particulate Sampler                              1                    1,2,3,4            42                      l
: d. Sampler Flow Rate Measuring Device                1                  1,2,3,4              42                  .
: 4. PLANT VENT SYSTEM                                                                                                      '
A. Low Range Monitors                      ,    .
l
: a. Noble Gas Activity Monitor #RU-143                                    *                                '
1                                      37          j
: b. Iodine, Sampler                                  1
* 40        :
: c. Particulate Sampler                              1
* 40
                                                                                                                            /
: d. Flow Rate Monitor                                1
* 36
: e. Sampler Flow Rate Measuring Device                                    *                          /                              '
1                                      36 /
i
                                                                                                                        /
                        ~'
                              .-                                                                          /
N.                                                                  ,,e        g
                                                                                                                                          -.        ,e. .
 
                                                                                                                                                                      - - - - - - - -          ~
Q a
w%
                                                                                                                                                                                            ~-                                                    *%.
y 7 ''
TA8LE 3.3-13 (Continued) x~
h                      j/                      RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION                                                    \
K E                                                                                                  MINIMUM CHANNELS 7                        . INSTRUMENT                                                                    OPERABLE              APPLICABILITY            ACTION E                .
PLANT VENT SYSTEM (Continued)
                                                                                          --s e                  B. High Range Monitors
                                          ,                                                                          a.        Noble Gas Activity Monitor #RU-144                                    1
* 42
: b.        Iodine Sampler                                                        1
* 42
: c.        Particulate Sampler                                                    1
* 42
: d.        Sampler flow Rate Measuring Device                                    1
* 42
                                                                                          ,                5. FUEL BUILDING VENTILATION SYSTEM A                    A. Low Range Monitors 4                          a.      Noble Gas Activity Monitor #RU-145                                        1                        ##                  37,41
: b.        Iodine Sampler                                                          1                        ##                  40
: c.      Particulate Sampler                                                      1                        ##                    40
: d.      Flow Rate Monitor                                                        1                        ##                    36
{                e.      Sampler Flow Rate Measuring Device                                        1                        ##                    36 B. High Range Monitors                                                ''
{              a.      Noble Gas Activity Monitor #RU-146                                        1                        ##.                  41,42
                                                                                                      ;            b.        Iodine Sampler                                                          1                        ##                    42
: c.      Particulate Sampler                                                      1                        ##                    42
: d.      Sampler Flow Rate Measuring Device                                        1                        ##                    42    ,
ll j
                                                                                                                                        ~ ~ ~ . _ _ . _ _ _ . _
e s                                                                                                                            .,                      _ . = _ ~
 
t J.
                                              ,f-                        " ~.                                                                                    l
                                          ''-                                            %                                                                      1 i                                  ,=
TABLE 3.3-13 (Continued f~ ~ x ...                            .
x TABLE NOTATION                              N w                                  i i
N            '
i t
j
* r          US RADWASTE SYSTEM operation.                                                                          -
                                # During waste gas release.
l
          .QdA [! N In MODES 1, 2, 3, and 4 or when irradiated fuel is in the fuel storage pool, l
j ACTION 35 -        With the number of channels OPERABLE less than required by the f                                Minimum Channels OPERABLE requirement, the' contents of the                                            l
{                                tank (s) may be released to the environment provided that prior                                      ;
to initiating the release:                                                                          ;
                  ;                              a.      At least two independent samples of the tank's contents                                    f
,                  !                                    are analyzed, and
                    \                                                                                                                              /
l l                            b.      At least two technically qualified members of the facility                            l l                                    staff independently verify the release rate calculations                              ;
i                                    and discharge valve lineup;
                                                                                                                                              /
                      \                          Otherwise, suspend release of radioactive affluents via this                                f                      i
                      ;                        pathway.                                                                                  I j ACTION 36 -            With the number of channels OPERABLE less than required by the
{                        Minimum Channels OPERABLE requirement, effluent releases via i                        this pathway may continue provided the flow rate is estimated                        i at least once per 4 hours.                                                          !
l                                      -
t
                      ! . ACTION 37 -          With the number of channels OPERA 8LE less than required by the                      '
Minimus Channels OPERABLE requirement, effluent releases via                          5 this pathway may continue provided the actio'ns of (A) or (8)                          1
                      ;                        are performed.
: a.      Initiate the Preplanned Alternate Sampling Program of                            I Specification 6.16 to monitor the appropriate parameter (s).
k-                    b.      Place moveable air monitors in-line or take grab samples l                              at least once per 12 hours.                                                      ,
                    .I                                                                                                                      ,
1 ACTION 38 -            With the number of channels OPERABLE less than required by the
                  ,l                            Minimus Channels OPERABLE requirement, immediately suspend
{                          PURGING of radioactive effluents via this pathway.
          ~            \ ACTION 39- With                the number of channnels OPERABLE one less than required by the Minimum Channels OPERA 8LE requirement, operation of the
                , i
                          \                    GASE0US RADWASTE SYSTEM may continue provided grab samples                      .,
.                          \                    are taken and analyzed daily. With both channels inoperable X. N              operation may continue provided grab samples are taken and s          analyzed (1) every 4 hours during degassing operations, and N      (2) daily during other operations.                                                                        ^
W PALO VERDE - UNIT 1                              3/4 3-75
  '-. .                                                                                .  - - - , , . . _ _          , . - .                            p. m.
 
i
                                                                                                      .          l TABLE 3.3-13 (Continued)                          .c.  ,      _
TABLE NOTATION 1
1 With the number of channels OPERABLE less than required by the Y'.9) j ; ACTION 40 - Minimum Channels OPERABLE requirement, effluent releases via                              -
(          l                the effected pathway may continue provided samples are contin-
            !                uously collected with auxiliary sampling equipment as required I                  in Table 4.11-2 within one hour after the channel has been
        !                    declared inoperable.
ACTION 41 -    With the number of channels OPERABLE less than required by the              '
j      !
Minimum Channels OPERABLE requirements, comply with the ACTION b of Specification 3.9.12 or operate the fuel building essential              ,
l                    ventilation system while moving irradiated fuel.                            -
l      ACTION 42 -    With the number of channels OPERABLE less than required by the i
Minimum Channels OPERABLE requirement restore the channel to OPERABLE status within 72 hours or:                                        !
: a. Initiate the Preplanned Alternate Sampling Program of Specification 6.16 to monitor the appropriate parameter (s) when it is needed.
: b. Prepare and submit a Special Report to the Commission
;                                  pursuant to Specification 6.9.2 within 30 days following the event outlining the action (s) taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.                                -
N,
                          'N
* g l                                                                  .
l i
PALO VERDE - UNIT 1                  3/4 3-76 i
 
D TABLE 4.3-8
                                                                                                                                          ~ ~
sg
                                                                                                                                                              ~
                  $                RADI0 ACTIVE GASEQUS EFFLUENT MONITORIP8G INSTRUMENTATION SURVEILLANCE REQUIREMENTS N
E                                                                                                                            CHANNEL              MODES IN WHICH 7    INSTRUMENT CHANNEL                        SOURCE                CHANNEL      FUNCTIONAL                SURVEILLANCE c                                                          CHECK                      ,
CHECK        CALIBRATION              TEST                  IS REQUIRED x
U    1. GASEOUS RADWASTE SYSTEM H                                                                                                                                                                                .
: a. Noble Gas Activity Monitor -                                                                                                                                        $
Providing Alarm and Automatic Termination of Release RU-12                    P                          P                    R(3)                  Q(1),P###            #
: b. Flow Rate Monitor                                P                        N.A.                    R                    Q,P###              #
: 2. GASE0US RADWASTE SYSTEM
{          EXPLOSIVE GAS MONITORING SYSTEM
: a. Hydrogen Monitor (continuous)                    D                          H.A.                Q(4)                    M                    **
: b. Hydrogen Monitor (sequential)                    D                          N.A.                Q(4)                    M                    **
: c. Oxygen Monitor (continuous)                      D-                        N.A.                Q(5)                    M                    **
: d. Oxygen Monitor (sequential)                    D                          N.A.                Q(5)                  M                    **
i 4,
                                                                                                                                      ~_                                            ;
_ ,.J
                                                                                                                                                        ~          _
                                                                                                                                                            ~
                                                                    ..... .    ._                        -    . _ _ . _ _ _ ~
 
r-to'6o h.
                                                                                                                    ~
TABLE 4.3-8 (Continued) g            RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS                                                i in
"'                                                                                        CHANNEL                MODES IN WHICH                          N.
CHANNEL'      SOURCE    CHANNEL            FUNCTIONAL              . SURVEILLANCE-INSTRUMENT                                CHECK        CHECK  CALIBRATION              TEST                    IS REQUIRED                              \
h  3. CONDENSER EVACUATION SYSTEM
                                                                                                                                                                  \
(RU-141 and RU-142)                                                                                                                                          i
                                                                                                                                                                      \
: a. Noble Gas Activity Monitor          D            M        R(3)                    Q(2)                                1, 2, 3, 4                        j
                                                                                                                                                                        \
: b. Iodine Sampler                      N.A.          N.A.      N.A.                    N.A.                                1, 2, 3, 4
: c. Particulate Sampler                N.A.          N.A.      N.A.                    N.A.                                1, 2, 3, 4
: d. Flow Rate Monitor                    D            N.A.      R                          Q                              1, 2, 3, 4                        l t
: e. Sampler Flow Rate Measuring          D    .
N.A.      R                          Q                              1,2,3,4                          [
T          Device 5
: 4. PLANT VENT SYSTEM (RU-143 and EU-144)                                                                                                                                      !
: a. Noble Gas Activity Monitor          D            M        R(3)                    Q(2)                                *
: b. Iodine Sampler                      N.A.          N.A.      N. A.                  N.A.                                *
: c. Particulate Sampler                N.A.          N.A.      N.A.                    N.A.
* j
: d. Flow Rate Monitor                    D            N.A.      R                          Q
* I i
: e. Sampler Flow Rate Measuring          D            N.A.      R                          Q Device                                      .                                                                                          e
                                                                                                                                                      /
N                                                                                                                                          :l
        'x                                                                                                                                  /
ll
                                                                                                                                            /
                                                                                                                                          /
                                          ~.._''~"~-.,                                                                            '
                                                                                                                                    /
                                                              '~~~    . , , _ ,
w.x . -.      - . ~ . .      -.-#              ..                                              .
 
7,                                        ,    ,,
c.
O-TABLE 4.3-8 (Continued)                                              *
                                                                                                                                                        ~.
  $o              -                RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATICN SURVEILLANCE REQUIREMENTS                                        -s g                                                                                                                                                                s 5
  "                                                                                                                                CHANNEL
          '                                                                                                                                        MODES IN MIICN ..
CHANNEL                      SOURCE          CHANNEL                    FUNCTIONAL        SURVEILLANCE i
CHECK
[t['
z      INSTRUMENT                                                                    CHECK        CALIBRATION                    TEST            IS REQUIRED Qi            S. FUEL BUILDING VENTILATION SYSTEM i
4 sl                  (RU-145 and RU-146)
: a. Noble Gas Actvity Monitor
      ;,                                                          D                        M            R(3)                          Q(2)                  ##                        l
: b. Iodine Sampler                                                                                                                                                .\
N.A.                      H.A.            N.A.                          N.A.                  ##
: c. Particulate Sampler                  N.A.                      N.A.            N.A.                          N.A.                  ##                        l
  ,        ,          d. Flow Rate Monitor                        D                    N.A.            R                            Q                    ##                      -f s          ,
* I
: e. Sampler Flow Rate Measuring              D                    N.A.            R                            Q                    ##
  "4          ,          Device                                                                                                                                                  <
                                                                                                                                                                                    /
l l
l t
                                                                                                                                                                            ?
                                                                ,                                                                                                        /
                                                                                                                                                                      /                    .
                                                                                                                                                                      /
                                                                                                                                                                  .l
                                                                    '            ~
                                                                                          ~~-
i f
t
                                                                                                                        ,.+"*'
* f--                                    -~~.
                                                                                                                            ~
f
                                                        ./                                                                            N p-/.            TA8LE 4.3-8 (Continued)                                                            x      ,
                              ,/
                                    /                                  TABLE NOTATIONS                                                  c                          4
                        /* At all times.                                                                                                                                            i
[80                ,/ ** During GASEOUS RA0 WASTE SYSTEM operation.                                                                                                            "
l# During waste gas release.                                                                                          ,
                          ## During MODES 1, 2, 3 or 4 or with irradiated fuel in the fuel storage pool.
                        ### Functional test should consist of, but not be . limited to, a verification of
                ;                system isolation capability by the insertion of a simulated alarm condition.
'              [(1) The          of thisCHANNEL            FUNCTIONAL pathway and' control      room alarm annunciation              TEST shall            occurs also if anydemonstrate of the                      that a i                following conditions exists:
i i              1.      Instrument indicates measured. levels above the alare/ trip setpoint.
: 2.      Circuit failure.
: 3.      Instrument indicates a downscale failure.
4 l                4.      Instrument controls not set in operate mode.
j(2) The        alarm CHANNEL        FUNCTIONAL annunciation  occurs if anyTEST  of the shall                      alsoconditions following            demonstrate  exists:that control room
              ;                  1.    . Instrument indicates measured levels above the alarm setpoint.
.                              2.      Circuit failure.                                                                                                                    I
: 3.      InstrumentindicatesadowIscalefailure.
4.-      Instrument cdntrols not set in operate mode. _
f  '
                                                                                                                                                ~
(3) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards                                                                                  -
(NBS) or using standards that have been obtained from suppliers that                                                                        i participate in measurement assurance activities with NBS. These staadards                                                                  f j                      shall permit calibrating the system over its intended range of energy and
          ;                      measurement range. For subsequent CHANNEL CALIBRATION, sources that'have                                                                  /
i' i
been related to the initial calibration shall be used.                                                                                  ,'
: g.                                                                                                                            !
i (4) The CHANNEL CALIBRATION shall include the use of standard gas samples                                                                                l                    :
            !                    containing a nominal:                                                                          .
i                    !
f                  ~ 1.        One volume percent' hydrogen, balance nitrogen, and                                                          ~
l f
: 2.      Four volume percent hydrogen, balance nitrogen.
4 i
(                      l
        ! (5) The CHANNEL CALIBRATION shall faclude the use of standard gas samples                                                                              i i                        containing a nominal:                                                                                                        /
[                                1.      One volume percent oxygen, balance nitrogen, and f
I      j.                      2.      Four volume percent oxygen, balance nitrogen.
            \
                                        *%=    .%    ,_, _
                                                                                                        ~~~
PALO VERDE - UNIT 1                                  3/4 3-80                                -          ~
1 I
                .                  I w              -      ---
                                                                %.e-.      - - , , _ . - _ . - , , , ,            . - , - _        -_        ,            y.                      , . . _
 
3/4.4 REACTOR COOLANT SYSTEM                                                      *-
3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 Both reactor coolant loops and both reactor coolant pumps in each loop shall be.in operation.
APPLICABILITY: MODES 1 and 2.*
ACTION:
With less than the above required reactor coolant pumps in operation, be in at
          -least HOT STANDBY within 1 hour.
l SURVEILLANCE REQUIREMENTS 4.4.1.1 The above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.
i l
l 1.
i "See Special Test Exception 3.10.3.  ,
I i
                                                                                                  ~
l          PALO VER0E - UNIT 1                  3/4 4-1 l
l' l' .  ._.        .            ._ .    .        .        . _ - _ _ _ .  ,  -. _.
 
7-~
jc(,({ REACTOR COOLANT SYSTEM)
HOT STANOBY                                                        -            -    --
i LIMITING CONDITION FOR OPERATION
      ~
        ~~
                          '3.4.1.2 The reactor coolant loops listed below shall be OPERABLE and at least
                                                                                                                    ~
one of these reactor coolant loops shall be in operation.*
: a. Reactor Coolant Loop 1 and its associated steam generator and at least one associated reactor coolant pump.
                                . b .' Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump.
g                        APPLICABILITY: MODE 3.
l                        ACTION:
l 4
: a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours or be in HOT SHUTDOWN within the next 12 hours.
!                                b. With no reactor coolant loop in operation, suspend all operations
,                                      involving a reduction in baron concentration of the Reactor Coolant
,'                                    System and immediately initiate corrective action to return the required reactor coolant loop to operation.
i
                        ' SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required reactor coolant pumps, if not in
,                        operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2. At least one reactor coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours.                                                  .
4.4.1.2.3 The required steam generator (s) shall be determined OPERABLE verifying the secondary side water level to be > 25% indicated wide range level at least once per 12 hours.
_ _ . . _ _ . _ - _          u. _
g                        *All reactor coolant pumps may be deenergized for up to I hour provided
;                          (1) no operations are permitted that would cause dilution of the Reactor l                          Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
l
                    ,    PALOVERDE-UNIT 1);                        3/4 4-2 L
: c.  -
                  --q.
 
                        .-                w                                                                                    '
    . ,      ' REACTOR COOLANT SYSTEMN*
ew E
                  ' HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loop (s)/ train (s) listed below shall be OPERA 8LE and at least one reactor coolant and/or shutdown cooling loops shall be in operation.*
: a. Reactor Coolant Loop 1 and its associated steam generator and at least one associated reactor coolant pump,**
: b.  ' Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump,**
Shutdown Cooling Train K,
                  @ c.d.
                                                          ~
Shutdown Cooling TrainilCS ~,
APPLICABILITY: MODE 4.
ACTION:
: a. With less than the above required reactor coolant and/or shutdown cooling loops OPERABLE, immediately initiate corrective action to return the required loops to OPERA 8LE status as soon as possible; l                                if the remaining OPERABLE loop is a shutdown cooling loop, be in
;                                COLD SHUTDOWN within 24 hours.
~                                                                                            .__._ _ _..              . _ . _
El- With no reactor coolant or shutdown cooling loop ~iii''ojierafi6n~,'ii0i~pe'n~d~~ -_ _ . _ _ _ .
                                                                                                                                      ~
all. operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
                    "All reactor coolant pumps and shutdown cooling pumps may be deenergized for up to I hour provided (1) no operations are permitted that would cause dilution.of the Reactor Coolant System boron' concentration, and (2) core outlet temperature is maintained at least 10*F below saturation tem f            **A reactor coolant pump shall not be started                                        .
with one or mo f the Reactor dd        3 Coolant System cold leg temperatures less than or equal to 255'F during
  @      " c661down,~or(255?F during heatup_,_unless the secon_d_ary water _lemperature gf          s  (saturation-temperature steam generator ii~less;thancorresponding 100*F above _toTsteam~
each of the generator Reactor Coolant      _pressur_elof System each cold leg temperatures,
    &        %+M Ca appiu'av.ta SFR.                                  .
i
                .PALO VERDE - UNIT 1                        -3/4 4-3
 
N (REACTOR CLANT SYSTEM l
HOT SHUTOOWN                                                                            i SURVEILLANCE REQUIREMENTS                                                                !
t 4.4.1.3.1  The required reactor coolant pump (s), if not in. operation, sh'all be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.                      -
          -4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying the secondary side water level to be > 25% indicated wide range level at least once per 12 hours.
4.4.1.3.3 At least one reactor' coolant or shutdown cooling loop shall be
          -verified to be in operation and circulating reactor coolant at a flow rate greater than or equal to 4000 gpm at least once per 12 hours.
i i                                      .
I' l
(          PALO VERDE'- UNIT 1                  3/4 4-4
                                                                                              .      )
i                                                                                                      l i
t-                                                                                                    )
 
ccL(REACTOR COOLANT SYSTEM                    3                                                                                -
COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION
{-
t 3.4.1.4.1 At least one shutdown cooling loop shall be OPERABLE and in                                                      ,
operation *, and either:                                                                                                    I
: a. One additional shutdown cooling loop shall be OPERABLE #, or
: b. The secondary side water level of at least two steam generators                                              i shall be greater than 25% indicated wide-range level.                                                        j APPLICA8ILITY: MODE 5 with reactor coolant loops filledN.
ACTION:
i
: a. With less than the above required loops OPERABLE or with less than the required steam generator level, immediately initiate corrective action to return the required loops to OPERABLE status or to restore the required level as soon as possible.
: b. With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation.
~
SURVEILLANCE REQUIREMENTS
              '4.4.1.4.1.1 The secondary side water level of both steam generators when required shall be determined to be within limits at least once per 12 hours.                                          .
4.4.1.4.1.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm at least once per 12 hours..
4
                  #0ne shutdown cooling loop may be inoperable for up to 2 hours for surveillance testing provided the other shutdown cooling loop is OPERABLE.
L                  and in operation.
of the Reactor l    h N ACoolant g              reactor        coolant System        _ cold legpump    shalllessnot temperatures                      thanbe or equalstarted to 1 with one or mor $5'F l
7
          .fc6oldown, F29.5'F during heatup, unless the secondary water temperature X M Jaturation temperature corresponding to steam generator pressure) of each steam generator is less than 100*F above each of the Reactor Coolant Systeg j                  cold leg temperatures, l
              *The shutdown cooling pump may be deenergized for up to I hour provided
!                (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is l                maintained at least 10'F below saturation temperature.
              -R - K f;a agijca.d 56R PALO VERDE - UNIT 1                                3/4 4-5
 
M,  REACTORC0dLANT~SYSEMA COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two shutdown cooling loops shall b,e OPERABLE # and at least one shutdown cooling loop shall be in operation.
I APPLICABILITY: . MODE 5 with reactor' coolant loops not filled.
ACTION:
: a.      WithlessthantheaboverequiredloopsOPERABLE,immediatelyinitiate corrective action to return the required loops to OPERABLE status as l                      soon as possible.                                                                ,
: b.      With no' shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return th.e
                  ' required shutdown cooling loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.4(2' At least one shutdown cooling loop shall be dete ned to be in operatilin and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpa at least once per 12 hours.
I                                  .
        #0ne shutdown cooling loop may be inoperable for up to 2 hours for surveillance                '
        ' testing provided the other shutdown cooling loop is OPERABLE and in operation.
The shutdown cooling pump may be de-energized for up to'1 hour provided (1) no operations are permitted that would cause dilution of thedteactorcCoolant SSystem boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
1 l
PALO VERDE - UNIT                          3/4 4-6 L
l t
 
p_..______..
44 (vREACTOR COOLANT SYSTEM                    '%
w 3/4.4.2 SAFETY VALVES                                    .
SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.2.1 A minimum of one pressurizer code safety valve shall be OPERABLE with
                  .a lift setting of 2500 psia i 1L*
APPLICABILITY: MODE 4.
                                                                                  ~~
                                                                                                                          ~
ACTION:
: a. With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation.
: b. The provisions of Specification 3.0.4 may be suspended for up to 12 hours for entering into and during operation in MODE 4 for purposes of setting the pressurizer code safety valves under ambient (HOT) conditions provided a preliminary cold setting was made prior to heatup.
?'
SURVEILLANCE REQUIREMENTS
                  ~4.4.2.1      No additional Surveillance Requirements other than those required by Specification 4.0.5.
1
!                '"The lift setting pressure shall correspond to ambient conditions of the valve l                      a.t nominal operating temperature and pressure.
l-3/4 4-7 PALO VERDE - UNIT 1}
f
 
                . REACTOR COOLANT SYSTEM OPERAhING'    ~
LIMITING CONDITION FOR OPERATION 3.4.2.2 All pressurizer code safety valves shall be OPERABLE with a lift setting of 2500 psia i 1%.*
APPLICABILITY: MODES 1, 2, and 3.
ACTION:                                                                                              -
With one pressurizar code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in at least HOT STAND 8Y within 6 hours and in HOT SHUTDOWN within the following 6 hours with the shutdown cooling system suction line relief valves aligned to provide overpressure protection for the Reactor Coolant System.
s SURVEILLANCE REQUIREMENTS
(
_          _4.4.2.2._.No additional Surveillance Requirements other than_those required by
                    -Specification 4.0.5.                                                                                            - ~ - - ~        ~ -~~ ~~~
l l                                                      4
                      *The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
l i
                          .,___%      -s 3/4 4-8
                                                      ~
PALO VERDE - UNIT 1
            *                              -.-., -      y      ,_-g. - -,--,.,y--  -- ., y,.,  - ,.  --,#~              - * - --- .            --              -.
m-. y
 
l                                                                                                                                              l J      [ 0R COOLANT 3/4.4.3 PRESSURIZER PRESSURIZER LIMITING CONDITION FOR OPERATION
: f.                  [3.4.3.1 The pressurizer shall be OPERABLE with a minimum steady-state water level of greater than or equal to 27% indicated level (425 cubic feet) and a maximum steady-state water level of less than or equal to 56% indicated level (948 cubic feet) and at least two groups of pressurizer heaters capable of being
:                            powered from Class 1E buses each having a . nominal capacity of at least 150 kW.
APPLICA8ILITY: MODES 1, 2, and 3.
ACTION:
: a. With only one group of the above required pressurizar heaters OPERABLE, restore at least two groups to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours,
: b. With the pressurizer otherwise inoperable, restore the pressurizer to
    ~'
            ~
OPERABLE status within 1 hour, or be in at least-HOT STANDBY with the                    -
                                            -reactor trip breakers open within 6 hours and in~ HOT' SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.3.1.1 The pressurizer water volume shall be determined to be within its limits at least once per 12 hours.
4.4.3.1.2 -The capacity of the above required groups of pressurizer heaters shall be verified to be at least 150 kW at least once per 92 days.
                          ~
4.4.3.1.3 The emergency power supply for the pressurizer-heaters shall be demonstrated OPERABLE at least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss-of-
                          'offsite power:
: a. The pressurizer heaters are automatically shed from the emergency power sources, and
: b. The pressurizer heaters can be reconnected to their respective buses manually from the control room.
l                        (PALO VERDE --UNIT 1)                                    3/4 4-9
                              ,  . - - - .      ,, ., ,,.  -,..n-    ,, - - , , , _ , , ,        , , . , - ,                -- , ,
 
h
                                                                    ~
3/4.4.3          PRESSURIZER LIMITING CONDITION FOR OPERATION i  -r  3.4.3 - The pressurizer shall be OPERABLE w'itE              ~~
                                                                                  ^
i
: a. A steady state water volume less than or equal _to 58% indicated-level (1010 cu. ft) but greater than 27% indicated level (445 cu.                        .}
ft.), and ( 99G                                                      0                  ,;
: b. At least two groups of pressurizer heaters capable of being powered                        f from IE buses each having a nominal capacity of at least 150 kw.
APPLICABILITY            S1  2 and 3.                          ' --              -d                          !
ACTION:
l i
: a. With one group of the above required pressurizer heaters operable, 4
restore at least two groups to OPERABLE status within 72 hours or be in at least NOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
: b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in
,                  HOT SHUTDOWN within the following 6 hours.
l                                                                    -
SURVEILLANCE REQUIREMENTS i
4.4.3.1    The pressurizer water volume shall be determined to be within.its i      limit at least once per 12 hours.                                  --          ~ ' ~ - ~ ~ - ~
4.4.3.2 The capacity of each of the above required groups of pressurizer heaters shall be verified to be at least 150 kw at least once per 92 days.
4.4.3.3 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by verifying that on an Engineered Safety Features Actuation test signal concurrent with a loss of offsite power:
            'a. the pressurizer heaters are automatically shed'from the emergency power sources, and
: b. the pressurizer heaters can be reconnected to their respective buses manually from the control room.
l l
l 3/4 4-9                      Amendment Number 9 February 27, 1984
 
                                                                                              ~
REACTOR COOLANT SYSTEM f
AUXILIARY SPRAY                                                                                '
LIMITING CONDITION FOR OPERATION l
3.4.3.2 Both auxiliary spray valves shall be OPERABLE.
APPLICA8ILITY: MODES 1, 2, 3, and 4.
ACTION:
: a. Wfth only one of .the above required auxiliary spray valves OPERABLE,
                  . restore both valves to OPERABLE status within 72 hours or be in at least HOT STAN08Y within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
: b. With none of the above required auxiliary spray valves OPERABLE, restore at least one valve to OPERA 8LE status within the next 6 hours or be in at least HOT STANOBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.
SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The auxiliary spray valves shall be verified to have power available to each valve every 24 hours.                      .
4.4.3.2.2 The auxiliary spray valves shall be cycled at least once per 18 months.
O PALO VERDE - UNIT 1                    3/4 4-10
 
1
,      -
* REACTOR COOLANT SYSTEM 3/4.4.4 ~ STEAM GENERATORS' i                LIMITING CONDITION FOR OPERATION M f # !# M E.                                                        l'            -
                                                        -~ %                                                    3/y.4.Q S O p 'redC,*M6I 4          / 3.4.4 Each steam generator shall be OPERABLE. '                                                        Qy,N#'n'                                        h                                    [,
w APPLICA8ILITY: MODES 1, 2, 3, and 4.                                                                                        N ACTION:
1 With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T
* cold above 2104.                                      3 i
i" SURVEILLANCE REQUIREMENTS
                                                                                                                                                                    \
4.4.4.0 'Each steam generator-shall be demonstrated OPERABLE by performance of                                                                        '
the following augmented inservice inspection program.                                                                                                  i t
4.4.4.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.                                                                                      ,i i
4.4.4.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the                                                                                ,
corresponding action required shall be as specified in Table 4.4-2. The                                                                                        i inservice inspection of steam generator tubes shall be performed at the
              ' frequencies specified in Specification 4.4.4.3 and the inspected tubes shall                                                                                  ,
be verified acceptable per the acceptance criteria of Specification 4.4.4.4.                                                                                i The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for                                                                                  ,
these inspections shall be selected on a random basis except:                                                                                            ;                              '
: a.        Where experience in similar plants with similar water chemistry                                                                    /
indicates critical areas to be inspected, then at least 50% of the                                                                '
tubes inspected shall be from these critical areas.
: b.        The first sample of- tubes selected for each inservice inspection                                                          ,
(subsequent to the preservice inspection) of each steam generator shall. include:
                                                                                                                                          /
                                                                                                                                        /
                                                                                                                                    /                                                                          n
                                                                                                                              /
                                                                                                                        /
r
                                                                                                                      /
                                                                                                                  /
PALO VERDE - UNIT 1                                3/4 4-11                                  ,                          .
                                                                                  - /
                                                                                                                                            .n.---- , ,, . ,                    - - , , - . - - . - - .
    --                -,,--w.,                    .-. , _ _ . ,            , , - , , . - .-,,,-----..-,.--                    .-
 
M
                                                                                                              ^W
                                                                                                      -~
p  7'                                                                                                      N.                  ;
j 4
f ftEACTOR COOLANT SYSTEM                                                                                                            i i.
SURVEILLANCE REQUIREMENTS (Continued)
: 1.      All nonplugged tubes that previously had detectable wall penetrations (greater than 20%).
l                    2.      Tubes in those areas where experience has indicated potential                                            \
p5                    problems.
: 3.      A tube inspection (pursuant to Specification 4.4.4.4a.8.) shall be performed on each selected tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shal1 i                            be selected and subjected to a tube inspection.                                                            ,
4
: c. The tubes selected as the second and third samples (if required by Table 4.4-2) during each inservice inspection may be subjected to a partial tube inspection provided:
: 1.      The tubes selected for these samples include the tubes from those areas of the tube sheet array where tubes with imperfections were previously found.
: 2.      The inspections include those_ portions of the tubes where imperfections were previously found.
;          The results of each sample inspection shall be classified into one of the                                                      !
following three categories:
~
1 Category                                                    Inspection Results C-1                                Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.                                                            j C-2                                One or more tubes, but not more than 1% of the total tubes inspected are defective, or between
,                                                            5% and 1.0% of the total tubes inspected are degraded tubes.
C-3                                More than'10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are, defective.
Note:      In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations
,                                  to be included in the above percentage calculations.
i PALO VERDE - UNIT 1                                              3/4 4-12
                                                              , . , , , , ,    ,-.    -                - - -    ---w 7-
                                                                                                                          -  ---g-.w-      w
 
              '          ~~ ~~~ ~                                                      _ . .
REACTOR COOLANT SYSTEMS                                                                                        .
SURVEILLANCE REQUIREMENTS (Continued)
                                                                                                                                        ~
4.4.4.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be; performed at the following frequencies:
: a. The first inservice i'spection n                shall be performed after 6 Effective Full Power Months but within 24 calender months of initial crit-                                          :
    @                  icality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months after i
the previous inspection. If two consecutive inspections following
                      ' service under AVT conditions, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation h.as not continued and no additional degradation has
                      . occurred, the inspection interval may be extended to a maximum of once per 40 months.                                                                                        ,      i
: b. If the results of the inservice inspection of a steam generator                                                    I conducted in accordance with Table 4.4-2 at 40 month intervals fall                                      j into Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of L                      Specification 4.4.4.3a.; the interval may then be extended to a                                        l          .
maximum of once per 40 months.                                                                        j
: c. Additional, unscheduled inservice inspections shall be performed on                              ,
each steam generator in accordance with the first sample inspection                              /
j                      specified in Table 4.4-2 during the shutdown subsequent to any of                              /                  l f
the following conditions:                                                                      /
: 1.      Primary-to-secondary tubes leaks (not including leaks                              '
                                                                                                                    /
originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.5.2.                                            . ,/                          ;
: 2.      A seismic occurrence greater than the Operating Basis                      /
Earthquake.
: 3.      A loss-of-coolant accident requiring actuation of the                    /
engineered safeguards.
l
: 4.      A main steam line or feedwater line break.
l                                                                                                    l
                                                                                                    /
                                                                                              /
PALO VERDE - UNIT 1                            3/4 4-13
_ _ ~ . --.              _
 
3                                                                                                                                            ,
l Y.y 3
                                                                                  ~
N
                    /      REACTOR COOLANT SYSTEM                                                                    N                                            ,
N
                                                                                                                                  \
SURVEILLANCE REQUIREMENTS (Continued)                                                                      i i
                                                                                                                                              \
4.4.4.4  Acceptance' Criteria                                                                                    \                        l v                                                                                                                                              \
: a. As used in this Specification                                                                              i
                                                                                                                                            ~
                                                                                                                                                  \
: 1.            Imperfection means an exception to the dimensions, finish, or                                                      i contour of a tube from that required by fabrication drawings or                                l
;;                                                  specifications. Eddy-current testing indications below 20% of                                  !
the nominal tube wall thickness, if detectable, may be considered I as imperfections.                                                                              !
* l
: 2.            Deoradation means a service-induced cracking, wastage, wear, or                                  ;
                                                . general corrosion occurring on either inside or outside of a                                      l tube.                                                                                            l
: 3.            Decraded Tube means a tube containing imperfections greater                                      ,
l 4      ,
            ' y, than or equal to 20% of the nominal wall thickness caused by                                      ;                  l j(;                                      degradation.                                                                                    l o:        T'                                                                                                                                l                  l g                                      4.            % Deoradation means the percentage of the tube wall thickness                                    l                  !
4                          affected or removed by degradation.                                                              j
      .s                                                                                                                                            ;
: 5.            Defect means an imperfection of such severity that it exceeds                                    I
                                              .      the plugging limit. A tube containing a. defect is defective.                                                  j s                                                                                              F,
: 6.          ' Pluccing Limit means the imperfection depth at or beyond which                                    ;
l                                                    the tube shall be removed from se:vice and is equal to 40%                                      ;
of the nominal tube wall thickness.                                                              !              ;
                                                                                                                                                .I                    :
3                                  7.        - Unserviceable describes the condition of a tube if it leaks or                                                    !
contains a defect large enough to affect its structural integrity                            il in the event of an Operating Basis Earthquake, a loss-of-coolant                          -
{
[-
accident, or a steam line or feedwater line break as specified                            i                    3, in 4.4.4.3c., above.                                                                    j l
l 8 .-          Tube Inspection means an inspection of the steam generator tube f gy                s,                                from the-point of entry (hot leg side) completely around the                        l                            l
              \                                      U-bend to the top support of the cold leg.                                          /
        .o                                                                                                                              7
: 9.            Preservice Inspection means an inspection of the full length of each tube in each steam generator performed by eddy current                    !
techniques prior to service to establish a baseline                            /
[.*
:                                                                                                                                  l                                    .
\                                                                      .
L Li 4
!                                                                                                                                                                        l PALO VERDE - UNIT 1                            3/4 4-14                                                                                      '
t
                                            ,,.,...e      '"a
 
p f~                                                -~---...-.s.--.-
REACTOR COOLANT SYSTEM                                                                                                .
4 i
SURVEILLANCE REQUIREMENTS (Continued)                                                                                                      ,
I i                                                                                                                                  .
condition of the tubing. This inspection was performed prior to the field hydrostatic test and prior to initial POWER OPERATION
                    !                          using the equipment and techniques expected to be used during i                          subsequent inservice inspections.                                                          )
                      \
i              b. - The steam generator shall be determined OPERA 8LE after completing t                the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by i
Table 4.4-2.
4.4.4.5              Reports
: a. Within 15 days following the completion of each inservice inspection                                          ;
of steam generator tubes, the number of tubes plugged in each steam                                          ;
generator shall be reported to the Commission in a Special Report o,                                      pursuant to Specifigation 6.9.2.
: b. The complete results of the steam generator tube inservice inspection -                                            i
                              ,.          shall be submitted to the Commission in a Special Report pursuant to                                              ,
Specification 6.9.2 within 12 months following completion of the inspection.- This Special Report shall include:
i
: 1. Numbar and extent of tubes inspected.                                  .
:                                                                                                        i l            2. ,  Location and percent of wall-thickness penetration for each                                                i indication of an imperfection.                                                                              ,
: 3. Identification of tubes plugged.                                                    !
l
: c. Results of' steam generator tube inspections which fall into
<                                          Category C-3 shall be reported in a Special Report to the Commission                                              j-pursuant to Specification 6.9.2 within 30 days and prior to resumption of plant operation and shall provide a description of investigations conducted to determine cause of the tube degradation; and corrective measures taken to prevent recurrence.
l
                                                                            ~
l l
l l:                    '                                                                                              '
                                                                                                                                                    .          l PALO VERDE - UNIT 1                                3/4 4-15              -
 
I l
                                                                  ,. / ..
                                                                                        ----~ ~ .
                                                                                                                ' ~~ *                                            ,\
                                        ,                                                                                            , x TABLE 4.4-1
                          /
[                        MINIMUM NUMBER OF STEAM GENERATORS TO BE                                                                        ;
INSPECTED DURING INSERVICE INSPECTION                                                                      +
                                                                                                                                                              .I p5          'The inservice inspection may be limited to one steam generator on a rotating i              schedule encompassing 6%-of the tubes if the results of the first or previous
                    ; inspections indicate that all steam generators are performing in a like manner.
                  ! Note that under some circumstances, the operating conditions in one or more
                  ' steam generators may be found to be more severe than those in other steam generato rs. Under such circumstances the sample sequence shall be modified to inspect the most. severe conditions.                                                                                                .
e i=
i
* l l
l                                                                                                                                ,1 l                                                                                                                              I r                                                                                                                            i l                                                                                                                        ?
l                                                                                                                      '
: l.                                                                                                                j
                                                                                                                  /
P
                                                                                                          /
PALO VERDE - UNIT 1                                3/4 4-16 i              . . _ _ . . .                                  =                  - . -
L
 
    ._                                          _                                      _ ~_.
V
                                                                                                                                                                                                            ~
                                                                                                                                            ~ TABLE 4'.4-2~
                                                                                                                                                                  ~ ~
                                                                                                                                                                                                                              'N g-N, 4                                                                                                                                                                                                                            i 3
[                                                          STEAM GENERATOR TUBE INSPECTION                                                                                  \
m                                                                                                                                                                              \
IST SAMPLE INSPECTION                                          2ND SAMPLE INSPECTION                        3RD SAMPLE INSPECTION
[      Sample Size            Result            Action Required                  Result'        Action Required              Result                Action Required z
1                                                        U    A minimum of                C-1                    None                        N. A.                  N. A.                N. A.                            N. A.
e    S Tubes per i                                                              S. G.                                                                                                                                                                      i C-2            Plug defective tubes                C-1                    None                  N. A.                            N. A.          I and inspect additional                                                                                                              i Plug delactive tubes            C-1            None 2S tubes in this S. G.                C-2          and inspect additional          C-2            Plug defective tubes t                                                                                                                                                          4S tubes in this S. G.
Perform action for C-3              C-3 result of first sample R                                                                                              Perform action for
* C-3        C-3' result of first              N. A.                            N. A.            \
a                                                                                              sample O
                                                          "                              C-3            inspect all tubes in              All other                                                                                          \
this S. G., plug de.              S. G.s are                None                N. A.
                                                                                                                                                                                                                                                \
                                                                                                                                                                .                                                          N. A.                  j factive tubes aml .                C-1 inspect 2S tubes in                                                                                                                        \,
Some S. G.s p , form action for each other S. G.                                                                  N. A.                            N. A.                    {
C-2 but no C-2 restdt of second additional    sample Notification to NRC              S. G are pursuant to $50.72                C-3 (b)(2) of 10 CFR                  Atklitional Inspect all tubes in
                                                                        .                                Part60                            S. ,G. is C-3 each S. G. and plug                                                                          !
defective tubes.
Notification to NHC              N. A.                            N. A.
pursuant to $50.72                                                                          'f (b)(2) of 10 CFR Part 50                                                                                  /
                                                            .                                                                                                                                                                                  /
S=3            % Where N is the number of steam generators in the unit. and n is the number of steam generators inspected n          during an inspection                                                                                .
                                                                                                " ~ ~ '                                                  *
                                                                                                                      .-_...._,,__.m,.--"-            **
: m. e- e,w    *ee.+ee..s.      ..m  e s>e.ee
 
i REACT 0'R COOLANT SYSTEM                                                                                    -
j.
    .              3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE l
LEAKAGE DETECTION SYSTEMS                                                                                                    ,
LIMITING CONDITION FOR OPERATION Obdo GCO'/iLLAMr                                      REco! Pcm/3'T3
  @                  h A The            vefollowing          h sReactor m e.      us.9.s. I va, rms. kam
              . / 3' .4.5.1                                              Cliolant System leakage' detection systems shall
                / be OPERA 8LE:
i l                            a.            A containment atmosphere particulate radioactivity monitoring system,
: b.            The containment sump level and flow monitoring system, and
: c.          .The containment atmosphere gaseous radioactivity monitoring system.
                  'APPLICA8ILITY: MODES 1, 2, 3, and 4.
A ACTION:
With only two of the above required leakage detection systems OPERABLE, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours when the required gaseous and/or particulate radioactivity monitoring system is inoperable;'othenvise, be in at least HOT STANOBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
: j.                  SURVEILLANCE REQUIREMENTS 4.4.5.1 The leakage detection systems shall be demonstrated OPERABLE by:                                              /
l
: a.            Containment atmosphere gaseous and particulate monitoring                                  !
system performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL-                    ,
FUNCTIONAL TEST at the frequencies specified in Table 4.3-3,                          j i
                          \b.            Containment sump level and flow monitoring system performance of                      i CHANNEL CALIBRATION at least once per 18 months.                                ,-
l                                                                                                                            ,
l                                                                          '~                    -      . _
l l-i p                                                          ,
I l
PALO VERDE - UNIT 1                                          3/4 4-18 i
l~                                                                                                                                                  l 1
        . _ . .              . - . _ _ _ . _ - _ _ _ _ _ _ _ . _ _ _ _                    _ _ _ _ ..-.- _ _ _....--_ _ .-._-- . s          -
 
                                                                                                                                  ,            !)I
                                                                                                                                                    ;1 7 -                      -
          - c cL . .(REACTOR COOLANT SYSTEM,                                                                                                        ,
OPERATIONAL LEAKAGE                                                                                                            ;,
LIMITING CONDITION FOR OPERATION t
3.4.5.2          . Reactor Coolant System. leakage shall be limited to:
: a.      No PRESSURE BOUNDARY LEAKAGE,
: b.      1 gpa UNIDENTIFIED LEAKAGE,                                                                          .
: c.      1 gpa total primary-to-secondary _ leakage through hteam generators,
                                                      ~
                        ."          qarid'720Tal1ons per day thridiah any one steam generato6                                                        ,
: d.      10 gpsi IDENTIFIED ' LEAKAGE from the Reactor Coolant System, @
                                      -1gpaleakageataReactorCoolantSystempressureof2250t20psTai
                @            e.
                                      'from any Reactor Coolant System pressure isolation valve specified in Table 3.4-1.
APPLICABILITY: MODES 1, 2, 3, and 4
                    .ACTIO!D
                                                                                                                                            ~
: t.      With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 6 hours and in COLD SHUTOOWN within the following 30 hours,
: b.      With any Reactor Coolant System leakage greater than any_one_of the S,            limits,excludingPRESSUREB0UNDARYLEAKAGE(andleakagTe_from_Reactae
                              't
                                      <Co61      ant Sysles_p~ressure71 withinslimits.within          4 hours or solation    'be in at~ val _yes)least HOT STAND 8Y withinred W
                                  *- the neg hours and in COLD SHUTDOWN within the following 30 hours.                            .
With~any Reactor Coolant System pressure isolifion valve leikage s
[c. greater than the above limit, isolate the high pressure portion of                                    -
            @                          the affected system from the low pressure portion ~within 4 hours by~^
                                                                                                                                        ~~
use of at least one closed manual or deactivated automatic valve, i                                      or be in at least HOT STAND 8Y within the next 6 hours and in COLD
:                                      SHUTOOWN within the following 30 hours.                                                                  .
                          !' 'd.        With RCS leakage alarmed and confirmed in a flow path with no flow rate indicators, commence an RCS water inventory balance within 1-hour to determine the leak rate. __ _ _ _                  .
SURVEILLANCE REQUIREMENTS 4.4.5.2(1) Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:
* y                                        l g            a.      Monitoringthecontainmentatmospheregaseous&id; particulate radioactivity monM o;r at least once per 12 hours.                                                            j
: b.      Monitoring the containment sump inventory and discharge at least once per 12 hours.
PAL 0 VERDE - UNIT 1 '                                            3/4 4-19
      ,7 . . . . . . . .
_ ~ . . _ _ . - _ - - __ .
 
      .                          ~.
iI
                                                                                                                                    ,t1
                                ,7          .
h
* g,            REACTOR COOLANT SYSTEMi SURVEILLANCE REQUIREMENTS (Continued)
: c. Performance of a Reactor Coolant System water inventory balance at least once per 72 hours.
7                                                                  N Monitoring the reactor head flange leako.ff_ system _at least once per, 24 hours.
                                                                                  ~
[f 4.4.5.2.2            EachReactorCoolantSystempressureisolationvalvespecifiediD
        ,                      Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within
  -.          1                its limit.                                                                            y
          .7
: a. At least once per 18 months, l
            'l                        b.* Prior to entering MODE 2 whenever the plant has been in COLD                l
              !                            SHUTDOWN for 72 hours or more and if leakage testing has not been          !
l                            performed in the previous-9 months, i
j                        c. Prior to returning the valve to service following maintenance,                .
repair or replacement work on the valve,                                      I
: d.
* Within 24 hours following valve actuation due to automatic or                    l 3
manual action or flow through the valve,                                        i t                                                                                                              j 7
i                .      e.* Within 72 hours following a system response to an Engineered Safety l            l                            Feature actuation signal.                                        ~
l            l l            i                  The provisions of Specification 4.0.4 are not applicable for entry into MODE 3              .
l
              !                  or 4.                                                                                        l
            .l
              \
l l
t i
: l.              i-                                                                                                    !
,                I l                \                                                                                                    !
                  \                                                                                                  l
                                                                                                                    /
L                                "The provisions of Specifications 4.4.5.2.2.b, 4.4.5.2.2.d, and 4.4.5.2.2.e l                      ;
are not applicable for valves UV 651, UV 652, UV 653 and UV 654 due to          /
l                        i position indication of valves in the control room.                              '
l                          \s                                                                                  /
                                \                                                                      /-
PALO VERDE - UNIT 1                  3/4 4-20 I
l
 
                                                                        -- - - ~
e                                        TABLE'3.4-1                _
s REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE                                      DESCRIPTION
: 1) SIE-V237                                LOOP 1A RC/SI CHECK-
: 2) SIE-V247                              LOOP 18 RC/SI CHECK                            !
: 3) SIE-V217                              LOOP 2A RC/SI CHECK
: 4) SIE-V227                                LOOP 28 RC/SI CHECK
: 5) SIE-V235                                LOOP 1A SIT CHECK                                    -
i                                6)- SIE-V245                              LOOP 18 SIT CHECK
                                '7) SIE-V215                                LOOP 2A SIT CHECK                                .
_. y                        8) SIE-V225                                LOOP 28 SIT CHECK                                  .
i                      9) SIE-V542                                LOOP 1A SI HEADER CHECK
: 10) SIE-V543                                LOOP 18 SI HEADER CHECK
#~
: 11) .SIE-V540                              LOOP 2A SI HEADER CHECK i                  12) SIE-V541                                LOOP 28 SI HEADER CHECK
{                13) SIA-V522                                LOOP 1 HP LONG TERM RECIRCULATION CHECK j              14) SIA-VS23                                LOOP 1 HP LONG TERM RECIRCULATION CHECK l              3-              15)-SIB-V532                                LOOP 2 HP LONG TERM RECIRCULATION CHECK
!      (        I            16) SIB-V533                                LOOP 2 HP LONG TERM RECIRCULATION CHECK
: 17) SIA-UV651*,#
                  \'                                .                    LOOP 1 SHUTOOWN COOLING ISOLATION 1
: 18)    SIB-UV652*,#                          LOOP 2 SHUTDOWN C00 LING' ISOLATION
: 19) SIC-UV653*,#                            LOOP 1 SHUTDOWN COOLING ISOLATION
: 20) SID-UV654*,#                            LOOP 2 SHUTDOWN COOLING ISOLATION
* Testing per Specification 4.4.5.2.2.d is not applicable due to positive indica-tion of valve position in the control room.
                                #1. Leakage rates greater than 1.0 gpa but less than or equa1 to 5.0 gpm are
                        ,            considered acceptable if the latest measured rate has not exceeded the
                                  ,  rate determined by previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm j                by 50% or greater, j        -2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpa are I              considered unacceptable if the latest measured rate exceeded the rate i            determined by the previous test by an amount that reduces the margin i
between measured leakage rate and the maximum permissible rate of 5.0 gpm
{            by 50% or greater.                                                    .
: 3. Leakage rates greater than 5.0 gpm are considered unacceptable.
,                    ;        PALO VERGE - UNIT 1                    3/4 4-21
!                    l-                                                                                            '
i
    .                    N
                                ~%
 
                                                                                                                  +
                    ,-                ~
        .s    bEACTORCOOLANTSYSTEM
      ~
              %x                          --
3/4.4.6 CHEMISTRY.
LIMITING CONDITION FOR OPERATION W
                .3.4.6.~ The Reactor Coolant System chemistry shall be maintained within the h
limits specified in Table 3.4-2.T      ,
f                APPLICA8ILITY: At all times.
ACTION:
MODES 1, 2,.3, and 4:
: a. With any one or more' chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Steady State Limit within 24 hours or be in at least HOT
                                -STAN08Y within the next 6 hours and in COLD SHUT 00WN within the following 30 hours.
: b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours.
At All Other Times:
With the concentration of either chloride or fluoride 'in the Reactor                      f Coolant System in excess of its Steady State Limit for more than 24 hours.
or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psia, if applicable, and perform an engineering                  -k evaluation to determine the effects of the out-of-limit condition on the
            .            structural integrity of the Reactor Coolant' System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psia or prior to proceeding to MODE 4.
* SURVEILLANCE REQUIREMENTS 4.4.6 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.
I i
                                    ~
R~,                        . -
3/4 4-22 PALO VERDE - UNIT 1 [
l:      . .      .    .
L
 
r, TABLE 3.4 2'I' REACTOR COOLANT SYSTEM CHEMISTRY STEADY STATE      TRANSIENT PARAMETER                                      LIMIT            LIMIT a
DISSOLVED OXYGEN                          1 0.10 ppe          5 1.00 ppm CHLORIDE                                  1 0.15 ppa          1 1.50 ppe FLUORIDE                                  5 0.10 ppe          5 1.00 ppm cold less than or equal to 250*F.
                " Limit not applicable with T 4
                          ^*,,,
PALO VERDE - UNIT 1  ,s                3/4 4-23 m *
* ep
 
TABLE 4.4 C3''-
REACTOR COOLANT SYSTEM                                i-CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS SAMPLE AND                                                                                        ;
              - PARAMETER                                              ANALYSIS FREQUENCY                                                                                    j i
                                        "                                                                                                                                    l DISSOLVED OXYGEN                                    At least once per 72 hours                                                                              i I
CHLORIDE                                            At least once per 72 hours FLUORIDE                                            At least once per 72 hours eold less than or equal to 250*F "Not required with T t
9 e
8                    9 j+                  '' ,,
PALO VERDE - UNIT 1                  ;
3/4 4-24                                              -
<--+r    ,        s      -  , - - - -        _                  _                  , . . , , . __ ., - - , - , . - - , , + , _ , , . . , - - - - , . -      ,,-y, _,~ -
 
I
                        -              ~
                                                                                                              !)
j OpCREACTORCOOLANTSYSTEM).                                                                                j
                    ~
3/4.4.7 SPECIFIC ACTIVITY
['
LIMITING CONDITION FOR OPERATION 3.4.7 The specific activity of the primary coolant shall be limited to:
: a.        Less than or equal to 1.0 microcurie / gram DOSE EQUIVALENT I-131, and
: b.        Less than or equal to 100/l microcuries/gran.
* APPLICA8ILITY: MODES 1, 2, 3, 4, and 5.                      .
ACTION:
MODES 1, 2, and 3*:
: a.        With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.4-1, operation may continue for up to 48 hours provided that the cumulative operating time under these circumstances does not exceed 800 hours in any consecutive 12 month period. With the total cumulative operating time at a primary coolant specific activity greater than 1.0 microcurie / gram DOSE EQUIVALENT I-131 exceeding
(.
500 hours in any consecutive 6 month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2      -
within 30 days indicating the number of hours above'this limit. The provisions of Specification 3.0.4 are not applicable.                        .
: b.        With the specific activity of the primary coolant greater than 1.0 microcurie / gram DOSE EQUIVALENT'I-131 for more than 48 hours during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T than 500*F within 6 hours.                                cold less
: c.        With the specific activity of the primary coolant greater than 100/E microcuries/ gram, be in at least HOT STAN08Y with T      less than 500*F within 6 hours.                                Eold i
            "With cold greater than or equal to 500*F.
                    +-******="w-,            ,
s PALO VERDE - UNIT 1
                                            } '
3/4 4-25
 
R COOLANT SYS 4
LIMITING CONDITION FOR OPERATION (Continued)
                                                                              ~
ACTION: (Continued)
MODES 1, 2, 3, 4, and 5:                      ,
: d. With the specific activity of the primary coolant greater than 1 microcurie / gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries/ gram, perform th ' sampling and analysis require-
            @          ments of item 4.(a) of Table 4.4          until the specific activity of the primary coolant is restored to within its limits. A Special Report shall be prepared and snhraitted to the Commission pursuant to Speci-            '
fication 6.9.25within 30 days with a copy to the Director, Nuclear Reactor Megulation, Attention: Chief, Core Performance Branch, and g          Chief, Accident Evaluation Branch, U.S. Nuclear Regulatory Commis-tion. Washington, D.C. 20555.f This report,snais contain the results of the specific activity analyses together with the following information:
: 1. Reactor power history starting 48 hours prior to the first sample in which the limit was exceeded,
: 2. Fuel burnup by core region,
: 3. Clean-up flow history starting 48 hours prior to the first sample in which the limit was exceeded,
: 4. History of degassing operation, if any, starting 48 hours prior to the first sample in which the limit was exceeded, and
: 5. The time duration when the specific activity of the primary coolant exceeded 1 microcurie / gram DOSE EQUIVALENT I-131.              .
I SURVEILLANCE REQUIREMENTS
                                                                                ^
i            4.4.7 Thespecificactivityoftheprimarycoolantshallbebeterminedtobe within the limits by performance of the sampling and analysis program of
  @        Table 4.4-@ @
e a
PALO VERDE - UNIT 1                      3/4 4-26 d
me 960
 
TABLE 4.44)
[g l
l i      ,    5                                            PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE h                                                      AND ANALYSIS PROGRAM i
;              7        TYPE OF MEASUREMENT                                        SAMPLE AND ANALYSIS          H0 DES IN )AlICH SAMPLE 1
c-  j        AND ANALYSIS                                                  FREQUENCY            AND ANALYSIS REQUIRED
!              a
:              * /      1. Gross Ac'tivity Determination            At least once per 72 hours                I , 2, 3, 4 l            7:
{                        2. Isotopic Analysis for DOSE              1 per 14 days                              1 EQUIVALENT I-131 Concentration i
j                        3. Radiochemical for E Determination        1 per 6 months
* 1 i
: 4. Isotopic Analysis for Iodine            (a) Once per 4 hours,                      1#, 2#,,3#, 4#, 5#
l                              Including I-131, I-133, and I-135                      whenever the specific
,                                                                                    activity exceeds
!              1:*                                                                    1.0 pCi/ gram, DOSE
                                                                                    . EQUIVALENT I-131
.            t                                                ,
or 100/E pCi/ gram, and
!              U (b) One sample'between                    1, 2, 3
:                                                                                    2 and 6 hours following I
a THERMAL POWER
,                                                                                    change exceeding 15%
i                                              -
of the RATED THERMAL j      ,                                                                              POWER within a 1-hour
!                                                                                    period. One sample is i                                                                                    sufficient if plant has gone through a SHUTDOWN
:l                                                                                    or if transient is complete in 6 hours.
{
i l                    # Until the specific activity of the primary coolant system is restored within its limits.
1 j
* Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours or longer.
l s                                        .
1
__~_1._._-
 
1 A
s                                                            .          .
I.
_                            x              ..                    . ..          ',              .            ..,              . ...                          ..              .          ...
g                                i    .                          . ..              .      ..              .        .      . ..                                .          .              i.
ay                                  t.              .              , ,.                ,...                .                    . ...                          ..,,                    ..i l.
gg                              . x x.
3_                  .            .                              .                      ....                ..            .      i      .              .      .            ..        ....
                                  .            .        . s.                  .        ..            .;ii                    ,..              .; i.                                                                                                              ;
p            ' ..            .        . y .                  . . .                  ....                . ...                ....                          ....                    ....
i l-              5                ...                  ,    ,  .'x
                                                                                                                          . .i.
ii              .'            .
a                                                      x                      .        ..          .      .                    .. ..                          ...,
p                                                          t.          .              .      .              ,                                ..              .                        ..          ,
1 p                                                            g          ..            ..                  . .          .      . ...                                i.,                    ...
,              g                              .
x            .
i X                                        .
,              E c
00                                  ,            .          y i
: s.                        ..                    ' UNACCEPTABLE                                                        '
o i
                                                                              . . . x
                                                                                        .x            . ...
                                                                                                                          . ... OPERATION
                                                                                                                                                ..                      .      ..                              e.
g'                ,            .      . ...                  . . .          .            . ..          . ..                                  ..            ....                    ....
                                        ...            . ...                  . .          .  .ny . . .                                      .
g                                                                                                                                                              ...,                    ....
                                                                                                                          .        .,                          i.
                                                                    .                  , ,            xi .              i        ..
;.      ,      c., ,                          .                                        .            .x.                          ..                                                .                  ..
i              u                                                              . . .                        .x            4  .        .      . ..                                        ..                  ..          -
I l
k z 150
              <i
                                                                                                      ....t
                                                                                                          .' .'x    l .
                                                                                                              ..(.            ...
o                      ,..            . ...                  . . . .
                                                                                                      . . .,          .s...
C                              .                . .          . . . ,
u                .                              ,            , , , ,
xi . .
1.,
t g
                                      .      .      . . .                  . . . .                    .e .          ..N .                .                      .      .            ..
                        .            .      .      . . .                  . . . .                      . ..          i .s,                .. ,                          .                      ...,
4                ....                      .
                                                                              . . ..                      i t .        .    ..x              . . . ..                      ...                    ....
3                .            .        .      . .                . . .            ..i.                .    ...x            ....                                ...                    ..
g 100            ,,,.                  , , ,                  . , , .                ,,,.                . ..,              g...                                  ...                    ..                                .,,
6                .
                                          ..          . .,.                  . . ..                . ...                ....                x...                                ...              ...
              ,                                                                                          .i;                                  X. ..                                      i          ...,
guy              .        ..          .    .      .        s . , .                          ..        ...i                  .\i                            .,..                    ..,,
                                                                              . .            .      .                  . ...                .y                      i      ..,                    .. .
1
                                ..                                                . . .            . ...                ..                    . . s,                        ..                .    ....                          -
g                            .        i ........                                    .            .      .,          .      .. 1                          ....                    ,,,,
a            '-                      ACCEPTABLE                                    . ..                *''.                . ...i                                          *          .
3              l                      OPERATION                                    llll                                      .      l.'                    ',.                    .l.'
50      ,
                                                      . .                                          . .          .      ....                .                      .      i...                          ...
5              *            -        ' '-                        - - -            -        .-              . .-                                          .-
e w
            . m                      .      .        ...                    . ...                . ...                              .      ,,,                                        ..        ....
                              .        ..          6 ..                                                ,,.
: 8.              ...,                  . ..              i.
t 6 e .
i ..
0                                                                                                                                                                                      '
20                    30                40                        50                  60                  70                            80                        90              100 PERCENT OF RATED THERMAL POWER                                                                                                                                        .
FIGURE 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS .                                                                                                                                                                                      ,
PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC                                                                                                                                                                                            i r
ACTIVITY > 1.0 uCi/ GRAM DOSE EQUIVALENT I-131
* PALO VERDE - UNIT 1                                                                        3/4 4-28 l
u  . ..
l l
L
 
j 4
                                              ~
g i, REACTOR COOLANT SYSTEM ,
t 3/4.4.8 -PRESSURE / TEMPERATURE LIMITS
                                                                                                              -l 4
REACTOR COOLANT SYSTEM
                  -LIMITING CONDITION FOR OPERATION-3.4.8.1' The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak            Lp
  ',              and hydrostatic testing with:
Nk 6      A maximum heatup rate of 20*F per hour with the RCS cold leg temper ',
I        ature less than or equal to 95'F, 40*F per hour with RCS cold leg temperature greater than 95'F but less than or equal to 400*F, and 100*F per hour with RCS cold leg temperature greater than 400*F.
                    !  b. A maximum cooldown rate of 10*F per hour with RCS cold leg temperature less than or equal to 100*F, 40*F per hour with RCS cold leg i
temperature greater than 100*F but less than or equal to 130*F, and 100*F per hour with RCS cold leg temperature greater than 130*F.
: c. A maximum temperature change of 10*F in any 1-hour period during inservice hydrostatic and leak testing operations.
APPLICA81LITY: At all times.*
                ' ACTION:                              -
With any of the above limits exceeded, restore the temperature and/or pressure            .
                . to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANOBY within the next 6 hours and reduce the RCS Teold and pressure to.less than 210*F and 500 psia, respectively, within the following 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System tesperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operation,s.
: 4. 4. 8.1. 2 The reactor vessel material irradiation surveillance specimens shall be -removed and examined, to determine changes in material properties, at the intervals required by 10 CFR Part 50 Appendix H in accordance with the 8~) schedule in Table 4.4-1. The results of these examinations shall be used to update Figure 3.4-2. 3                                                _
                  "see special Test Exception 3.10.5.
PALO VERDE - UNIT 1                            3/4 4-29                                  .
D
 
3/4.4.'8        PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM'                    -
                                                                                          ')
i LIMITING CONDITION FOR OPERATION 3.4'.8.1  The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in 'accordance with the limit lines shown on Figure 3.4-2 during heatup, cooldown, criticality, and inservice leak and hydrostatic testingwjh:                                          _.
  ,/  a. A maximum heatup rate of 100*F/hr.                          'N j      b. A maximum cooldown rate of 100*F/hr.                                  )
\'                                                                                  1
  \:  c. A maximum temperature change of 10'F'in any one hour period during /
    \        inservice hydrostatic.and leak testing operations above the heatup/
(      and cooldown limit curves.                                        _f APPLICABILITY: At all times.                                    -
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perfonn an engineering evaluation to
' determine the effects of the out-of-limit condition on the fracture toughness          ,'N properties of the Reactor Coolant System; determine that the Reactor Coolant            '
                                                                                            )
System remains acceptable for continued operations or be .in at least HOT STANOBY within the next 6 hours and reduce ttie RCS T      and pressure to less
* than210*Fand500 psia,respectively,withinthefofibNing30 hours.
SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown,'and irservice leak and hydrostatic testing operations.
4.4.8.1.2 The reactor ' essel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50 Appendix H in accordance with the schedule
.in Table 4.4-3. The results of these examinations shall be used to update Figure 3.4-2.
Amendment Number 9 February 27, 1984 3/4 4-20
 
                                                                                                                                                                                                                                    .t w
V m
2 o
INDICATED PRESSURE (PSIA)                                                                                                      ,
                                                                                                                                                                                                                                                                                                                                                      -~.'%                                                  -
                    <                                                                                                                                                                                                                                                                  n                                                              n                                                                u%
m                                                                                                                                        a                                                          a o                                                                                                                                              s
                    =                                                                            Un                                                          o                                                                                                                                                                                                                                                                            N o                                                                            o m                                                                            o                                                          8                                                                                                                        8-                                                                                                                                                          x
                                      =
1            I                  1                .I                                I        I I
i                        1            I        .I        -      I
                                                                                                                                                                                                                                                                                                                                              ...                      ,        ...              . .I              :1    .ip      .                \
is n:                                                                        '
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:; . .l,;                :.;
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                                                                                                                                                                                                                                                                                                                                                              .        .j.I "h;:;::i.t;;:nl i I.:
:!:-:1 ;l..:-
t!.:
                    -i                m
                                                                                                                          .:                        .. .                              1. ..:                                      .                            .:..            ..;                                              ...: .:.:                  .
:          .    .ta                        .
e                                                                                                                                                                                                                                                                        ., t.!.
g                          .
I                                        ;-                                                            :..                                                  ,
                                                                                                                                                                                                                                                                                                                        ..        :: :- ;::: r . ..ll                              g,                            i                  1-g.
i m
                                                                                                                                                                                      .:. a ..;                                                                    ..
                                                                                                                                                                                                                                                                                                                              .                .j:.        :t-          i.        ,
                                                                                                                                                                                                                                                                                                                                                                                                -        l    :        I
                                                                                                                                                                                                                                                                                                                                                                                                                        ,: I l w                    -                                                -
:r; I                                                                            -            -
                                                                              . .I IrA i
e m                  .                                                                                          .
                                                                                                                                                                                                                                                                                            -            -.. :i i. ;.l::
                                                                                                                                                                                                                                                                                                                              ; rp;t r j;ti            .,n::r ji:l i;? ! li              :}}

Latest revision as of 16:22, 12 December 2024

Forwards Proposed Final Version of CESSAR Chapter 16 Sys 80 Tech Specs,Per 850502 Meeting W/Nrc.Completion of Review of Tech Specs Requested by End of June 1985 to Include Revs in Amend 10 to CESSAR
ML20127H458
Person / Time
Site: 05000470
Issue date: 05/16/1985
From: Scherer A
ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To: Thompson H
Office of Nuclear Reactor Regulation
References
LD-85-024, LD-85-24, NUDOCS 8505210305
Download: ML20127H458 (868)
v • d • e


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